WorldWideScience

Sample records for passive reactor plants

  1. Role of Passive Safety Features in Prevention And Mitigation of Severe Plant Conditions in Indian Advanced Heavy Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jain, Vikas; Nayak, A.; Dhiman, M.; Kulkarni, P. P.; Vijayan, P. K.; Vaze, K. K. [Bhabha Atomic Research Centre, Mumbai (India)

    2013-10-15

    Pressing demands of economic competitiveness, the need for large-scale deployment, minimizing the need of human intervention, and experience from the past events and incidents at operating reactors have guided the evolution and innovations in reactor technologies. Indian innovative reactor 'AHWR' is a pressure-tube type natural circulation based boiling water reactor that is designed to meet such requirements, which essentially reflect the needs of next generation reactors. The reactor employs various passive features to prevent and mitigate accidental conditions, like a slightly negative void reactivity coefficient, passive poison injection to scram the reactor in event of failure of the wired shutdown systems, a large elevated pool of water as a heat sink inside the containment, passive decay heat removal based on natural circulation and passive valves, passive ECC injection, etc. It is designed to meet the fundamental safety requirements of safe shutdown, safe decay heat removal and confinement of activity with no impact in public domain, and hence, no need for emergency planning under all conceivable scenarios. This paper examines the role of the various passive safety systems in prevention and mitigation of severe plant conditions that may arise in event of multiple failures. For the purpose of demonstration of the effectiveness of its passive features, postulated scenarios on the lines of three major severe accidents in the history of nuclear power reactors are considered, namely; the Three Mile Island (TMI), Chernobyl and Fukushima accidents. Severe plant conditions along the lines of these scenarios are postulated to the extent conceivable in the reactor under consideration and analyzed using best estimate system thermal-hydraulics code RELAP5/Mod3.2. It is found that the various passive systems incorporated enable the reactor to tolerate the postulated accident conditions without causing severe plant conditions and core degradation.

  2. ROLE OF PASSIVE SAFETY FEATURES IN PREVENTION AND MITIGATION OF SEVERE PLANT CONDITIONS IN INDIAN ADVANCED HEAVY WATER REACTOR

    Directory of Open Access Journals (Sweden)

    VIKAS JAIN

    2013-10-01

    Full Text Available Pressing demands of economic competitiveness, the need for large-scale deployment, minimizing the need of human intervention, and experience from the past events and incidents at operating reactors have guided the evolution and innovations in reactor technologies. Indian innovative reactor ‘AHWR’ is a pressure-tube type natural circulation based boiling water reactor that is designed to meet such requirements, which essentially reflect the needs of next generation reactors. The reactor employs various passive features to prevent and mitigate accidental conditions, like a slightly negative void reactivity coefficient, passive poison injection to scram the reactor in event of failure of the wired shutdown systems, a large elevated pool of water as a heat sink inside the containment, passive decay heat removal based on natural circulation and passive valves, passive ECC injection, etc. It is designed to meet the fundamental safety requirements of safe shutdown, safe decay heat removal and confinement of activity with no impact in public domain, and hence, no need for emergency planning under all conceivable scenarios. This paper examines the role of the various passive safety systems in prevention and mitigation of severe plant conditions that may arise in event of multiple failures. For the purpose of demonstration of the effectiveness of its passive features, postulated scenarios on the lines of three major severe accidents in the history of nuclear power reactors are considered, namely; the Three Mile Island (TMI, Chernobyl and Fukushima accidents. Severe plant conditions along the lines of these scenarios are postulated to the extent conceivable in the reactor under consideration and analyzed using best estimate system thermal-hydraulics code RELAP5/Mod3.2. It is found that the various passive systems incorporated enable the reactor to tolerate the postulated accident conditions without causing severe plant conditions and core degradation.

  3. Passive Safety Features for Small Modular Reactors

    International Nuclear Information System (INIS)

    Ingersoll, Daniel T.

    2010-01-01

    The rapid growth in the size and complexity of commercial nuclear power plants in the 1970s spawned an interest in smaller, simpler designs that are inherently or intrinsically safe through the use of passive design features. Several designs were developed, but none were ever built, although some of their passive safety features were incorporated into large commercial plant designs that are being planned or built today. In recent years, several reactor vendors are actively redeveloping small modular reactor (SMR) designs with even greater use of passive features. Several designs incorporate the ultimate in passive safety they completely eliminate specific accident initiators from the design. Other design features help to reduce the likelihood of an accident or help to mitigate the accidents consequences, should one occur. While some passive safety features are common to most SMR designs, irrespective of the coolant technology, other features are specific to water, gas, or liquid-metal cooled SMR designs. The extensive use of passive safety features in SMRs promise to make these plants highly robust, protecting both the general public and the owner/investor. Once demonstrated, these plants should allow nuclear power to be used confidently for a broader range of customers and applications than will be possible with large plants alone.

  4. Balancing passive and active systems for evolutionary water cooled reactors

    International Nuclear Information System (INIS)

    Fil, N.S.; Allen, P.J.; Kirmse, R.E.; Kurihara, M.; Oh, S.J.; Sinha, R.K.

    1999-01-01

    Advanced concepts of the water-cooled reactors are intended to improve safety, economics and public perception of nuclear power. The potential inclusion of new passive means in addition or instead of traditional active systems is being considered by nuclear plant designers to reach these goals. With respect to plant safety, application of the passive means is mainly intended to simplify the safety systems and to improve their reliability, to mitigate the effect of human errors and equipment malfunction. However, some clear drawbacks and the limited experience and testing of passive systems may raise additional questions that have to be addressed in the design process for each advanced reactor. Therefore the plant designer should find a reasonable balance of active and passive means to effectively use their advantages and compensate their drawbacks. Some considerations that have to be taken into account when balancing active/passive means in advanced water-cooled reactors are discussed in this paper. (author)

  5. Passive systems for light water reactors

    International Nuclear Information System (INIS)

    Adinolfi, R.; Noviello, L.

    1990-01-01

    The paper reviews the most original concepts that have been considered in Italy for the back-fitting of the nuclear power plants in order to reduce the probability and the importance of the release to the environment in case of a core melt. With reference either to BWR or PWR, passive concepts have been considered for back-fitting in the following areas: pump seals damage prevention and ECCS passive operation; reactor passive depressurization; molten reactor core passive cooling; metal containment passive water cooling through a water tank located at high level; containment isolation improvement through a sealing system; containment leaks control and limitation of environmental release. In addition some considerations will be made on the protection against external events introduced from the beginning on the PUN design either on building and equipment lay-out either on structure design. (author). 5 figs

  6. PSA in design of passive/active safety reactors

    International Nuclear Information System (INIS)

    Sato, T.; Tanabe, A.; Kondo, S.

    1995-01-01

    PSAs in the design of advanced reactors are applied mainly in level 1 PSA areas. However, even in level 1 PSA, there are certain areas where special care must be taken depending on plant design concepts. This paper identifies these areas both for passive and active safety reactor concepts. For example, 'long-term PSA' and shutdown PSA are very important for a passive safety reactor concept from the standpoint of effectiveness of a grace period and passive safety systems. External events are also important for an active safety reactor concept. These kinds of special PSAs are difficult to conduct precisely in a conceptual design stage. This paper shows methods of conducting these kinds of special PSAs simply and conveniently and the use of acquired insights for the design of advanced reactors. This paper also clarifies the meaning or definition of a grace period from the standpoint of PSA

  7. Interim results of the study of control room crew staffing for advanced passive reactor plants

    International Nuclear Information System (INIS)

    Hallbert, B.P.; Sebok, A.; Haugset, K.

    1996-01-01

    Differences in the ways in which vendors expect the operations staff to interact with advanced passive plants by vendors have led to a need for reconsideration of the minimum shift staffing requirements of licensed Reactor Operators and Senior Reactor Operators contained in current federal regulations (i.e., 10 CFR 50.54(m)). A research project is being carried out to evaluate the impact(s) of advanced passive plant design and staffing of control room crews on operator and team performance. The purpose of the project is to contribute to the understanding of potential safety issues and provide data to support the development of design review guidance. Two factors are being evaluated across a range of plant operating conditions: control room crew staffing; and characteristics of the operating facility itself, whether it employs conventional or advanced, passive features. This paper presents the results of the first phase of the study conducted at the Loviisa nuclear power station earlier this year. Loviisa served as the conventional plant in this study. Data collection from four crews were collected from a series of design basis scenarios, each crew serving in either a normal or minimum staffing configuration. Results of data analyses show that crews participating in the minimum shift staffing configuration experienced significantly higher workload, had lower situation awareness, demonstrated significantly less effective team performance, and performed more poorly as a crew than the crews participating in the normal shift staffing configuration. The baseline data on crew configurations from the conventional plant setting will be compared with similar data to be collected from the advanced plant setting, and a report prepared providing the results of the entire study

  8. Plant experience with check valves in passive systems

    Energy Technology Data Exchange (ETDEWEB)

    Pahladsingh, R R [GKN Joint Nuclear Power Plant, Dodewaard (Netherlands)

    1996-12-01

    In the design of the advanced nuclear reactors there is a tendency to introduce more passive safety systems. The 25 year old design of the GKN nuclear reactor is different from the present BWR reactors because of some special features, such as the Natural Circulation - and the Passive Isolation Condenser system. When reviewing the design, one can conclude that the plant has 25 years of experience with check valves in passive systems and as passive components in systems. The result of this experience has been modeled in a plant-specific ``living PSA`` for the plant. A data-analysis has been performed on components which are related to the safety systems in the plant. As part of this study also the check valves have been taken in consideration. At GKN, the check valves have shown to be reliable components in the systems and no catastrophic failures have been experienced during the 25 years of operation. Especially the Isolation Condenser with its operation experience can contribute substantially to the insight of check valves in stand-by position at reactor pressure and operating by gravity under different pressure conditions. With the introduction of several passive systems in the SBWR-600 design, such as the Isolation Condensers, Gravity Driven Cooling, and Suppression Pool Cooling System, the issue of reliability of check valves in these systems is actual. Some critical aspects for study in connection with check valves are: What is the reliability of a check valve in a system at reactor pressure, to open on demand; what is the reliability of a check valve in a system at low pressure (gravity), to open on demand; what is the reliability of a check valve to open/close when the stand-by check wave is at zero differential pressure. The plant experience with check valves in a few essential safety systems is described and a brief introduction will be made about the application of check valves in the design of the new generation reactors is given. (author). 6 figs, 1 tab.

  9. Materials for passively safe reactors

    International Nuclear Information System (INIS)

    Simnad, T.

    1993-01-01

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

  10. Conceptual benefits of passive nuclear power plants and their effect on component design

    International Nuclear Information System (INIS)

    DeVine, J.C. Jr.

    1996-01-01

    Today, nearly ten years after the advanced light water reactor (ALWR) Program was conceived by US utility leaders, and a decade and a half since a new nuclear power plant was ordered in the US, the ALWR passive plant is coming into its own. This design concept, a midsized simplified light water reactor, features extremely reliable passive systems for accident prevention and mitigation and combines proven experience with state-of-the-art engineering and human factors. It is now emerging as the front runner to become the next generation reactor in the US and perhaps around the world. Although simple and straightforward in concept, the passive plant is in many respects a significant departure from previous trends in reactor engineering. Successful implementation of this concept presents numerous challenges to the designers of passive plant systems and components. This paper provides a brief history of the ALWR program, it outlines the ALWR passive plant design objectives and principles, and it summarizes with examples their implications on component design. (orig.)

  11. Application of heat pipes in nuclear reactors for passive heat removal

    Energy Technology Data Exchange (ETDEWEB)

    Haque, Z.; Yetisir, M., E-mail: haquez@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2013-07-01

    This paper introduces a number of potential heat pipe applications in passive (i.e., not requiring external power) nuclear reactor heat removal. Heat pipes are particularly suitable for small reactors as the demand for heat removal is significantly less than commercial nuclear power plants, and passive and reliable heat removal is required. The use of heat pipes has been proposed in many small reactor designs for passive heat removal from the reactor core. This paper presents the application of heat pipes in AECL's Nuclear Battery design, a small reactor concept developed by AECL. Other potential applications of heat pipes include transferring excess heat from containment to the atmosphere by integrating low-temperature heat pipes into the containment building (to ensure long-term cooling following a station blackout), and passively cooling spent fuel bays. (author)

  12. The Westinghouse AP1000 plant design: a generation III+ reactor with unique proven passive safety technology

    International Nuclear Information System (INIS)

    Demetri, K. J.; Leipner, C. I.; Marshall, M. L.

    2015-09-01

    The AP1000 plant is an 1100-M We pressurized water reactor with passive safety features and extensive plant simplifications and standardization that simplify construction, operation, maintenance, safety, and cost. The AP1000 plant is based on proven pressurized water reactor (PWR) technology, with an emphasis on safety features that rely solely on natural forces. These passive safety features are combined with simple, active, defense-in-depth systems used during normal plant operations which also provide the first level of defense against more probable events. This paper focuses on specific safety and licensing topics: the AP1000 plant robustness to be prepared for extreme events that may lead to catastrophic loss of infrastructure, such as the Fukushima Dai-ichi event, and the AP1000 plant compliance with the safety objectives for new plants. The first deployment of the AP1000 plant formally began in July 2007 when Westinghouse Electric Company and its consortium partner, the Shaw Group, signed contracts for four AP1000 units on coastal sites of Sanmen and Haiyang, China. Both sites have the planned ability to accommodate at least six AP1000 units; construction is largely concurrent for all four units. Additionally, the United States (U.S.) Nuclear Regulatory Commission (NRC) issued combined licenses (COLs) to allow Southern Nuclear Operating Company (SNC) and South Carolina Electric and Gas Company (SCE and G) to construct and operate AP1000 plants. Within this paper, the various factors that contribute to an unparalleled level of design, construction, delivery, and licensing certainty for any new AP1000 plant projects are described. These include: 1) How the AP1000 plant design development and reviews undertaken in the United States, China and Europe increase licensing certainty. 2) How the AP1000 passive plant robustness against extreme events that result in large loss of infrastructure further contributes to the licensing certainty in a post

  13. The Westinghouse AP1000 plant design: a generation III+ reactor with unique proven passive safety technology

    Energy Technology Data Exchange (ETDEWEB)

    Demetri, K. J.; Leipner, C. I.; Marshall, M. L., E-mail: demetrkj@westinghouse.com [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2015-09-15

    The AP1000 plant is an 1100-M We pressurized water reactor with passive safety features and extensive plant simplifications and standardization that simplify construction, operation, maintenance, safety, and cost. The AP1000 plant is based on proven pressurized water reactor (PWR) technology, with an emphasis on safety features that rely solely on natural forces. These passive safety features are combined with simple, active, defense-in-depth systems used during normal plant operations which also provide the first level of defense against more probable events. This paper focuses on specific safety and licensing topics: the AP1000 plant robustness to be prepared for extreme events that may lead to catastrophic loss of infrastructure, such as the Fukushima Dai-ichi event, and the AP1000 plant compliance with the safety objectives for new plants. The first deployment of the AP1000 plant formally began in July 2007 when Westinghouse Electric Company and its consortium partner, the Shaw Group, signed contracts for four AP1000 units on coastal sites of Sanmen and Haiyang, China. Both sites have the planned ability to accommodate at least six AP1000 units; construction is largely concurrent for all four units. Additionally, the United States (U.S.) Nuclear Regulatory Commission (NRC) issued combined licenses (COLs) to allow Southern Nuclear Operating Company (SNC) and South Carolina Electric and Gas Company (SCE and G) to construct and operate AP1000 plants. Within this paper, the various factors that contribute to an unparalleled level of design, construction, delivery, and licensing certainty for any new AP1000 plant projects are described. These include: 1) How the AP1000 plant design development and reviews undertaken in the United States, China and Europe increase licensing certainty. 2) How the AP1000 passive plant robustness against extreme events that result in large loss of infrastructure further contributes to the licensing certainty in a post

  14. Reactor core and passive safety systems descriptions of a next generation pressure tube reactor - mechanical aspects

    Energy Technology Data Exchange (ETDEWEB)

    Yetisir, M.; Gaudet, M.; Rhodes, D.; Hamilton, H.; Pencer, J. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    Canada has been developing a channel-type supercritical water-cooled nuclear reactor concept, often called the Canadian SCWR. The objective of this reactor concept is to meet the technology goals of the Generation IV International Forum (GIF) for the next generation nuclear reactor development, which include enhanced safety features (inherent safe operation and deploying passive safety features), improved resource utilization, sustainable fuel cycle, and greater proliferation resistance than Generation III nuclear reactors. The Canadian SCWR core concept consists of a high-pressure inlet plenum, a separate low-pressure heavy water moderator contained in a calandria vessel, and 336 pressure tubes surrounded by the moderator. The reactor uses supercritical water as a coolant, and a direct steam power cycle to generate electricity. The reactor concept incorporates advanced safety features such as passive core cooling, long-term decay heat rejection to the environment and fuel melt prevention via passive moderator cooling. These features significantly reduce core damage frequency relative to existing nuclear reactors. This paper presents a description of the design concepts for the Canadian SCWR core, reactor building layout and the plant layout. Passive safety concepts are also described that address containment and core cooling following a loss-of coolant accident, as well as long term reactor heat removal at station blackout conditions. (author)

  15. Passive Decay Heat Removal System for Micro Modular Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Jangsik; Lee, Jeong Ik; Jeong, Yong Hoon [KAIST, Daejeon (Korea, Republic of)

    2015-10-15

    Dry cooling system is applied as waste heat removal system therefore it is able to consider wide construction site. Schematic figure of the reactor is shown in Fig. 1. In safety features, the reactor has double containment and passive decay heat removal (PDHR) system. The double containment prevents leakage from reactor coolant system to be emitted into environment. The passive decay heat removal system copes with design basis accidents (DBAs). Micros Modular Reactor (MMR) which has been being developed in KAIST is S-CO{sub 2} gas cooled reactor and shows many advantages. The S-CO{sub 2} power cycle reduces size of compressor, and it makes small size of power plant enough to be transported by trailer.The passive residual heat removal system is designed and thermal hydraulic (TH) analysis on coolant system is accomplished. In this research, the design process and TH analysis results are presented. PDHR system is designed for MMR and coolant system with the PDHR system is analyzed by MARS-KS code. Conservative assumptions are applied and the results show that PDHR system keeps coolant system under the design limitation.

  16. Passive safety systems for integral reactors

    International Nuclear Information System (INIS)

    Kuul, V.S.; Samoilov, O.B.

    1996-01-01

    In this paper, a wide range of passive safety systems intended for use on integral reactors is considered. The operation of these systems relies on natural processes and does not require external power supplies. Using these systems, there is the possibility of preventing serious consequences for all classes of accidents including reactivity, loss-of-coolant and loss of heat sink as well as severe accidents. Enhancement of safety system reliability has been achieved through the use of self-actuating devices, capable of providing passive initiation of protective and isolation systems, which respond immediately to variations in the physical parameters of the fluid in the reactor or in a guard vessel. For beyond design base accidents accompanied by complete loss of heat removal capability, autonomous self-actuated ERHR trains have been proposed. These trains are completely independent of the secondary loops and need no action to isolate them from the steam turbine plant. Passive safety principles have been consistently implemented in AST-500, ATETS-200 and VPBER 600 which are new generation NPPs developed by OKBM. Their main characteristic is enhanced stability over a wide range of internal and external emergency initiators. (author). 10 figs

  17. Passive safety systems for integral reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kuul, V S; Samoilov, O B [OKB Mechanical Engineering (Russian Federation)

    1996-12-01

    In this paper, a wide range of passive safety systems intended for use on integral reactors is considered. The operation of these systems relies on natural processes and does not require external power supplies. Using these systems, there is the possibility of preventing serious consequences for all classes of accidents including reactivity, loss-of-coolant and loss of heat sink as well as severe accidents. Enhancement of safety system reliability has been achieved through the use of self-actuating devices, capable of providing passive initiation of protective and isolation systems, which respond immediately to variations in the physical parameters of the fluid in the reactor or in a guard vessel. For beyond design base accidents accompanied by complete loss of heat removal capability, autonomous self-actuated ERHR trains have been proposed. These trains are completely independent of the secondary loops and need no action to isolate them from the steam turbine plant. Passive safety principles have been consistently implemented in AST-500, ATETS-200 and VPBER 600 which are new generation NPPs developed by OKBM. Their main characteristic is enhanced stability over a wide range of internal and external emergency initiators. (author). 10 figs.

  18. CLASSIFICATION OF SYSTEMS FOR PASSIVE AFTERHEAT REMOVAL FROM REACTOR CONTAINMENT OF NUCLEAR POWER PLANT WITH WATER-COOLED POWER REACTOR

    Directory of Open Access Journals (Sweden)

    N. Khaled

    2014-01-01

    Full Text Available A classification on systems for passive afterheat removal from reactor containment has been developed in the paper.  The classification permits to make a detailed analysis of various concepts pertaining to systems for passive afterheat removal from reactor containment of new generation. The paper considers main classification features of the given systems.

  19. Passive Nuclear Plants Program (UPDATE)

    International Nuclear Information System (INIS)

    Chimeno, M. A.

    1998-01-01

    The light water passive plants program (PCNP), today Advanced Nuclear Power Plants Program (PCNA), was constituted in order to reach the goals of the Spanish Electrical Sector in the field of advanced nuclear power plants, optimize the efforts of all Spanish initiatives, and increase joint presence in international projects. The last update of this program, featured in revision 5th of the Program Report, reflects the consolidation of the Spanish sector's presence in International programs of the advanced power plants on the basis of the practically concluded American ALWR program. Since the beginning of the program , the PCNP relies on financing from the Electrical sector, Ocide, SEPI-Endesa, Westinghouse, General Electric, as well as from the industrial cooperators, Initec, UTE (Initec- Empresarios Agrupados), Ciemat, Enusa, Ensa and Tecnatom. The program is made up of the following projects, already concluded: - EPRI's Advanced Light Water Plants Certification Project - Westinghouse's AP600 Project - General Electric's SBWR Project (presently paralyzed) and ABWR project Currently, the following project are under development, at different degrees of advance: - EPP project (European Passive Plant) - EBWR project (European Advanced Boiling Water Reactor)

  20. EP 1000 -The European Passive Plant

    International Nuclear Information System (INIS)

    Cummins, Ed; Oyarzabal, Mariano; Saiu, Gianfranco

    1998-01-01

    A group of European utilities, along with Westinghouse and its industrial partner GENESI (an Italian consortium including ANSALDO and FIAT) initiated a program to evaluate Westinghouse passive nuclear plant technology for application in Europe. The European utility group consisted of: Agrupacion electrica para al Desarrollo Technologico Nuclear (DTN), Spain; Electricite de France; ENEL, SpA., Italy; IVO Power Engineering, Ltd., Finland; Scottish Nuclear Limited (acting for itself on behalf of Nuclear Electric plc, U.K.; Tractebel Energy Engineering, Belgium; UAK (represented by NOK-Beznau), Switzerland; and Vattenfall AB, Ringhals, Sweden. The European Passive Plant (EPP) program, which began in 1994, is an evaluation of the Westinghouse 600 MWe AP 600 and 1000 MWe Simplified Pressurized Water Reactor (SPWR) designs in meeting the European Utility Requirements (EUR), and where necessary, modifying the design to achieve compliance. Phase 1 or the EPP program was completed and included the two major tasks of evaluating the effect of the EUR on the Westinghouse nuclear island and developing the EP 1000, a 1000 MWe passive plant reference design that conforms to the EUR and would be licensable in Europe. The EP 1000 closely follows the Westinghouse SPWR design for safety systems and containment and the AP 600 design for auxiliary systems. It also includes features that where required to meet the EUR and key European licensing requirements. The primary circuit of the EP 1000 retains most of the general features of the current-day designs, but some evolutionary features to enhance reliability, simplicity of operation, ease of maintenance, and plant safety have been incorporated into the design. The core, reactor vessel, and reactor internals of the EP 1000 are similar to those of currently operating Westinghouse PWR plants, but several new features are included to enhance the performance characteristics. The basic EP 1000 safety philosophy is based on use of inherent

  1. Westinghouse Small Modular Reactor passive safety system response to postulated events

    International Nuclear Information System (INIS)

    Smith, M. C.; Wright, R. F.

    2012-01-01

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor. This paper is part of a series of four describing the design and safety features of the Westinghouse SMR. This paper focuses in particular upon the passive safety features and the safety system response of the Westinghouse SMR. The Westinghouse SMR design incorporates many features to minimize the effects of, and in some cases eliminates the possibility of postulated accidents. The small size of the reactor and the low power density limits the potential consequences of an accident relative to a large plant. The integral design eliminates large loop piping, which significantly reduces the flow area of postulated loss of coolant accidents (LOCAs). The Westinghouse SMR containment is a high-pressure, compact design that normally operates at a partial vacuum. This facilitates heat removal from the containment during LOCA events. The containment is submerged in water which also aides the heat removal and provides an additional radionuclide filter. The Westinghouse SMR safety system design is passive, is based largely on the passive safety systems used in the AP1000 R reactor, and provides mitigation of all design basis accidents without the need for AC electrical power for a period of seven days. Frequent faults, such as reactivity insertion events and loss of power events, are protected by first shutting down the nuclear reaction by inserting control rods, then providing cold, borated water through a passive, buoyancy-driven flow. Decay heat removal is provided using a layered approach that includes the passive removal of heat by the steam drum and independent passive heat removal system that transfers heat from the primary system to the environment. Less frequent faults such as loss of coolant accidents are mitigated by passive injection of a large quantity of water that is readily available inside containment. An automatic depressurization system is used to

  2. U. S. Utility Leadership in Requirements For Passive Reactors

    International Nuclear Information System (INIS)

    Kim, Jcng H.; Layman, William H.

    1991-01-01

    Utility leadership from both U.S. utilities and international utilities, is a key element in the U. S. Advanced Light Water Reactor Program. International utilities have played a very import Design reviews by the utilities participating in the ALRR Program will ensure that all of the utility requirements are met while design work is being carried out. Our mission is to achieve NRC certification of designs that reflect the needs of the utilities and we believe that this will play an important role in the resurgence of nuclear plant construction in the United States. As stated in the Nuclear Power Oversight Committee's Strategic Plan For Building New Nuclear Power Plants : 'The extensive operating experience with today's light water reactors (LWRs), and the promise shown in recent technical developments, leads the industry to the conclusion that the next nuclear plants ordered in the United States will be advanced light water reactors (A LWRs). Two types are under development : units of large output (1300 MW) called 'evolutionary' A LWRs and units of mid-size output (600 MW) called 'Passive' A LWRs. The term 'passive' refers to the safety features which depend more on natural processes such as gravity and buoyancy than on powered equipment such as pumps

  3. Passive cooling system for nuclear reactor containment structure

    Science.gov (United States)

    Gou, Perng-Fei; Wade, Gentry E.

    1989-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  4. Plant maintenance and advanced reactors, 2006

    Energy Technology Data Exchange (ETDEWEB)

    Agnihotri, Newal (ed.)

    2006-09-15

    The focus of the September-October issue is on plant maintenance and advanced reactors. Major articles/reports in this issue include: Advanced plants to meet rising expectations, by John Cleveland, International Atomic Energy Agency, Vienna; A flexible and economic small reactor, by Mario D. Carelli and Bojan Petrovic, Westinghouse Electric Company; A simple and passively safe reactor, by Yury N. Kuznetsov, Research and Development Institute of Power Engineering (NIKIET), Russia; Gas-cooled reactors, by Jeffrey S. Merrifield, U.S. Nuclear Regulatory Commission; ISI project managment in the PRC, by Chen Chanbing, RINPO, China; and, Fort Calhoun refurbishment, by Sudesh Cambhir, Omaha Public Power District.

  5. Performance of the prism reactor's passive decay heat removal system

    International Nuclear Information System (INIS)

    Magee, P.M.; Hunsbedt, A.

    1989-01-01

    The PRISM modular reactor concept has a totally passive safety-grade decay heat removal system referred to as the Reactor Vessel Auxiliary Cooling System (RVACS) that rejects heat from the reactor by radiation and natural convection of air. The system is inherently reliable and is not subject to the failure modes commonly associated with active cooling systems. The thermal performance of RVACS exceeds requirements and significant thermal margins exist. RVACS has been shown to perform its function under many postulated accident conditions. The PRISM power plant is equipped with three methods for shutdown: condenser cooling in conjunction with intermediate sodium and steam generator systems, and auxiliary cooling system (ACS) which removes heat from the steam generator by natural convection of air and transport of heat from the core by natural convection in the primary and intermediate systems, and a safety- grade reactor vessel auxiliary cooling system (RVACS) which removes heat passively from the reactor containment vessel by natural convection of air. The combination of one active and two passive systems provides a highly reliable and economical shutdown heat removal system. This paper provides a summary of the RVACS thermal performance for expected operating conditions and postulated accident events. The supporting experimental work, which substantiates the performance predictions, is also summarized

  6. Design criteria for the electrical system in advanced passive reactors. Special features of the AP-600 Reactor

    International Nuclear Information System (INIS)

    Moraleda Lopez, A.

    1997-01-01

    The design of the electrical system of an Passive Advanced Reactor is determined by the concept of passive actuation of safety systems, simplification of process systems and optimisation of equipment performance. The system that results from these criteria is very different to those designed for present plants. The main differences are: No class 1E alternating current systems No emergency diesel generators Fewer safety and non-safety class electricity consumers System for continuous monitoring of battery status Use of electronic speed regulators for reactor feedwater pump motors Outsite battery backup safety power supply Motor-operated valves are the only safety electrical actuators Portable power supply for post 72 hour equipment This paper develops these concepts and applies them to the AP-600 project and describes the electrical system of this type of plant. (Author)

  7. NRC review of passive reactor design certification testing programs: Overview, progress, and regulatory perspective

    Energy Technology Data Exchange (ETDEWEB)

    Levin, A.E.

    1995-09-01

    New reactor designs, employing passive safety systems, are currently under development by reactor vendors for certification under the U.S. Nuclear Regulatory Commission`s (NRC`s) design certification rule. The vendors have established testing programs to support the certification of the passive designs, to meet regulatory requirements for demonstration of passive safety system performance. The NRC has, therefore, developed a process for the review of the vendors` testing programs and for incorporation of the results of those reviews into the safety evaluations for the passive plants. This paper discusses progress in the test program reviews, and also addresses unique regulatory aspects of those reviews.

  8. Thermal-hydraulic modeling needs for passive reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, J.M. [Nuclear Regulatory Commission, Washington, DC (United States)

    1997-07-01

    The U.S. Nuclear Regulatory Commission has received an application for design certification from the Westinghouse Electric Corporation for an Advanced Light Water Reactor design known as the AP600. As part of the design certification process, the USNRC uses its thermal-hydraulic system analysis codes to independently audit the vendor calculations. The focus of this effort has been the small break LOCA transients that rely upon the passive safety features of the design to depressurize the primary system sufficiently so that gravity driven injection can provide a stable source for long term cooling. Of course, large break LOCAs have also been considered, but as the involved phenomena do not appear to be appreciably different from those of current plants, they were not discussed in this paper. Although the SBLOCA scenario does not appear to threaten core coolability - indeed, heatup is not even expected to occur - there have been concerns as to the performance of the passive safety systems. For example, the passive systems drive flows with small heads, consequently requiring more precision in the analysis compared to active systems methods for passive plants as compared to current plants with active systems. For the analysis of SBLOCAs and operating transients, the USNRC uses the RELAP5 thermal-hydraulic system analysis code. To assure the applicability of RELAP5 to the analysis of these transients for the AP600 design, a four year long program of code development and assessment has been undertaken.

  9. Thermal-hydraulic modeling needs for passive reactors

    International Nuclear Information System (INIS)

    Kelly, J.M.

    1997-01-01

    The U.S. Nuclear Regulatory Commission has received an application for design certification from the Westinghouse Electric Corporation for an Advanced Light Water Reactor design known as the AP600. As part of the design certification process, the USNRC uses its thermal-hydraulic system analysis codes to independently audit the vendor calculations. The focus of this effort has been the small break LOCA transients that rely upon the passive safety features of the design to depressurize the primary system sufficiently so that gravity driven injection can provide a stable source for long term cooling. Of course, large break LOCAs have also been considered, but as the involved phenomena do not appear to be appreciably different from those of current plants, they were not discussed in this paper. Although the SBLOCA scenario does not appear to threaten core coolability - indeed, heatup is not even expected to occur - there have been concerns as to the performance of the passive safety systems. For example, the passive systems drive flows with small heads, consequently requiring more precision in the analysis compared to active systems methods for passive plants as compared to current plants with active systems. For the analysis of SBLOCAs and operating transients, the USNRC uses the RELAP5 thermal-hydraulic system analysis code. To assure the applicability of RELAP5 to the analysis of these transients for the AP600 design, a four year long program of code development and assessment has been undertaken

  10. Passive and inherent safety technologies for light-water nuclear reactors

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    1990-07-01

    Passive/inherent safety implies a technical revolution in our approach to nuclear power safety. This direction is discussed herein for light-water reactors (LWRs) -- the predominant type of power reactor used in the world today. At Oak Ridge National Laboratory (ORNL) the approach to the development of passive/inherent safety for LWRs consists of four steps: identify and quantify safety requirements and goals; identify and quantify the technical functional requirements needed for safety; identify, invent, develop, and quantify technical options that meet both of the above requirements; and integrate safety systems into designs of economic and reliable nuclear power plants. Significant progress has been achieved in the first three steps of this program. The last step involves primarily the reactor vendors. These activities, as well as related activities worldwide, are described here. 27 refs., 7 tabs

  11. From Dodewaard to a modern economic passive plant-ESBWR

    International Nuclear Information System (INIS)

    Arnold, H.; Yadigaroglu, G.; Stoop, P.M.; Gonzales, A.; Rao, A.S.

    1996-01-01

    For over 25 years the Dodewaard nuclear plant has produced electricity with one of the highest reliability of any nuclear power plant in the world. Almost 10 years ago when some designers looked at the features to incorporate in a modern mid-size Boiling Water Reactor (SBWR) design that was developed by an international team and consisted of natural circulation and passive safety injection and decay heat removal. Since the passive decay heat removal was a major new technology area, an extensive test program was developed and conducted - utilizing newly constructed large scale integrated system test facility at the Paul Scherrer Institute, Switzerland. The development of any modern reactor is time consuming and expensive, hence, the design and technology was done as part of an international team effort. This paper provides an overview of the international design and technology effort and discusses how the development costs were minimized through cooperation. The result of the design effort showed that overall plant economics could be improved by a power level increase, supported by some modest design changes. This paper provides the approach and key results of this multi-year international design effort to develop a modern, larger scale, passive plant - the European Simplified Boiling Water Reactor (ESBWR). This effort, in addition to developing an economic nuclear power station, also demonstrates how the benefits of international cooperation can be utilized to keep development costs reasonable. An overview is also provided of the planned research efforts to extend existing technology to the ESBWR. (authors)

  12. Utility requirements for safety in the passive advanced light-water reactor

    International Nuclear Information System (INIS)

    Marston, T.U.; Layman, W.H.; Bockhold, G. Jr.

    1993-01-01

    The objective of the passive plant design is to use passive systems to replace all the active engineered safety systems presently used in light-water reactors. The benefits derived from such an approach to safety design are multiple. First, it is expected that a passive design approach will significantly simplify the overall plant design, including a reduction in the number of components, and reduce the operation and maintenance burden. Second, it is expected that the overall safety and reliability of the passive systems will be improved over active systems, which will result in extremely low risk to public health and safety. Third, challenges to the operating staff will be minimized during transient and emergency conditions, which will reduce the uncertainty associated with human behavior. Finally, it is expected that reliance on passive safety features will lead to a better understanding by the general public and recognition that a major improvement in public safety has been achieved

  13. Passive cooling of a fixed bed nuclear reactor

    International Nuclear Information System (INIS)

    Petry, V.J.; Bortoli, A.L. de; Sefidwash, F.

    2005-01-01

    Small nuclear reactors without the need for on-site refuelling have greater simplicity, better compliance with passive safety systems, and are more adequate for countries with small electric grids and limited investment capabilities. Here the passive cooling characteristic of the fixed bed nuclear reactor (FBNR), that is being developed under the International Atomic Energy Agency (IAEA) Coordinated Research Project, is studied. A mathematical model is developed to calculate the temperature distribution in the fuel chamber of the reactor. The results demonstrate the passive cooling of this nuclear reactor concept. (authors)

  14. SWR 1000: an advanced boiling water reactor with passive safety features

    International Nuclear Information System (INIS)

    Brettschuh, W.

    1999-01-01

    The SWR 1000, an advanced BWR, is being developed by Siemens under contract from Germany's electric utilities and with the support of European partners. The project is currently in the basic design phase to be concluded in mid-1999 with the release of a site-independent safety report and costing analysis. The development goals for the project encompass competitive costs, use of passive safety systems to further reduce probabilities of occurrence of severe accidents, assured control of accidents so no emergency response actions for evacuation of the local population are needed, simplification of plant systems based on operator experience, and planning and design based on German codes, standards and specifications put forward by the Franco-German Reactor Safety Commission for future nuclear power plants equipped with PWRs, as well as IAEA specifications and the European Utility Requirements. These goals led to a plant concept with a low power density core, with large water inventories stored above the core inside the reactor pressure vessel, in the pressure suppression pool, and in other locations. All accident situations arising from power operation can be controlled by passive safety features without rise in core temperature and with a grace period of more than three days. In addition, postulated core melt is controlled by passive equipment. All new passive systems have been successfully tested for function and performance using large-scale components in experimental testing facilities at PSI in Switzerland and at the Juelich Research Centre in Germany. In addition to improvements of the safety systems, the plant's operating systems have been simplified based on operating experience. The design's safety concept, simplified operating systems and 48 months construction time yield favourable plant construction costs. The level of concept maturity required to begin offering the SWR 1000 on the power generation market is anticipated to be reached, as planned in the year

  15. Design and development of innovative passive valves for Nuclear Power Plant applications

    Energy Technology Data Exchange (ETDEWEB)

    Sapra, M.K., E-mail: sapramk@barc.gov.in; Kundu, S.; Pal, A.K.; Vijayan, P.K.; Vaze, K.K.; Sinha, R.K.

    2015-05-15

    Highlights: • Passive valves are self-acting valves requiring no external energy to function. • These valves have been developed for Advanced Heavy Water Reactor (AHWR) of India. • Passive valves are core components of passive safety systems of the reactor. • Accumulator Isolation Passive Valve (AIPV) has been developed and tested for ECSS. • AIPV provided passive isolation and flow regulation in ECCS of Integral Test Loop. - Abstract: The recent Fukushima accident has resulted in an increased need for passive safety systems in upcoming advanced reactors. In order to enhance the global contribution and acceptability of nuclear energy, proven evidence is required to show that it is not only green but also safe, in case of extreme natural events. To achieve and establish this fact, we need to design, demonstrate and incorporate reliable ‘passive safety systems’ in our advanced reactor designs. In Nuclear Power Plants (NPPs), the use of passive safety systems such as accumulators, condensing and evaporative heat exchangers and gravity driven cooling systems provide enhanced safety and reliability. In addition, they eliminate the huge costs associated with the installation, maintenance and operation of active safety systems that require multiple pumps with independent and redundant electric power supplies. As a result, passive safety systems are preferred for numerous advanced reactor concepts. In current NPPs, passive safety systems which are not participating in day to day operation, are kept isolated, and require a signal and external energy source to open the valve. It is proposed to replace these valves by passive components and devices such as self-acting valves, rupture disks, etc. Some of these innovative passive valves, which do not require external power, have been recently designed, developed and tested at rated conditions. These valves are proposed to be used for various passive safety systems of an upcoming Nuclear Power Plant being designed

  16. Natural circulating passive cooling system for nuclear reactor containment structure

    Science.gov (United States)

    Gou, Perng-Fei; Wade, Gentry E.

    1990-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  17. Study of Cost Effective Large Advanced Pressurized Water Reactors that Employ Passive Safety Features

    International Nuclear Information System (INIS)

    Winters, J.W.; Corletti, M.M.; Hayashi, Y.

    2003-01-01

    A report of DOE sponsored portions of AP1000 Design Certification effort. On December 16, 1999, The United States Nuclear Regulatory Commission issued Design Certification of the AP600 standard nuclear reactor design. This culminated an 8-year review of the AP600 design, safety analysis and probabilistic risk assessment. The AP600 is a 600 MWe reactor that utilizes passive safety features that, once actuated, depend only on natural forces such as gravity and natural circulation to perform all required safety functions. These passive safety systems result in increased plant safety and have also significantly simplified plant systems and equipment, resulting in simplified plant operation and maintenance. The AP600 meets NRC deterministic safety criteria and probabilistic risk criteria with large margins. A summary comparison of key passive safety system design features is provided in Table 1. These key features are discussed due to their importance in affecting the key thermal-hydraulic phenomenon exhibited by the passive safety systems in critical areas. The scope of some of the design changes to the AP600 is described. These changes are the ones that are important in evaluating the passive plant design features embodied in the certified AP600 standard plant design. These design changes are incorporated into the AP1000 standard plant design that Westinghouse is certifying under 10 CFR Part 52. In conclusion, this report describes the results of the representative design certification activities that were partially supported by the Nuclear Energy Research Initiative. These activities are unique to AP1000, but are representative of research activities that must be driven to conclusion to realize successful licensing of the next generation of nuclear power plants in the United States

  18. The Westinghouse Advanced Passive Pressurized Water Reactor, AP1000

    International Nuclear Information System (INIS)

    Schene, R.

    2009-01-01

    Featuring proven technology and innovative passive safety systems, the Westinghouse AP1000 pressurized water reactor can achieve competitive generation costs in the current electricity market without emitting harmful greenhouse gases and further harming the environment. Westinghouse Electric Company, the pioneer in nuclear energy once again sets a new industry standard with the AP1000. The AP1000 is a two-loop pressurized water reactor that uses simplified, innovative and effective approach to safety. With a gross power rating of 3415 megawatt thermal and a nominal net electrical output of 1117 megawatt electric, the AP1000 is ideal for new base load generation. The AP1000 is the safest and most economical nuclear power plant available in the worldwide commercial marketplace, and is the only Generation III+ reactor to receive a design certification from the U.S. Nuclear Regulatory Commission (NRC). Based on nearly 20 years of research and development, the AP1000 builds and improves upon the established technology of major components used in current Westinghouse designed plants. These components, including steam generators, digital instrumentation and controls, fuel, pressurizers, and reactor vessels, are currently in use around the world and have years of proven, reliable operating experience. Historically, Westinghouse plant designs and technology have forged the cutting edge technology of nuclear plant around the world. Today, nearly 50 percent of the world's 440 nuclear plants are based on Westinghouse technology. Westinghouse continues to be the nuclear industry's global leader. (author)

  19. ESBWR passive heat exchanger design and performance - reducing plant development costs

    International Nuclear Information System (INIS)

    Lumini, E.; Upton, H.A.; Billig, P.F.; Masoni, P.

    1996-01-01

    The EUROPEAN Simplified Boiling Water Reactor (ESBWR) is a nuclear plant that builds on the solid technological foundation of the Simplified Boiling Reactor (SBWR) design. The major objective of the ESBWR program is to develop a plant design that utilizes the basic simplicity of the SBWR design that utilizes the basic simplicity of the SBWR design features to improve overall economics and to meet the specific requirements found in the European Utility Requirements Documents (EUR). The design is being developed by an international team of utilities, designers and researchers with the objective of meeting European utility and regulatory requirements. The overall approach to improve the commercial attractiveness of the ESBWR compared to the SBWR was to take advantage of the modular design of the passive safety system, the economy of scale, as well as the advantage of simpler systems of the passive plant to reduce overall material quantities and improve plant economics. To take advantage of the economy of scale, the power level of ESBWR was increased to 1190 MWe. Because of the modular nature of the passive safety systems in SBWR, in increase in thermal power of ESBWR to 3613 MWt only requires that the number of Passive Containment Condensers to maintain the passive safety features of ESBWR to four 33 MWt units for ESBWR. This paper reviews the Passive Containment Cooling (PCC) and Isolation Condenser (IC) unit design and addresses their use in the passive safety systems of the 3613 MWt ESBWR. The specific design differences and the applicability of the test completed at the SIET PANTHERS test facility in Piacenza, Italy are addressed as well as outlining additional qualification tests that must be completed on the PCC and IC unit design if they are to used in the passive safety systems of the ESBWR. This paper outlines the test results obtained from the prototype PCC and IC PANTHERS tests facility in Piacenza, Italy which have been used to design the ESBWR PCC/1C

  20. European passive plant program A design for the 21st century

    International Nuclear Information System (INIS)

    Adomaitis, D.; Oyarzabal, M.

    1998-01-01

    In 1994, a group of European utilities initiated, together with Westinghouse and its industrial partner GENESI (an Italian consortium including ANSALDO and FIAT), a program designated EPP (European Passive Plant) to evaluate Westinghouse passive nuclear plant technology for application in Europe. The following major tasks were accomplished: (1) the impacts of the European utility requirements (EUR) on the Westinghouse nuclear island design were evaluated; and (2) a 1000 MWe passive plant reference design (EP1000) was established which conforms to the EUR and is expected to be licensable in Europe. With respect to safety systems and containment, the reference plant design closely follows that of the Westinghouse simplified pressurized water reactor (SPWR) design, while the AP600 plant design has been taken as the basis for the EP1000 reference design in the auxiliary system design areas. However, the EP1000 design also includes features required to meet the EUR, as well as key European licensing requirements. (orig.)

  1. An evaluation of designed passive Core Makeup Tank (CMT) for China pressurized reactor (CPR1000)

    International Nuclear Information System (INIS)

    Wang, Mingjun; Tian, Wenxi; Qiu, Suizheng; Su, Guanghui; Zhang, Yapei

    2013-01-01

    Highlights: ► Only PRHRS is not sufficient to maintain reactor safety in case of SGTR accident. ► The Core Makeup Tank (CMT) is designed for CPR1000. ► Joint operation of PRHRS and CMT can keep reactor safety during the SGTR transient. ► CMT is a vital supplement for CPR1000 passive safety system design. - Abstract: Emergency Passive Safety System (EPSS) is an innovative design to improve reliability of nuclear power plants. In this work, the EPSS consists of secondary passive residual heat removal system (PRHRS) and the reactor Core Makeup Tank (CMT) system. The PRHRS, which has been studied in our previous paper, can effectively remove the core residual heat and passively improve the inherent safety by passive methods. The designed CMT, representing the safety improvement for CPR1000, is used to inject cool boron-containing water into the primary system during the loss of coolant accident. In this study, the behaviors of EPSS and transient characteristics of the primary loop system during the Steam Generator Tube Rupture (SGTR) accident are investigated using the nuclear reactor thermal hydraulic code RELAP5/MOD3.4. The results show that the designed CMT can protect the reactor primary loop from boiling and maintain primary loop coolant in single phase state. Both PRHRS and CMT operation ensures reactor safety during the SGTR accident. Results reported in this paper show that the designed CMT is a further safety improvement for CPR1000

  2. Issues affecting advanced passive light-water reactor safety analysis

    International Nuclear Information System (INIS)

    Beelman, R.J.; Fletcher, C.D.; Modro, S.M.

    1992-01-01

    Next generation commercial reactor designs emphasize enhanced safety through improved safety system reliability and performance by means of system simplification and reliance on immutable natural forces for system operation. Simulating the performance of these safety systems will be central to analytical safety evaluation of advanced passive reactor designs. Yet the characteristically small driving forces of these safety systems pose challenging computational problems to current thermal-hydraulic systems analysis codes. Additionally, the safety systems generally interact closely with one another, requiring accurate, integrated simulation of the nuclear steam supply system, engineered safeguards and containment. Furthermore, numerical safety analysis of these advanced passive reactor designs wig necessitate simulation of long-duration, slowly-developing transients compared with current reactor designs. The composite effects of small computational inaccuracies on induced system interactions and perturbations over long periods may well lead to predicted results which are significantly different than would otherwise be expected or might actually occur. Comparisons between the engineered safety features of competing US advanced light water reactor designs and analogous present day reactor designs are examined relative to the adequacy of existing thermal-hydraulic safety codes in predicting the mechanisms of passive safety. Areas where existing codes might require modification, extension or assessment relative to passive safety designs are identified. Conclusions concerning the applicability of these codes to advanced passive light water reactor safety analysis are presented

  3. Progress in Methodologies for the Assessment of Passive Safety System Reliability in Advanced Reactors. Results from the Coordinated Research Project on Development of Advanced Methodologies for the Assessment of Passive Safety Systems Performance in Advanced Reactors

    International Nuclear Information System (INIS)

    2014-09-01

    Strong reliance on inherent and passive design features has become a hallmark of many advanced reactor designs, including several evolutionary designs and nearly all advanced small and medium sized reactor (SMR) designs. Advanced nuclear reactor designs incorporate several passive systems in addition to active ones — not only to enhance the operational safety of the reactors but also to eliminate the possibility of serious accidents. Accordingly, the assessment of the reliability of passive safety systems is a crucial issue to be resolved before their extensive use in future nuclear power plants. Several physical parameters affect the performance of a passive safety system, and their values at the time of operation are unknown a priori. The functions of passive systems are based on basic physical laws and thermodynamic principals, and they may not experience the same kind of failures as active systems. Hence, consistent efforts are required to qualify the reliability of passive systems. To support the development of advanced nuclear reactor designs with passive systems, investigations into their reliability using various methodologies are being conducted in several Member States with advanced reactor development programmes. These efforts include reliability methods for passive systems by the French Atomic Energy and Alternative Energies Commission, reliability evaluation of passive safety system by the University of Pisa, Italy, and assessment of passive system reliability by the Bhabha Atomic Research Centre, India. These different approaches seem to demonstrate a consensus on some aspects. However, the developers of the approaches have been unable to agree on the definition of reliability in a passive system. Based on these developments and in order to foster collaboration, the IAEA initiated the Coordinated Research Project (CRP) on Development of Advanced Methodologies for the Assessment of Passive Safety Systems Performance in Advanced Reactors in 2008. The

  4. Operational and passive safety aspects of the STAR-LM natural convection HLMC reactor. Study on operational aspects of a natural circulation HLMC reactor. 2

    International Nuclear Information System (INIS)

    Sienicki, J.J.; Petkov, P.V.

    2001-09-01

    The concept of a heavy liquid metal cooled fast reactor that achieves 100+% natural circulation heat removal from the core has the potential to attain improved cost competitiveness through extreme simplification, proliferation resistance, and heightened passive safety. The concept offers the potential for simplifications in plant control strategies wherein inherent reactor feedbacks may restore balance between energy release and heat removal from the reactor during operation as well as providing passive reactivity shutdown in the event of transients involving failure to scram. This study was initiated to evaluate the operational characteristics of the 100+% natural circulation reactor under normal and transient states using a plant dynamics analysis computer code and to seek design and operational optimization of the concept. In the earlier Phase 1 of the project, the stage for the overall study was prepared. A coupled thermal hydraulics-kinetics plant dynamics analysis code was developed that has the capabilities to calculate operational and accident transients. Code input was prepared for the heavy liquid metal cooled natural circulation reactor concept. A preliminary analysis using the plant dynamics code and its input to calculate three illustrative cases relevant to initial startup, shutdown following long-term operation, and change-in-turbine load demonstrated the capability to analyze typical transient cases. The present second phase of the study involves documentation of the plant dynamics analysis computer code including major assumptions and thermal hydraulic equations as well as application of the code to calculate operational transients and postulated accidents. The following normal and accident scenarios are calculated: initial startup; normal shutdown; startup from hot standby; decrease-in-turbine load; increase-in-turbine load; loss-of-heat sink without scram; overcooling event without scram; and unprotected transient overpower. For the decrease

  5. Comparison of advanced mid-sized reactors regarding passive features, core damage frequencies and core melt retention features

    International Nuclear Information System (INIS)

    Wider, H.

    2005-01-01

    New Light Water Reactors, whose regular safety systems are complemented by passive safety systems, are ready for the market. The special aspect of passive safety features is their actuation and functioning independent of the operator. They add significantly to reduce the core damage frequency (CDF) since the operator continues to play its independent role in actuating the regular safety devices based on modern instrumentation and control (I and C). The latter also has passive features regarding the prevention of accidents. Two reactors with significant passive features that are presently offered on the market are the AP1000 PWR and the SWR 1000 BWR. Their passive features are compared and also their core damage frequencies (CDF). The latter are also compared with those of a VVER-1000. A further discussion about the two passive plants concerns their mitigating features for severe accidents. Regarding core-melt retention both rely on in-vessel cooling of the melt. The new VVER-1000 reactor, on the other hand features a validated ex-vessel concept. (author)

  6. Considerations on nuclear reactor passive safety systems

    International Nuclear Information System (INIS)

    2016-01-01

    After having indicated some passive safety systems present in electronuclear reactors (control bars, safety injection system accumulators, reactor cooling after stoppage, hydrogen recombination systems), this report recalls the main characteristics of passive safety systems, and discusses the main issues associated with the assessment of new passive systems (notably to face a sustained loss of electric supply systems or of cold water source) and research axis to be developed in this respect. More precisely, the report comments the classification of safety passive systems as it is proposed by the IAEA, outlines and comments specific aspects of these systems regarding their operation and performance. The next part discusses the safety approach, the control of performance of safety passive systems, issues related to their reliability, and the expected contribution of R and D (for example: understanding of physical phenomena which have an influence of these systems, capacities of simulation of these phenomena, needs of experimentations to validate simulation codes)

  7. PWR passive plant heat removal assessment: Joint EPRI-CRIEPI advanced LWR studies

    International Nuclear Information System (INIS)

    1991-03-01

    An independent assessment of the capabilities of the PWR passive plant heat removal systems was performed, covering the Passive Residual Heat Removal (PRHR) System, the Passive Safety Injection System (PSIS) and the Passive Containment Cooling System (PCCS) used in a 600 MWe passive plant (e.g., AP600). Additional effort included a review of the test programs which support the design and analysis of the systems, an assessment of the licensability of the plant with regard to heat removal adequacy, and an evaluation of the use of the passive systems with a larger plant. The major conclusions are as follows. The PRHR can remove core decay heat, prevents the pressurizer from filling with water for a loss-of-feedwater transient, and provides safety-grade means for maintaining the reactor coolant system in a safe shutdown condition for the case where the non-safety residual heat removal system becomes unavailable. The PSIS is effective in maintaining the core covered with water for loss-of-coolant accident pipe breaks to eight inches. The PCCS has sufficient heat removal capability to maintain the containment pressure within acceptable limits. The tests performed and planned are adequate to confirm the feasibility of the passive heat removal system designs and to provide a database for verification of the analytical techniques used for the plant evaluations. Each heat removal system can perform in accordance with Regulatory requirements, with the exception that the PRHR system is unable to achieve the required cold shutdown temperature of 200 F within the required 36-hour period. The passive heat removal systems to be used for the 600 MWe plant could be scaled up to a 900 MWe passive plant in a straightforward manner and only minimal, additional confirmatory testing would be required. Sections have been indexed separately for inclusion on the data base

  8. Nuclear desalination in the Arab world - Part II: Advanced inherent and passive safe nuclear reactors

    International Nuclear Information System (INIS)

    Karameldin, A.; Samer S. Mekhemar

    2004-01-01

    Rapid increases in population levels have led to greater demands for fresh water and electricity in the Arab World. Different types of energies are needed to contribute to bridging the gap between increased demand and production. Increased levels of safeguards in nuclear power plants have became reliable due to their large operational experience, which now exceeds 11,000 years of operation. Thus, the nuclear power industry should be attracting greater attention. World electricity production from nuclear power has risen from 1.7% in 1970 to 17%-20% today. This ratio had increased in June 2002 to reach more than 30%, 33% and 42% in Europe, Japan, and South Korea respectively. In the Arab World, both the public acceptance and economic viability of nuclear power as a major source of energy are greatly dependent on the achievement of a high level of safety and environmental protection. An assessment of the recent generation of advanced reactor safety criteria requirements has been carried out. The promising reactor designs adapted for the Arab world and other similar developing countries are those that profit from the enhanced and passive safety features of the new generation of reactors, with a stronger focus on the effective use of intrinsic characteristics, simplified plant design, and easy construction, operation and maintenance. In addition, selected advanced reactors with a full spectrum from small to large capacities, and from evolutionary to radical types, which have inherent and passive safety features, are discussed. The relevant economic assessment of these reactors adapted for water/electricity cogeneration have been carried out and compared with non-nuclear desalination methods. This assessment indicates that, water/electricity cogeneration by the nuclear method with advanced inherent and passive safe nuclear power plants, is viable and competitive. (author)

  9. Approaches to passive safety in advanced thermal reactors

    International Nuclear Information System (INIS)

    Moses, D.L.

    1986-01-01

    Since 1980, there has been a proliferation of thermal reactor designs which incorporate passive safety features. The evolution of this trend is briefly traced, and the nature of various passive safety features is discussed with regard to how they have been incorporated into evolving design concepts. The key aspects of the passive safety features include reduced core power density, enhanced passive heat sinks, inherent assured shutdown mechanisms, elimination/minimization of potential leak paths from the primary coolant systems, enhanced robustness of fuel elements and improved coolant chemistry and component materials. An increased reliance on purely passive safety features typically translates into larger reactor structures at reduced power ratings. Proponents of the most innovative concepts seek to offset the increased costs by simplifying licensing requirements and reducing construction time

  10. Integral nuclear power reactor with natural coolant circulation. Investigation of passive RHR system

    International Nuclear Information System (INIS)

    Samoilov, O.B.; Kuul, V.S.; Malamud, V.A.; Tarasov, G.I.

    1996-01-01

    The development of a small power (up to 240 MWe) integral PWR for nuclear co-generation power plants has been carried out. The distinctive features of this advanced reactor are: primary circuit arrangement in a single pressure vessel; natural coolant circulation; passive safety systems with self-activated control devices; use of a second (guard) vessel housing the reactor; favourable conditions for the most severe accident management. A passive steam condensing channel has been developed which is activated by the direct action of the primary circuit pressure without an automatic controlling action or manual intervention for emergency cooling of an integral reactor with an in-built pressurizer. In an emergency situation as pressure rises in the reactor a self-activated device blows out non-condensable gases from the condenser tube bundle and returns them in the steam-condensing mode of the operation with the returing primary coolant condensate into the reactor. The thermo-physical test facility is constructed and the experimental development of the steam-condensing channels is performed aiming at the verification of mathematical models for these channels operation in integral reactors both at loss-of-heat removal and LOCA accidents. (orig.)

  11. Passive safety systems and natural circulation in water cooled nuclear power plants

    International Nuclear Information System (INIS)

    2009-11-01

    Nuclear power produces 15% of the world's electricity. Many countries are planning to either introduce nuclear energy or expand their nuclear generating capacity. Design organizations are incorporating both proven means and new approaches for reducing the capital costs of their advanced designs. In the future most new nuclear plants will be of evolutionary design, often pursuing economies of scale. In the longer term, innovative designs could help to promote a new era of nuclear power. Since the mid-1980s it has been recognized that the application of passive safety systems (i.e. those whose operation takes advantage of natural forces such as convection and gravity), can contribute to simplification and potentially improve economics of new nuclear power plant designs. The IAEA Conference on The Safety of Nuclear Power: Strategy for the Future, which was convened in 1991, noted that for new plants 'the use of passive safety features is a desirable method of achieving simplification and increasing the reliability of the performance of essential safety functions, and should be used wherever appropriate'. Some new designs also utilize natural circulation as a means to remove core power during normal operation. The use of passive systems can eliminate the costs associated with the installation, maintenance, and operation of active systems that require multiple pumps with independent and redundant electric power supplies. However, considering the weak driving forces of passive systems based on natural circulation, careful design and analysis methods must be employed to ensure that the systems perform their intended functions. To support the development of advanced water cooled reactor designs with passive systems, investigations of natural circulation are conducted in several IAEA Member States with advanced reactor development programmes. To foster international collaboration on the enabling technology of passive systems that utilize natural circulation, the IAEA

  12. From dodewaard to a modern economic passive plant-ESBWR

    International Nuclear Information System (INIS)

    Gonzalez Lopez, A.; Arnold, H.; Yadigaroglu, G.; Stoop, P.M.; Rao, A.

    1997-01-01

    For over 25 years the Dodewaard nuclear plant has produced electricity with one of the highest reliability of any power plant in the world. Almost 10 years ago when some designers looked at the features to incorporate in a modern mid-size BWR design, they chose some of the key features of Dodewaard natural circulation and passive safety. Over the years this design evolved into a 670 SBWR design that was developed by an international team and consisted of natural circulation and passive safety injection and decay heat removal. Since the passive decay heat removal was a major new technology area, an extensive test program was developed and conducted utilizing newly constructed large scale integrated system test facility at the Paul Scherrer Institute, Switzerland. The development of any modern reactor is time consuming and expensive, hence, the design and technology was done as part of an international team effort. This paper provides an overview of the international design and technology effort and discusses how the development costs were minimized through cooperation. (Author)

  13. Experimental and design experience with passive safety features of liquid metal reactors

    International Nuclear Information System (INIS)

    Lucoff, D.M.; Waltar, A.E.; Sackett, J.I.; Salvatores, M.; Aizawa, K.

    1992-10-01

    Liquid metal cooled reactors (LMRs) have already been demonstrated to be robust machines. Many reactor designers now believe that it is possible to include in this technology sufficient passive safety that LMRs would be able to survive loss of flow, loss of heat sink, and transient overpower events, even if the plant protective system fails completely and do so without damage to the core. Early whole-core testing in Rapsodie, EBR-II. and FFTF indicate such designs may be possible. The operational safety testing program in EBR-II is demonstrating benign response of the reactor to a full range of controls failures. But additional testing is needed if transient core structural response under major accident conditions is to be properly understood. The proposed international Phase IIB passive safety tests in FFTF, being designed with a particular emphasis on providing, data to understand core bowing extremes, and further tests planned in EBR-11 with processed IFR fuel should provide a substantial and unique database for validating the computer codes being used to simulate postulated accident conditions

  14. Passive depressurization accident management strategy for boiling water reactors

    International Nuclear Information System (INIS)

    Liu, Maolong; Erkan, Nejdet; Ishiwatari, Yuki; Okamoto, Koji

    2015-01-01

    Highlights: • We proposed two passive depressurization systems for BWR severe accident management. • Sensitivity analysis of the passive depressurization systems with different leakage area. • Passive depressurization strategies can prevent direct containment heating. - Abstract: According to the current severe accident management guidance, operators are required to depressurize the reactor coolant system to prevent or mitigate the effects of direct containment heating using the safety/relief valves. During the course of a severe accident, the pressure boundary might fail prematurely, resulting in a rapid depressurization of the reactor cooling system before the startup of SRV operation. In this study, we demonstrated that a passive depressurization system could be used as a severe accident management tool under the severe accident conditions to depressurize the reactor coolant system and to prevent an additional devastating sequence of events and direct containment heating. The sensitivity analysis performed with SAMPSON code also demonstrated that the passive depressurization system with an optimized leakage area and failure condition is more efficient in managing a severe accident

  15. Passive depressurization accident management strategy for boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Maolong, E-mail: liuml@vis.t.u-tokyo.ac.jp [Department of Nuclear Engineering and Management, School of Engineering, The University of Tokyo (Japan); Erkan, Nejdet [Nuclear Professional School, School of Engineering, The University of Tokyo (Japan); Ishiwatari, Yuki [Department of Nuclear Engineering and Management, School of Engineering, The University of Tokyo (Japan); Hitachi-GE Nuclear Energy, Ltd. (Japan); Okamoto, Koji [Nuclear Professional School, School of Engineering, The University of Tokyo (Japan)

    2015-04-01

    Highlights: • We proposed two passive depressurization systems for BWR severe accident management. • Sensitivity analysis of the passive depressurization systems with different leakage area. • Passive depressurization strategies can prevent direct containment heating. - Abstract: According to the current severe accident management guidance, operators are required to depressurize the reactor coolant system to prevent or mitigate the effects of direct containment heating using the safety/relief valves. During the course of a severe accident, the pressure boundary might fail prematurely, resulting in a rapid depressurization of the reactor cooling system before the startup of SRV operation. In this study, we demonstrated that a passive depressurization system could be used as a severe accident management tool under the severe accident conditions to depressurize the reactor coolant system and to prevent an additional devastating sequence of events and direct containment heating. The sensitivity analysis performed with SAMPSON code also demonstrated that the passive depressurization system with an optimized leakage area and failure condition is more efficient in managing a severe accident.

  16. Studies on plant dynamics of sodium-cooled fast breeder reactors - verification of a plant model

    International Nuclear Information System (INIS)

    Schubert, B.

    1988-01-01

    For the analysis of sodium-cooled FBR safety and dynamics theoretical models are used, which have to be verified. In this report the verification of the plant model SSC-L is conducted by the comparison of calculated data with measurements of the experimental reactors KNK II and RAPSODIE. For this the plant model is extended and adapted. In general only small differences between calculated and measured data are recognized. The results are used to improve and complete the plant model. The extensions of the plant model applicability are used for the calculation of a loss of heat sink transient with reactor scram, considering pipes as passive heat sinks. (orig./HP) With 69 figs., 10 tabs [de

  17. Passive safety features in current and future water cooled reactors

    International Nuclear Information System (INIS)

    1990-11-01

    Better understanding of the passive safety systems and components in current and future water-cooled reactors may enhance the safety of present reactors, to the extend passive features are backfitted. This better understanding should also improve the safety of future reactors, which can incorporate more of these features. Passive safety systems and components may help to prevent accidents, core damage, or release radionuclides to the environment. The Technical Committee Meeting which was hosted by the USSR State Committee for Utilization of Nuclear Energy was attended by about 80 experts from 16 IAEA Member States and the NEA-OECD. A total of 21 papers were presented during the meeting. The objective of the meeting was to review and discuss passive safety systems and features of current and future water cooled reactor designs and to exchange information in this area of activity. A separate abstract was prepared for each of the 21 papers published in this proceedings. Refs, figs and tabs

  18. An integral reactor design concept for a nuclear co-generation plant

    International Nuclear Information System (INIS)

    Lee, D.J.; Kim, J.I.; Kim, K.K.; Chang, M.H.; Moon, K.S.

    1997-01-01

    An integral reactor concept for nuclear cogeneration plant is being developed at KAERI as an attempt to expand the peaceful utilization of well established commercial nuclear technology, and related industrial infrastructure such as desalination technology in Korea. Advanced technologies such as intrinsic and passive safety features are implemented in establishing the design concepts to enhance the safety and performance. Research and development including laboratory-scale tests are concurrently underway to evaluate the characteristics of various passive safety concepts and provide the proper technical data for the conceptual design. This paper describes the preliminary safety and design concepts of the advanced integral reactor. Salient features of the design are hexagonal core geometry, once-through helical steam generator, self-pressurizer, and seismic resistant fine control CEDMS, passive residual heat removal system, steam injector driven passive containment cooling system. (author)

  19. Role of passive valves & devices in poison injection system of advanced heavy water reactor

    International Nuclear Information System (INIS)

    Sapra, M.K.; Kundu, S.; Vijayan, P.K.; Vaze, K.K.; Sinha, R.K.

    2014-01-01

    The Advanced Heavy Water Reactor (AHWR) is a 300 MWe pressure tube type boiling light water (H 2 O) cooled, heavy water (D 2 O) moderated reactor. The reactor design is based on well-proven water reactor technologies and incorporates a number of passive safety features such as natural circulation core cooling; direct in-bundle injection of light water coolant during a Loss of Coolant Accident (LOCA) from Advanced Accumulators and Gravity Driven Water Pool by passive means; Passive Decay Heat Removal using Isolation Condensers, Passive Containment Cooling System and Passive Containment Isolation System. In addition to above, there is another passive safety system named as Passive Poison Injection System (PPIS) which is capable of shutting down the reactor for a prolonged time. It is an additional safety system in AHWR to fulfill the shutdown function in the event of failure of wired shutdown systems i.e. primary and secondary shut down systems of the reactor. When demanded, PPIS injects the liquid poison into the moderator by passive means using passive valves and devices. On increase of main heat transport (MHT) system pressure beyond a predetermined value, a set of rupture disks burst, which in-turn actuate the passive valve. The opening of passive valve initiates inrush of high pressure helium gas into poison tanks to push the poison into the moderator system, thereby shutting down the reactor. This paper primarily deals with design and development of Passive Poison Injection System (PPIS) and its passive valves & devices. Recently, a prototype DN 65 size Poison Injection Passive Valve (PIPV) has been developed for AHWR usage and tested rigorously under simulated conditions. The paper will highlight the role of passive valves & devices in PPIS of AHWR. The design concept and test results of passive valves along with rupture disk performance will also be covered. (author)

  20. Westinghouse Advances in Passive Plant Safety

    International Nuclear Information System (INIS)

    Bruschi, H. J.; Manager, General; Gerstenhaber, E.

    1993-01-01

    On June 26, 1992, Westinghouse submitted the Ap600 Standard Safety Analysis Report and comprehensive PIRA results to the U. S. NRC for review as part of the Ap600 design certification program. This major milestone was met on time on a schedule set more than 3 years before submittal and is the result of the cooperative efforts of the U. S. Department of Energy (DOE), the Electric Power Requirements Program, and the Westinghouse Ap600 design team. These efforts were initiated in 1985 to develop a 600 MW advanced light water reactor plant design based on specific technical requirements established to provide the safety, simplicity, reliability, and economics necessary for the next generation of nuclear power plants. The Ap600 design achieves the ALRR safety requirements through ample design margins, simplified safety systems based on natural driving forces, and on a human-engineered man-machine interface system. Extensive Probabilistic Risk evolution, have recently shown that even if none of the active defense-in-depth safety systems are available, the passive systems alone meet safety goals. Furthermore, many tests in an extensive test program have begun or have been completed. Early tests show that passive safety perform well and meet design expectations

  1. Utility requirements for advanced LWR passive plants

    International Nuclear Information System (INIS)

    Yedidia, J.M.; Sugnet, W.R.

    1992-01-01

    LWR Passive Plants are becoming an increasingly attractive and prominent option for future electric generating capacity for U.S. utilities. Conceptual designs for ALWR Passive Plants are currently being developed by U.S. suppliers. EPRI-sponsored work beginning in 1985 developed preliminary conceptual designs for a passive BWR and PWR. DOE-sponsored work from 1986 to the present in conjunction with further EPRI-sponsored studies has continued this development to the point of mature conceptual designs. The success to date in developing the ALWR Passive Plant concepts has substantially increased utility interest. The EPRI ALWR Program has responded by augmenting its initial scope to develop a Utility Requirements Document for ALWR Passive Plants. These requirements will be largely based on the ALWR Utility Requirements Document for Evolutionary Plants, but with significant changes in areas related to the passive safety functions and system configurations. This work was begun in late 1988, and the thirteen-chapter Passive Plant Utility Requirements Document will be completed in 1990. This paper discusses the progress to date in developing the Passive Plant requirements, reviews the top-level requirements, and discusses key issues related to adaptation of the utility requirements to passive safety functions and system configurations. (orig.)

  2. Study of passive residual heat removal system of a modular small PWR reactor

    International Nuclear Information System (INIS)

    Araujo, Nathália N.; Su, Jian

    2017-01-01

    This paper presents a study on the passive residual heat removal system (PRHRS) of a small modular nuclear reactor (SMR) of 75MW. More advanced nuclear reactors, such as generation III + and IV, have passive safety systems that automatically go into action in order to prevent accidents. The purpose of the PRHRS is to transfer the decay heat from the reactor's nuclear fuel, keeping the core cooled after the plant has shut down. It starts operating in the event of fall of power supply to the nuclear station, or in the event of an unavailability of the steam generator water supply system. Removal of decay heat from the core of the reactor is accomplished by the flow of the primary refrigerant by natural circulation through heat exchangers located in a pool filled with water located above the core. The natural circulation is caused by the density gradient between the reactor core and the pool. A thermal and comparative analysis of the PRHRS was performed consisting of the resolution of the mass conservation equations, amount of movement and energy and using incompressible fluid approximations with the Boussinesq approximation. Calculations were performed with the aid of Mathematica software. A design of the heat exchanger and the cooling water tank was done so that the core of the reactor remained cooled for 72 hours using only the PRHRS

  3. Advanced passive technology: A global standard for nuclear plant requirements

    International Nuclear Information System (INIS)

    Novak, V.

    1994-01-01

    Since 1984, Westinghouse has been developing AP8OO, a 800 MW, two-loop advanced passive plant, in response to an initiative established by the Electric Power Research Institute (EPRI) and the U.S. Department of Energy' (DOE). The preliminary design was cornpleved in 1989. AP6OO's Standard Safety Analysis and Probabilistic Risk analysis Reports were submitted to the U.S. Nuclear Regulatory Commission for design certification in 1992. Design simplification is the key strategy behind the AP6OO. The basic technical concept Of simplification has resulted in a simplified reactor coolant systems, simplified plant systems, a simplified plant arrangement, reduced number of components, simplified operation and maintenance

  4. Advanced passive technology: A global standard for nuclear plant requirements

    Energy Technology Data Exchange (ETDEWEB)

    Novak, V

    1994-12-31

    Since 1984, Westinghouse has been developing AP8OO, a 800 MW, two-loop advanced passive plant, in response to an initiative established by the Electric Power Research Institute (EPRI) and the U.S. Department of Energy` (DOE). The preliminary design was cornpleved in 1989. AP6OO`s Standard Safety Analysis and Probabilistic Risk analysis Reports were submitted to the U.S. Nuclear Regulatory Commission for design certification in 1992. Design simplification is the key strategy behind the AP6OO. The basic technical concept Of simplification has resulted in a simplified reactor coolant systems, simplified plant systems, a simplified plant arrangement, reduced number of components, simplified operation and maintenance.

  5. Applied reliability assessment for the passive safety systems of nuclear power plants (NPPs) using system dynamics (SD)

    International Nuclear Information System (INIS)

    Kim, Yun Il; Woo, Tae Ho

    2018-01-01

    The passive system by the free-fall is investigated in the accident of nuclear power plants (NPPs). The complex algorithm of the system dynamics (SD) modeling is done in the passive cooling system. The nuclear passive system by free-fall is successfully modeled for the loss of coolant accident (LOCA). Conventional passive system of gravity or natural circulation is working only when the piping systems is in the good condition. The external coolant supply system is introduced in the case of the piping system failure. The water is poured into the reactor through the guiding piping or tube. If the explosion happens, the coolants could be showering into the reactor core and its building. New kind of passive system is expected successfully in the on-site black out where the drone could be operated by battery or engine.

  6. Thermal hydraulic studies for passive heat transport systems relevant to advanced reactors

    International Nuclear Information System (INIS)

    Vijayan, P.K.; Sharma, M.; Borgohain, A.; Srivastava, A.K.; Pilkhwal, D.S.; Maheshwari, N.K.

    2014-01-01

    Nuclear is the only non-green house gas generating power source that can replace fossil fuels and can be commercially deployed in large scale. However, the enormous developmental efforts and safety upgrades during the past six decades have somewhat eroded the economic competitiveness of water-cooled reactors which form the mainstay of the current nuclear power programme. Further, the introduction of the supercritical Rankine cycle and the gas turbine based advanced fuel cycles have enhanced the efficiency of fossil fired power plants (FPP) thereby reducing its greenhouse gas emissions. The ongoing development of ultra-supercritical and advanced ultra-supercritical turbines aims to further reduce the greenhouse gas emissions and economic competitiveness of FPPs. In the backdrop of these developments, the nuclear industry also initiated development of advanced nuclear power plants (NPP) with improved efficiency, sustainability and enhanced safety as the main goals. A review of the advanced reactor concepts being investigated currently reveals that excepting the SCWR, all other concepts use coolants other than water. The coolants used are lead, lead bismuth eutectic, liquid sodium, molten salts, helium and supercritical water. Besides, some of these are employing passive systems to transport heat from the core under normal operating conditions. In view of this, a study is in progress at BARC to examine the performance of simple passive systems using SC CO 2 , SCW, LBE and molten salts as the coolant. This paper deals with some of the recent results of these studies. The study focuses on the steady state, transient and stability behaviour of the passive systems with these coolants. (author)

  7. Advanced Reactor Passive System Reliability Demonstration Analysis for an External Event

    Directory of Open Access Journals (Sweden)

    Matthew Bucknor

    2017-03-01

    Full Text Available Many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general for the postulated transient event.

  8. Advanced reactor passive system reliability demonstration analysis for an external event

    Energy Technology Data Exchange (ETDEWEB)

    Bucknor, Matthew; Grabaskas, David; Brunett, Acacia J.; Grelle, Austin [Argonne National Laboratory, Argonne (United States)

    2017-03-15

    Many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general) for the postulated transient event.

  9. Advanced reactor passive system reliability demonstration analysis for an external event

    International Nuclear Information System (INIS)

    Bucknor, Matthew; Grabaskas, David; Brunett, Acacia J.; Grelle, Austin

    2017-01-01

    Many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general) for the postulated transient event

  10. Fusion reactor passive safety and ignitor risk-based regulation

    International Nuclear Information System (INIS)

    Zucchetti, M.

    1995-01-01

    Passive design features are more reliable than operator action of successful operation of active safety systems. Passive safety has usually been adopted for fission. The achievement of an inventory-based passive safety is difficult if the fusion reactor uses neutronic reactions. Ignitor is a high-magnetic field tokamak designed to study the physics of ignited plasmas. The safety goal for Ignitor is classification as a mobility-based passively safe machine

  11. Passive safety and the advanced liquid metal reactors

    International Nuclear Information System (INIS)

    Hill, D.J.; Pedersen, D.R.; Marchaterre, J.F.

    1988-01-01

    Advanced Liquid Metal Reactors being developed today in the USA are designed to make maximum use of passive safety features. Much of the LMR safety work at Argonne National Laboratory is concerned with demonstrating, both theoretically and experimentally, the effectiveness of the passive safety features. The characteristics that contribute to passive safety are discussed, with particular emphasis on decay heat removal systems, together with examples of Argonne's theoretical and experimental programs in this area

  12. Worldwide advanced nuclear power reactors with passive and inherent safety: What, why, how, and who

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Reich, W.J.

    1991-09-01

    The political controversy over nuclear power, the accidents at Three Mile Island (TMI) and Chernobyl, international competition, concerns about the carbon dioxide greenhouse effect and technical breakthroughs have resulted in a segment of the nuclear industry examining power reactor concepts with PRIME safety characteristics. PRIME is an acronym for Passive safety, Resilience, Inherent safety, Malevolence resistance, and Extended time after initiation of an accident for external help. The basic ideal of PRIME is to develop power reactors in which operator error, internal sabotage, or external assault do not cause a significant release of radioactivity to the environment. Several PRIME reactor concepts are being considered. In each case, an existing, proven power reactor technology is combined with radical innovations in selected plant components and in the safety philosophy. The Process Inherent Ultimate Safety (PIUS) reactor is a modified pressurized-water reactor, the Modular High Temperature Gas-Cooled Reactor (MHTGR) is a modified gas-cooled reactor, and the Advanced CANDU Project is a modified heavy-water reactor. In addition to the reactor concepts, there is parallel work on super containments. The objective is the development of a passive ''box'' that can contain radioactivity in the event of any type of accident. This report briefly examines: why a segment of the nuclear power community is taking this new direction, how it differs from earlier directions, and what technical options are being considered. A more detailed description of which countries and reactor vendors have undertaken activities follows. 41 refs

  13. Passive residual energy utilization system in thermal cycles on water-cooled power reactors

    International Nuclear Information System (INIS)

    Placco, Guilherme M.; Guimaraes, Lamartine N.F.; Santos, Rubens S. dos

    2013-01-01

    This work presents a concept of a residual energy utilization in nuclear plants thermal cycles. After taking notice of the causes of the Fukushima nuclear plant accident, an idea arose to adapt a passive thermal circuit as part of the ECCS (Emergency Core Cooling System). One of the research topics of IEAv (Institute for Advanced Studies), as part of the heat conversion of a space nuclear power system is a passive multi fluid turbine. One of the main characteristics of this device is its passive capability of staying inert and be brought to power at moments notice. During the first experiments and testing of this passive device, it became clear that any small amount of gas flow would generate power. Given that in the first stages of the Fukushima accident and even during the whole event there was plenty availability of steam flow that would be the proper condition to make the proposed system to work. This system starts in case of failure of the ECCS, including loss of site power, loss of diesel generators and loss of the battery power. This system does not requires electricity to run and will work with bleed steam. It will generate enough power to supply the plant safety system avoiding overheating of the reactor core produced by the decay heat. This passive system uses a modified Tesla type turbine. With the tests conducted until now, it is possible to ensure that the operation of this new turbine in a thermal cycle is very satisfactory and it performs as expected. (author)

  14. Passive cooling systems in power reactors

    International Nuclear Information System (INIS)

    Aharon, J.; Harrari, R.; Weiss, Y.; Barnea, Y.; Katz, M.; Szanto, M.

    1996-01-01

    This paper reviews several R and D activities associated with the subject of passive cooling systems, conducted by the N.R.C.Negev thermohydraulic group. A short introduction considering different types of thermosyphons and their applications is followed by a detailed description of the experimental work, its results and conclusions. An ongoing research project is focused on the evaluation of the external dry air passive containment cooling system (PCCS) in the AP-600 (Westinghouse advanced pressurized water reactor). In this context some preliminary theoretical results and planned experimental research are for the fature described

  15. NRC review of Electric Power Research Institute's advanced light water reactor utility requirements document. Passive plant designs, chapter 1, project number 669

    International Nuclear Information System (INIS)

    1994-08-01

    The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the open-quotes Advanced Light Water Reactor [ALWR] Utility Requirements Documentclose quotes, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume 1, open-quotes ALWR Policy and Summary of Top-Tier Requirementsclose quotes, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, open-quotes NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document - Program Summaryclose quotes, provided a discussion of the overall purpose and scope of the Requirements Document, the background of the staff's review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review

  16. Development of the containment transient analysis code for the passive reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Young Dong; Kim, Young In; Bae, Yoon Young; Chang, Moon Hi [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-05-01

    This study was performed to develop the analysis tools for the passively cooled steel containment and to construct the integrated code system which can analyze a thermal hydraulic behavior of the containment and reactor system during a loss of coolant accident. The computer code CONTEMPT4/MOD5/PCCS was developed by incorporating the passive containment cooling models to the containment pressure and temperature transient analysis computer code CONTEMPT4/MOD5. The integrated reactor thermal hydraulic analysis code system for passive reactor was constructed by coupling the best estimate thermal hydraulic system analysis code RELAP5/MOD3 and CONTEMPT4/MOD5/PCCS through the process control method. In addition, to evaluate the applicability of the code the CONTEMPT4/MOD5/PCCS was applied to the SMART(System-Integrated Modular Advanced Reactor). The pressure and temperature transient following the small break LOCA of SMART was analysed by modeling the safeguard vessel using both the newly added passive containment cooling model and existing pool model. (author). 16 refs., 22 figs., 7 tabs.

  17. The Passive Reactor SIRTM - Developments in the UK

    International Nuclear Information System (INIS)

    Hayns, M. R.

    1991-01-01

    We have briefly described the circumstances in the UK which lead to our interest in next generation light water reactors. Whilst some of these issues are today parochial to the UK, we believe that many of the elements of the design are more widely applicable and that it offers a radical, but realistic plant. We believe that the use of passive system and inherent safety features in this design offer a balance between expectation and realism. Thus, there are no unique or untried systems and all of the major components are drawn from existing technology. The final test of realism, of course, is a submission to a licensing authority. We are not at that stage yet, but through simplification and use of tried systems, we believe that licensability should not be an issue. Whilst all cost calculations for nuclear power plants are fraught with difficulty, we believe that be using established methods and subculture date we have provided as good an estimate of costs as is possible at this stage of a new design. Even allowing a margin for error, it is clear that, at least for UK and US conditions, the STR TM design is competitive. For the single unit plant at 400 MW capital costs slightly exceed those of large plant. However, the real advantages of this size of plant only become apparent for a run of plant. Then not only cost, but availability, flexibility and financial risk factors all weigh heavily for the smaller plant. of course at some point, the larger plant will overtake the smaller plant as it too benefits from series ordering. However, a utility or country would need to order a series of like designs of many tens of megawatts before this would happen. We believe that for most countries or utilities, cost advantages for generating capacity in the range 5-10 GW is a more realistic target. At the present it is not clear whether the STR TM design will progress beyond its present state. Even if it does not, we believe that the exercise has provided many useful lessons, and

  18. Pioneering SUPER - Small Unit Passively-safe Enclosed Reactor - 15559

    International Nuclear Information System (INIS)

    Bhownik, P.K.; Gairola, A.; Shamim, J.A.; Suh, K.Y.; Suh, K.S.

    2015-01-01

    This paper presents the basic features of the Small Unit Passively-safe Enclosed Reactor abbreviated as SUPER, a new reactor system that has been designed and proposed at the Seoul National University's Department of Energy Systems Engineering. SUPER is a small modular reactor system or SMR that is cooled by sub-cooled as well as supercritical water. As a new member of SMRs, SUPER is a small-scale nuclear plant that is designed to be factory-manufactured and shipped as modules to be assembled at a site. The concept offers promising answers to many questions about nuclear power including proliferation resistance, waste management, safety, and startup costs. SUPER is a customized paradigm of a supercritical water reactor or SCWR, a type sharing commonalities with the current fleet of light water reactors, or LWRs. SUPER has evolved from the System-integrated Modular Advance Reactor, or SMART, being developed at the Korea Atomic Energy Research Institute, or KAERI. SUPER enhanced the safety features for robustness, design/equipment simplification for natural convection, multi-purpose application for co-generation flexibilities, suitable for isolated or small electrical grids, just-in-time capacity addition, short construction time, and last, but not least, lower capital cost per unit. The primary objectives of SUPER is to develop the conceptual design for a safe and economic small, natural circulation SCWR, to address the economic and safety attributes of the concept, and to demonstrate its technical feasibilities. (authors)

  19. Passive Safety Systems in Advanced Water Cooled Reactors (AWCRS). Case Studies. A Report of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)

    International Nuclear Information System (INIS)

    2013-09-01

    This report presents the results from the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) collaborative project (CP) on Advanced Water Cooled Reactor Case Studies in Support of Passive Safety Systems (AWCR), undertaken under the INPRO Programme Area C. INPRO was launched in 2000 - on the basis of a resolution of the IAEA General Conference (GC(44)/RES/21) - to ensure that nuclear energy is available in the 21st century in a sustainable manner, and it seeks to bring together all interested Member States to consider actions to achieve innovation. An important objective of nuclear energy system assessments is to identify 'gaps' in the various technologies and corresponding research and development (R and D) needs. This programme area fosters collaboration among INPRO Member States on selected innovative nuclear technologies to bridge technology gaps. Public concern about nuclear reactor safety has increased after the Fukushima Daiichi nuclear power plant accident caused by the loss of power to pump water for removing residual heat in the core. As a consequence, there has been an increasing interest in designing safety systems for new and advanced reactors that are passive in nature. Compared to active systems, passive safety features do not require operator intervention, active controls, or an external energy source. Passive systems rely only on physical phenomena such as natural circulation, thermal convection, gravity and self-pressurization. Passive safety features, therefore, are increasingly recognized as an essential component of the next-generation advanced reactors. A high level of safety and improved competitiveness are common goals for designing advanced nuclear power plants. Many of these systems incorporate several passive design concepts aimed at improving safety and reliability. The advantages of passive safety systems include simplicity, and avoidance of human intervention, external power or signals. For these reasons, most

  20. Seawater desalination plant using nuclear heating reactor coupled with MED process

    Institute of Scientific and Technical Information of China (English)

    2000-01-01

    A small size plant for seawater desalination using nuclear heating reactor coupled with MED process was developed by the Institute of Nuclear Energy Technology, Tsinghua University, China. This seawater desalination plant was designed to supply potable water demand to some coastal location or island where both fresh water and energy source are severely lacking. It is also recommended as a demonstration and training facility for seawater desalination using nuclear energy. The design of small size of seawater desalination plant couples two proven technologies: Nuclear Heating Reactor (NHR) and Multi-Effect Destination (MED) process. The NHR design possesses intrinsic and passive safety features, which was demonstrated by the experiences of the project NHR-5. The intermediate circuit and steam circuit were designed as the safety barriers between the NHR reactor and MED desalination system. Within 10~200 MWt of the power range of the heating reactor, the desalination plant could provide 8000 to 150,000 m3/d of high quality potable water. The design concept and parameters, safety features and coupling scheme are presented.

  1. Seawater desalination plant using nuclear heating reactor coupled with MED process

    International Nuclear Information System (INIS)

    Wu Shaorong; Dong Duo; Zhang Dafang; Wang Xiuzhen

    2000-01-01

    A small size plant for seawater desalination using nuclear heating reactor coupled with MED process was developed by the Institute of Nuclear Energy Technology, Tsinghua University, China. this seawater desalination plant was designed to supply potable water demand to some coastal location or island where both fresh water and energy source are severely lacking. It is also recommended as a demonstration and training facility for seawater desalination using nuclear energy. The design of small size of seawater desalination plant couples two proven technologies: Nuclear Heating Reactor (NHR) and Multi-Effect Destination (MED) process. The NHR design possesses intrinsic and passive safety features, which was demonstrated by the experiences of the project NHR-5. the intermediate circuit and steam circuit were designed as the safety barriers between the NHR reactor and MED desalination system. Within 10-200 MWt of the power range of the heating reactor, the desalination plant could provide 8000 to 150,000 m 3 /d of high quality potable water. The design concept and parameters, safety features and coupling scheme are presented

  2. Thermal analysis and design of a passive reflux condenser for the simplified boiling water reactor

    International Nuclear Information System (INIS)

    Bijlani, C.; Patti, F.; Prasad, V.

    1993-01-01

    At present, the advanced light water reactors (ALWRS) in the United States are being designed to remove reactor decay heat for a period of 72 h following a postulated loss-of-coolant accident (LOCA). The water in the pools external to the containment is evaporated or boiled off to remove the decay heat. It is presumed that the water in the pools can be replenished within 72 h through operator actions or outside assistance. Some countries in Europe require that the plant be designed to remove the reactor decay heat for a much longer duration than 72 h without external assistance. This paper presents an analysis and design of a passive heat exchanger called a reflux condenser (RC), which was considered for an ALWR-the 600-MW(electric) simplified boiling water reactor. The RC is required to condense the steam formed when the water in the pool in which the passive containment cooling system (PCCS) is immersed boils following a LOCA. The RCs are nuclear non-safety related. This paper presents steady-state performance of an RC at various outdoor air dry-bulb temperatures under still air conditions

  3. Advanced passive PWR AC-600: Development orientation of nuclear power reactors in China for the next century

    International Nuclear Information System (INIS)

    Huang Xueqing; Zhang Senru

    1999-01-01

    Based on Qinshan II Nuclear Power Plant that is designed and constructed by way of self-reliance, China has developed advanced passive PWR AC-600. The design concept of AC-600 not only takes the real situation of China into consideration, but also follows the developing trend of nuclear power in the world. The design of AC-600 has the following technical characteristics: Advanced reactor: 18-24 month fuel cycle, low neutron leakage, low power density of the core, no any penetration in the RPV below the level of the reactor coolant nozzles; Passive safety systems: passive emergency residual heat removal system, passive-active safety injection system, passive containment cooling system and main control room habitability system; System simplified and the number of components reduced; Digital I and C; Modular construction. AC-600 inherits the proven technology China has mastered and used in Qirtshan 11, and absorbs advanced international design concepts, but it also has a distinctive characteristic of bringing forth new ideas independently. It is suited to Chinese conditions and therefore is expected to become an orientation of nuclear power development by self-reliance in China for the next century. (author)

  4. Application in nuclear engineering: methodology of innovative nuclear reactors: approaches to the safety of future nuclear power plants

    International Nuclear Information System (INIS)

    Alramady, A.M.K

    2008-01-01

    This thesis describes RELAP5 and MATLAB/SIMULINK computer codes for thermal hydraulic analysis of a typical pressurized water reactor (PWR). The two codes are used to calculate the thermal-hydraulic characteristics of the reactor core and the primary loop under steady-state and hypothetical accidents conditions.New designs of nuclear power plants are directed to increase safety by many methods like reducing the dependence on active parts (such as safety pumps, fans, and diesel generators ) and replacing them with passive features such as gravity draining of cooling water from tanks, and natural circulation of water and air. In this work, high and medium pressure injection pumps are replaced by passive injection components. Different break sizes in cold leg pipe are simulated to analyze to what degree the plant is safe (without any operator action) by using only these passive components. The passive design means operators would not need to take immediate action after an accident, with the reactor ,instead, safely shutting down on its own. Different accident scenarios were simulated in this thesis as loss of coolant accidents and station blackout accidents, and complete passive safety systems used to mitigate theses accidents.

  5. NRC review of Electric Power Research Institute's advanced light water reactor utility requirements document. Passive plant designs, chapters 2-13, project number 669

    International Nuclear Information System (INIS)

    1994-08-01

    The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the open-quotes Advanced Light Water Reactor [ALWR] Utility Requirements Documentclose quotes, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume I, open-quotes ALWR Policy and Summary of Top-Tier Requirementsclose quotes, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, open-quotes NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document - Program Summaryclose quotes, provided a discussion of the overall purpose and scope of the Requirements Document, the background of the staff's review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review

  6. The importance of carry out studies about the use of passive autocatalytic recombiners for hydrogen control in reactors type ESBWR

    International Nuclear Information System (INIS)

    Sanchez J, J.; Morales S, J. B.

    2009-10-01

    A way to satisfy and to guarantee the energy necessities in the future is increasing in a gradual way the creation of nuclear power plants, introducing advanced designs in its systems that contribute in way substantial in the security of the same nuclear plants. The tendency of new designs of these nuclear plants is the incorporation of systems more reliable and sure, and that the operation does not depend on external factors as the electric power, motors diesel or the action of the operator of nuclear plant, what is known as security passive systems. In this sense, the passive autocatalytic recombiners are a contribution toward the use of this type of systems. At the present time it is had studies of the incorporation of passive autocatalytic recombiners in nuclear plants in operation and that they have contributed to minimize the danger associated to hydrogen. The present work contains a first approach to the study of hydrogen recombiners incorporation in advanced nuclear plants, for this case in a nuclear power plant of ESBWR type. To achieve our objective it seeks to use specialized codes as RELAP/SCDAP to obtain simulations of passive autocatalytic recombiners behaviour and we can to estimate their operation inside the reactor contention, contemplating the possibility to use other codes like SCILAB and/or MATLAB for the simulation of a passive autocatalytic recombiner. (Author)

  7. Development of passive condensers for accident localization systems at nuclear power plants in the former USSR

    International Nuclear Information System (INIS)

    Kuznecov, M.V.

    1992-01-01

    The development is summarized of passive condensers for accident localization systems at nuclear power plants (with RBMK and WWER reactors) in the former USSR. Basic properties and criteria defining their availability are described, as are experimental tests and technical solution optimization results. (author) 2 fig

  8. A study of passive safety conditions for fast reactor core

    International Nuclear Information System (INIS)

    Shimizu, Akinao

    1991-01-01

    A study has been made for passive safety conditions of fast reactor cores. Objective of the study is to develop a concept of a core with passive safety as well as a simple safety philosophy. A simple safety philosophy, which is wore easy to explain to the public, is needed to enhance the public acceptance for nuclear reactors. The present paper describes a conceptual plan of the study including the definition of the problem a method of approach and identification of tasks to be solved

  9. Feasibility study of applying the passive safety system concept to fusion–fission hybrid reactor

    International Nuclear Information System (INIS)

    Yu, Zhang-cheng; Xie, Heng

    2014-01-01

    The fusion–fission hybrid reactor can produce energy, breed nuclear fuel, and handle the nuclear waste, etc., with the fusion neutron source striking the subcritical blanket. The passive safety system consists of passive residual heat removal system, passive safety injection system and automatic depressurization system was adopted into the fusion–fission hybrid reactor in this paper. Modeling and nodalization of primary loop, partial secondary loop and passive core cooling system for the fusion–fission hybrid reactor using relap5 were conducted and small break LOCA on cold leg was analyzed. The results of key transient parameters indicated that the actuation of passive safety system could mitigate the accidental consequence of the 4-inch cold leg small break LOCA on cold leg in the early time effectively. It is feasible to apply the passive safety system concept to fusion–fission hybrid reactor. The minimum collapsed liquid level had great increase if doubling the volume of CMTs to increase its coolant injection and had no increase if doubling the volume of ACCs

  10. Concept of passive safe small reactor for distributed energy supply system

    International Nuclear Information System (INIS)

    Ishida, Toshihisa; Nakajima, Nobuya; Sawada, Ken-ichi; Yoritsune, Tsutomu; Shimada, Shoichiro; Nakano, Yoshihiro; Yonomoto, Taisuke; Takahashi, Hiroki

    2003-01-01

    This paper presents a concept of a Passive Safe Small Reactor for Distributed energy supply system (PSRD). The PSRD is an integrated-type PWR with reactor thermal power of 100 to 300 MW aimed at supplying electricity, district heating, etc. In design of the PSRD, high priority is laid on enhancement of safety as well as improvement of economy. Safety is enhanced by the following means: i) Extreme reduction of pipes penetrating the reactor vessel, by limiting to only those of the steam, the feed water and the safety valves, ii) Adoption of the water filled containment and the passive safety systems with fluid driven by natural circulation force, and iii) Adoption of the in-vessel type control rod drive mechanism, accompanying a passive reactor shut-down. To comply with a severe operation condition of PSRD, material of the ball bearing with graphite retainer has been selected by test. For improvement of economy, simplification of the reactor system and long operation of the core are achieved. Optimization of core design concerning the burnable poison ensures the burn-up of 28 GWd/t for low enriched UO 2 fuel rods. (author)

  11. Safety significance of ATR [Advanced Test Reactor] passive safety response attributes

    International Nuclear Information System (INIS)

    Atkinson, S.A.

    1989-01-01

    The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory was designed with some passive safety response attributes which contribute to the safety posture of the facility. The three passive safety attributes being evaluated in the paper are: (1) In-core and in-vessel natural convection cooling, (2) a passive heat sink capability of the ATR primary coolant system (PCS) for the transfer of decay power from the uninsulated piping to the confinement, and (3) gravity feed of emergency coolant makeup. The safety significance of the ATR passive safety response attributes is that the reactor can passively respond for most transients, given a reactor scram, to provide adequate decay power removal and a significant time for operator action should the normal active heat removal systems and their backup systems both fail. The ATR Interim Level 1 Probabilistic Risk Assessment (PRA) model ands results were used to evaluate the significance to ATR fuel damage frequency (or probability) of the above three passive response attributes. The results of the evaluation indicate that the first attribute is a major safety characteristic of the ATR. The second attribute has a noticeable but only minor safety significance. The third attribute has no significant influence on the ATR Level 1 PRA because of the diversity and redundancy of the ATR firewater injection system (emergency coolant system). 8 refs., 4 figs., 1 tab

  12. A passive automated personnel accountability system for reactor emergency preparedness

    International Nuclear Information System (INIS)

    Zimmerman, R.O.; DeLisle, G.V.; Hickey, E.E.

    1988-04-01

    In 1985 a project was undertaken at the N Reactor on the Hanford Site to develop an automated personnel accountability system to ensure accountability of all personnel within 30 minutes of a site evacuation. The decision to develop such a system was made after a full-scale evacuation drill showed that the manual accountability system in use at the time was inadequate to meet the 30-minute requirement. Accountability systems at commercial nuclear power plants were evaluated, but found to be unsuitable because they were not passive, that is, they required action on part of the user for the system to work. Approximately 2500 people could be required to evacuate the 100-N Area. Therefore, a card key system or badge exchange system was judged not to be feasible. A passive accountability system was desired for N Reactor to allow personnel to enter and leave the site in a more timely manner. To meet the need for an automated accountability system at N Reactor, a special Evacuation Accountability System (EVACS) was designed and developed. The EVACS system has three basic components: the transponder, a credit card-sized device worn with the security badge; portal monitors, which are electronically activated by the transponder; and a computer information system that contains the personnel data base. Each person wearing a transponder is accounted for automatically by walking through a portal. In this paper, a description of the hardware and software will be presented, together with problems encountered and lessons learned while adapting an existing technology to this particular use. The system is currently installed and requires acceptance testing before becoming operational

  13. An approach for assessing ALWR passive safety system reliability

    International Nuclear Information System (INIS)

    Hake, T.M.

    1991-01-01

    Many of the advanced light water reactor (ALWR) concepts proposed for the next generation of nuclear power plants rely on passive rather than active systems to perform safety functions. Despite the reduced redundancy of the passive systems as compared to active systems in current plants, the assertion is that the overall safety of the plant is enhanced due to the much higher expected reliability of the passive systems. In order to investigate this assertion, a study is being conducted at Sandia National Laboratories to evaluate the reliability of ALWR passive safety features in the context of probabilistic risk assessment (PRA). The purpose of this paper is to provide a brief overview of the approach to this study. The quantification of passive system reliability is not as straightforward as for active systems, due to the lack of operating experience, and to the greater uncertainty in the governing physical phenomena. Thus, the adequacy of current methods for evaluating system reliability must be assessed, and alternatives proposed if necessary. For this study, the Westinghouse Advanced Passive 600 MWe reactor (AP600) was chosen as the advanced reactor for analysis, because of the availability of AP600 design information. This study compares the reliability of AP600 emergency cooling system with that of corresponding systems in a current generation reactor

  14. 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)

  15. Design of integrated passive safety system (IPSS) for ultimate passive safety of nuclear power plants

    International Nuclear Information System (INIS)

    Chang, Soon Heung; Kim, Sang Ho; Choi, Jae Young

    2013-01-01

    containment outside into integrated passive safety tanks (IPSTs). The coolant in the IPSTs is used for various functions in accident scenarios. Also, potential problems for the realistic installation of the IPSS are proposed and the solutions to these problems are schematically described. IPSS is the design for the passive safety enhancement in preparation for a loss of AC power. Consequently, it is designed for the supplementation and enhancement of current nuclear power plants, not as a replacement. The specific optimization design for each current or future reactor will be studied as further works

  16. Proposed and existing passive and inherent safety-related structures, systems, and components (building blocks) for advanced light-water reactors

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Moses, D.L.; Lewis, E.B.; Gibson, R.; Pearson, R.; Reich, W.J.; Murphy, G.A.; Staunton, R.H.; Kohn, W.E.

    1989-10-01

    A nuclear power plant is composed of many structures, systems, and components (SSCs). Examples include emergency core cooling systems, feedwater systems, and electrical systems. The design of a reactor consists of combining various SSCs (building blocks) into an integrated plant design. A new reactor design is the result of combining old SSCs in new ways or use of new SSCs. This report identifies, describes, and characterizes SSCs with passive and inherent features that can be used to assure safety in light-water reactors. Existing, proposed, and speculative technologies are described. The following approaches were used to identify the technologies: world technical literature searches, world patent searches, and discussions with universities, national laboratories and industrial vendors. 214 refs., 105 figs., 26 tabs

  17. Proposed and existing passive and inherent safety-related structures, systems, and components (building blocks) for advanced light-water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W.; Moses, D.L.; Lewis, E.B.; Gibson, R.; Pearson, R.; Reich, W.J.; Murphy, G.A.; Staunton, R.H.; Kohn, W.E.

    1989-10-01

    A nuclear power plant is composed of many structures, systems, and components (SSCs). Examples include emergency core cooling systems, feedwater systems, and electrical systems. The design of a reactor consists of combining various SSCs (building blocks) into an integrated plant design. A new reactor design is the result of combining old SSCs in new ways or use of new SSCs. This report identifies, describes, and characterizes SSCs with passive and inherent features that can be used to assure safety in light-water reactors. Existing, proposed, and speculative technologies are described. The following approaches were used to identify the technologies: world technical literature searches, world patent searches, and discussions with universities, national laboratories and industrial vendors. 214 refs., 105 figs., 26 tabs.

  18. Research and development on reduced-moderation light water reactor with passive safety features (Contract research)

    International Nuclear Information System (INIS)

    Iwamura, Takamichi; Okubo, Tsutomu; Akie, Hiroshi; Kugo, Teruhiko; Yonomoto, Taisuke; Kureta, Masatoshi; Ishikawa, Nobuyuki; Nagaya, Yasunobu; Araya, Fumimasa; Okajima, Shigeaki; Okumura, Keisuke; Suzuki, Motoe; Mineo, Hideaki; Nakatsuka, Toru

    2004-06-01

    The present report contains the achievement of 'Research and Development on Reduced-moderation Light Water Reactor with Passive Safety Features', which was performed by Japan Atomic Energy Research Institute (JAERI), Hitachi Ltd., Japan Atomic Power Company and Tokyo Institute of Technology in FY2000-2002 as the innovative and viable nuclear energy technology (IVNET) development project operated by the Institute of Applied Energy (IAE). In the present project, the reduced-moderation water reactor (RMWR) has been developed to ensure sustainable energy supply and to solve the recent problems of nuclear power and nuclear fuel cycle, such as economical competitiveness, effective use of plutonium and reduction of spent fuel storage. The RMWR can attain the favorable characteristics such as high burnup, long operation cycle, multiple recycling of plutonium (Pu) and effective utilization of uranium resources based on accumulated LWR technologies. Our development target is 'Reduced-moderation Light Water Reactor with Passive Safety Features' with innovative technologies to achieve above mentioned requirement. Electric power is selected as 300 MWe considering anticipated size required for future deployment. The reactor core consists of MOX fuel assemblies with tight lattice arrangement to increase the conversion ratio. Design targets of the core specification are conversion ratio more than unity, negative void reactivity feedback coefficient to assure safety, discharged burnup more than 60 GWd/t and operation cycle more than 2 years. As for the reactor system, a small size natural circulation BWR with passive safety systems is adopted to increase safety and reduce construction cost. The results obtained are as follows: As regards core design study, core design was performed to meet the goal. Sequence of startup operation was constructed for the RMWR. As the plant design, plant system was designed to achieve enhanced economy using passive safety system effectively. In

  19. A Methodology for Modeling Nuclear Power Plant Passive Component Aging in Probabilistic Risk Assessment under the Impact of Operating Conditions, Surveillance and Maintenance Activities

    Science.gov (United States)

    Guler Yigitoglu, Askin

    In the context of long operation of nuclear power plants (NPPs) (i.e., 60-80 years, and beyond), investigation of the aging of passive systems, structures and components (SSCs) is important to assess safety margins and to decide on reactor life extension as indicated within the U.S. Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) Program. In the traditional probabilistic risk assessment (PRA) methodology, evaluating the potential significance of aging of passive SSCs on plant risk is challenging. Although passive SSC failure rates can be added as initiating event frequencies or basic event failure rates in the traditional event-tree/fault-tree methodology, these failure rates are generally based on generic plant failure data which means that the true state of a specific plant is not reflected in a realistic manner on aging effects. Dynamic PRA methodologies have gained attention recently due to their capability to account for the plant state and thus address the difficulties in the traditional PRA modeling of aging effects of passive components using physics-based models (and also in the modeling of digital instrumentation and control systems). Physics-based models can capture the impact of complex aging processes (e.g., fatigue, stress corrosion cracking, flow-accelerated corrosion, etc.) on SSCs and can be utilized to estimate passive SSC failure rates using realistic NPP data from reactor simulation, as well as considering effects of surveillance and maintenance activities. The objectives of this dissertation are twofold: The development of a methodology for the incorporation of aging modeling of passive SSC into a reactor simulation environment to provide a framework for evaluation of their risk contribution in both the dynamic and traditional PRA; and the demonstration of the methodology through its application to pressurizer surge line pipe weld and steam generator tubes in commercial nuclear power plants. In the proposed methodology, a

  20. Sodium hydroxide injection passivation work for the reactor water clean-up system in a new ABWR plant

    International Nuclear Information System (INIS)

    Wen, Tung-Jen; Lu, Ju-Huang

    2012-09-01

    Several studies have identified that Co-58 and Co-60 as the primary source of radiation build up on out-of-core components in new BWR plants. The deposition rate of Co on stainless steel and carbon steel is shown to be controlled mainly by the thickness of oxide films and its morphology formed through pretreatment. The passivation treatment was implemented accordingly at Lungmen unit 1 in an ABWR plant in September 2010. It is determined that the passivation conditions should be maintained at the temperature of 180∼230 deg. C, pH of 8.0∼8.5 and dissolved oxygen content over 400 ppb. The films would provide effective protection against radioactive deposition. The application of the pre-filming process on piping before the pre-operation is done during the flow induced vibration test (FIV) period. The protectiveness of stable magnetite can be increased by the pH control under the specific condition. The pre-filming control process and evaluation of passivation effectiveness is discussed in detail based on the surface analysis of the passivated specimens. Many efforts have been devoted to sodium hydroxide injection method for pH control of the system through the filter demineralizer under smooth operation. A comparison of test specimens on the properties of oxide film formed between laboratory and in-plant tests through alkaline treatment are also shown in this report. (authors)

  1. Technical Meeting on Passive Shutdown Systems for Liquid Metal-Cooled Fast Reactors. Working Material

    International Nuclear Information System (INIS)

    2015-01-01

    A major focus of the design of modern fast reactor systems is on inherent and passive safety. Specific systems to improve reactor safety performance during accidental transients have been developed in nearly all fast reactor programs, and a large number of proposed systems have reached various stages of maturity. This Technical Meeting on Passive Shutdown Systems for Fast Reactors, which was recommended by the Technical Working Group on Fast Reactors (TWG-FR), addressed Member States’ expressed need for information exchange on projects and programs in the field, as well as for the identification of priorities based on the analysis of technology gaps to be covered through R&D activities. This meeting was limited to shutdown systems only, and did not include other passive features such as natural circulation decay heat removal systems etc.; however the meeting catered to passive shutdown safety devices applicable to all types of fast neutron systems. It was agreed to initiate a new study and produce a Nuclear Energy Series (NES) Technical Report to collect information about the existing operational systems as well as innovative concepts under development. This will be a useful source for member states interested in gaining technical expertise to develop passive shutdown systems as well as to highlight the importance and development in this area

  2. Unlimited cooling capacity of the passive-type emergency core cooling system of the MARS reactor

    International Nuclear Information System (INIS)

    Bandini, G.; Caira, M.; Naviglio, A.; Sorabella, L.

    1995-01-01

    The MARS nuclear plant is equipped with a 600 MWth PWR type nuclear steam supply system, with completely innovative engineered core safeguards. The most relevant innovative safety system of this plant is its Emergency Core Cooling System, which is completely passive (with only one non static component). The Emergency Core Cooling System (ECCS) of the MARS reactor is natural-circulation, passive-type, and its intervention follows a core flow decrease, whatever was the cause. The operation of the system is based on a cascade of three fluid systems, functionally interfacing through heat exchangers; the first fluid system is connected to the reactor vessel and the last one includes an atmospheric-pressure condenser, cooled by external air. The infinite thermal capacity of the final heat sink provides the system an unlimited autonomy. The capability and operability of the system are based on its integrity and on the integrity of the primary coolant boundary (both of them are permanently enclosed in a pressurized containment; 100% redundancy is also foreseen) and on the operation of only one non static component (a check valve), with 400% redundancy. In the paper, all main thermal hydraulic transients occurring as a consequence of postulated accidents are analysed, to verify the capability of the passive-type ECCS to intervene always in time, without causing undue conditions of reduced coolability of the core (DNB, etc.), and to verify its capability to guarantee a long-term (indefinite) coolability of the core without the need of any external intervention. (author)

  3. Performance Assessment of Passive Gaseous Provisions (PGAP). Report of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)

    International Nuclear Information System (INIS)

    2013-07-01

    The International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) was launched in 2000 on the basis of IAEA General Conference resolution GC(44)/RES/21. INPRO helps to ensure the availability of sustainable nuclear energy in the 21st century and seeks to bring together all interested Member States - both technology holders and technology users - to consider joint actions to achieve desired innovations. To contribute to an international consensus on the definition of the reliability of passive systems that involve natural circulation, and on a methodology to assess this reliability, INPRO initiated a collaborative project on Performance Assessment of Passive Gaseous Provisions (PGAP) in 2007. Advanced nuclear reactor designs incorporate several passive systems in addition to active ones, not only to enhance the operational safety of the reactors but also to mitigate the consequences of a severe accident should one occur. However, the reliability of passive safety systems is crucial and must be assessed before they are used extensively in future nuclear power plants. Several physical parameters affect the performance of a passive safety system, and their values at the time of operation are a priori unknown. The functions of many passive systems are based on thermohydraulic principles, which until recently were considered as not being subject to any kind of failure. Hence, large and consistent efforts are required to quantify the reliability of such systems. Three participants from three INPRO Member States were involved in this collaborative project. Reliability methods for passive systems (RMPS) and assessment of passive system reliability (APSRA) methodologies were used by the participants to assess the performance and reliability of the passive decay heat removal system of the French gas cooled fast reactor design for station blackout and a loss of coolant accident combined with loss of off-site power, respectively. This publication presents the

  4. Passive autocatalytic recombiners for combustible gas control in advanced light water reactors

    International Nuclear Information System (INIS)

    Wolff, U.; Sliter, G.

    2004-01-01

    A key aspect of the worldwide effort to develop advanced nuclear power plants is designing to address severe accident phenomena, including the generation of hydrogen during core melt progression (metal-water and core-concrete reactions). This design work not only resolves safety concerns with hydrogen, but also supports the development of a technical basis for simplification of off-site emergency planning. The dominant challenge to any emergency planning approach is a large, early containment failure due to pressure excursions. Among the potential contributors to large and rapid increases in containment pressure is hydrogen combustion. The more improbable a containment-threatening combustion becomes, the more appropriate the argument for significant emergency planning simplification. As discussed in this paper, catalytic recombiners provide a means to passively and reliably limit hydrogen combustion to a continuous oxidation process with virtually no potential for containment failure in passive advanced light water reactors (ALWRs). (author)

  5. Experimental research progress on passive safety systems of Chinese advanced PWR

    International Nuclear Information System (INIS)

    Xiao Zejun; Zhuo Wenbin; Zheng Hua; Chen Bingde; Zong Guifang; Jia Dounan

    2003-01-01

    TMI and Chernobyl accidents, having pronounced impact on nuclear industries, triggered the governments as well as interested institutions to devote much attention to the safety of nuclear power plant and public's requirements on nuclear power plant safety were also going to be stricter and stricter. It is obvious that safety level of an ordinary light water reactor is no longer satisfactory to these requirements. Recently, the safety authorities have recommended the implementation of passive system to improve the safety of nuclear reactors. Passive safety system is one of the main differences between Chinese advanced PWR and other conventional PWR. The working principle of passive safety system is to utilize the gravity, natural convection (natural circulation) and stored energy to implement the system's safety function. Reactors with passive safety systems are not only safer, but also more economical. The passive safety system of Chinese advanced PWR is composed of three independent systems, i.e. passive containment cooling system, passive residual heat removal system and passive core makeup tank injection system. This paper is a summary of experimental research progress on passive containment cooling system, passive residual heat removal system and passive core makeup tank injection system

  6. Conceptual design of the Purdue compact torus/passive liner fusion reactor

    International Nuclear Information System (INIS)

    Terry, W.K.

    1981-01-01

    This proposal describes a program for the conceptual development of a novel fusion reactor design, the Purdue Compact Torus/Passive Liner Reactor. The key features of the concept are described and a comparison is made with a conventional tokamak

  7. Trends vs. reactor size of passive reactivity shutdown and control performance

    International Nuclear Information System (INIS)

    Wade, D.C.; Fujita, E.K.

    1988-01-01

    The focus of the US advanced reactor program since the cancellation of CRBR has been on inherent safety and cost reduction. The notion is to so design the reactor that in the event of an off normal condition, it brings itself to a safe shutdown condition and removes decay heat by reliance on ''inherent processes'' i.e., without reliance on devices requiring switching and outside sources of power. Such a reactor design would offer the potential to eliminate costly ''Engineered Safety Features,'' to lower capital costs, and to assuage public unease concerning reactor safety. For LMR concepts, the goal of passive reactivity shutdown has been approached in the US by designing the reactors for favorable relationships among the power, power/flow, and inlet temperature coefficients of reactivity, for high internal conversion ratio (yielding small burnup control swing), and for a primary pump coastdown time appropriately matched to the delayed neutron hold back of power decay upon negative reactivity input. The use of sodium bonded metallic fuel pins has facilitated the achievement of the passive shutdown design goals as a consequence of their high thermal conductivity and high effective heavy metal density. Alternately, core designs based on derated oxide pins may be able to achieve the passive shutdown features at the cost of larger core volume and increased initial fissile inventory. 8 refs., 12 figs., 1 tab

  8. A compact, inherently safe liquid metal reactor plant concept for terrestrial defense power applications

    International Nuclear Information System (INIS)

    Magee, P.M.; Dubberley, A.E.; Lutz, D.E.; Palmer, R.S.

    1987-01-01

    A compact, inherently safe, liquid metal reactor concept based on the GE PRISM innovative LMR design has been developed for terrestrial defense power applications in the 2-50 MWe range. The concept uses a small, sodium-cooled, U-5%Zr metal fueled reactor contained within two redundant steel vessels. The core is designed to operate at a low power density and temperature (925 F) and can operate 30 years without refueling. One two primary coolant loops, depending upon the plant size, transport heat from the core to sodium-to-air, double-wall heat exchangers. Power is produced by a gas turbine operated in a closed ''bottoming'' cycle that employs intercoolers between the compressor stages and a recuperator. Inherent safety is provided by passive means only; operator action is not required to ensure plant safety even for events normally considered Beyond Design Basis Accidents. In addition to normal shutdown heat removal via the sodium-to-air heat exchangers, the design utilizes an inherently passive radiant vessel auxiliary cooling system similar to that designed for PRISM. The use of an air cycle gas turbine eliminates the cost and complexity of the sodium-water reactor pressure relief system required for a steam cycle sodium-cooled reactor

  9. Considerations on monitoring needs of advanced, passive safety light water reactors for severe accident management

    International Nuclear Information System (INIS)

    Bava, G.; Zambardi, F.

    1992-01-01

    This paper deals with problems concerning information and related instrumentation needs for Accident Management (AM), with special emphasis on Severe Accidents (SA) in the new advanced, passive safety Light Water Reactors (PLWR), presently in a development stage. The passive safety conception adopted in the plants concerned goes parallel with a deeper consideration of SA, that reflects the need of increasing the plant resistance against conditions going beyond traditional ''design basis accidents''. Further, the role of Accident Management (AM) is still emphasized as last step of the defence in depth concept, in spite of the design efforts aimed to reduce human factor importance; as a consequence, the availability of pertinent information on actual plant conditions remains a necessary premise for performing preplanned actions. This information is essential to assess the evolution of the accident scenarios, to monitor the performances of the safety systems, to evaluate the ultimate challenge to the plant safety, and to implement the emergency operating procedures and the emergency plans. Based on these general purposes, the impact of the new conception on the monitoring structure is discussed, furthermore reference is made to the accident monitoring criteria applied in current plants to evaluate the requirements for possible solutions. (orig.)

  10. A passive emergency heat sink for water-cooled reactors with particular application to CANDU reactors

    International Nuclear Information System (INIS)

    Spinks, N.J.

    1996-01-01

    Water in an overhead pool can serve as a general-purpose passive emergency heat sink for water-cooled reactors. It can be used for containment cooling, for emergency depressurization of the heat transport-system, or to receive any other emergency heat, such as that from the CANDU moderator. The passive emergency water system provides in-containment depressurization of steam generators and no other provision is needed for supply of low-pressure emergency water to the steam generators. For containment cooling, the pool supplies water to the tube side of elevated tube banks inside containment. The elevation with respect to the reactor heat source maximizes heat transport, by natural convection, of hot containment gases. This effective heat transport combines with the large heat-transfer coefficients of tube banks, to reduce containment overpressure during accidents. Cooled air from the tube banks is directed past the break in the heat-transport system, to facilitate removal of hydrogen using passive catalytic recombiners. (author)

  11. EP1000 passive plant description

    International Nuclear Information System (INIS)

    Saiu, G.

    1999-01-01

    In 1994, a group of European Utilities, together with Westinghouse and its Industrial Partner GENESI (an Italian consortium including ANSALDO and FIAT), initiated a program designated EPP (European Passive Plant) to evaluate Westinghouse Passive Nuclear Plant Technology for application in Europe. In Phase I of the European Passive Plant Program which was completed in 1996, a 1000 MWe passive plant reference design (EP1000) was established which conforms to the European Utility Requirements (EUR) and is expected to meet the European Safety Authorities requirements. Phase 2 of the program was initiated in 1997 with the objective of developing the Nuclear Island design details and performing supporting analyses to start development of Safety Case Report (SCR) for submittal to European Licensing Authorities. The first part of Phase 2, 'Design Definition' phase (Phase 2A) will be completed at the end of 1998, the main efforts being design definition of key systems and structures, development of the Nuclear Island layout, and performing preliminary safety analyses to support design efforts. The second part, 'Phase 2B', includes both the analyses and evaluations required to demonstrate the adequacy of the design, and to support the preparation of Safety Case Report. The second part of Phase 2 of the program will start at the beginning of 1999 and will be completed in the 2001. Incorporation of the EUR has been a key design requirement for the EP1000 from the beginning of the program. Detailed design solutions to meet the EUR have been defined and the safety approach has also been developed based on the EUR guidelines. This paper integrates and updates the plant description reported in the IAEA TECDOC-968. The most significant developments of the EP1000 plant design during Phase 2A of the EPP program are described and reference is made to the key design requirements set by the EUR Rev. B document. (author)

  12. The PRISM concept for a safe, economic and testable liquid metal fast reactor plant

    International Nuclear Information System (INIS)

    Berglund, R.C.; Salerno, L.N.; Tippets, F.E.

    1987-01-01

    The PRISM project is underway at General Electric as part of an advanced reactor conceptual design program sponsored by the US Department of Energy. The PRISM concept emphasizes inherent safety, modular construction, and factory fabrication. These features are intended to improve the basis for public acceptance, reduce cost,improve licensability, and reduce the risk of schedule delays and cost increases during construction. A PRISM power plant comprises a number of reactor modules. The relatively small size of the reactor module facilitates the use of passive, inherent self-shutdown and shutdown heat removal features for safe accommodation of accidents. These inherent safety features permit simplification and reduction of conventional safety-related systems in the plant. Testing of a full-size prototype reactor module is planned in the late 1990's to demonstrate these inherent safety characteristics. It is intended that the results of the test be used to obtain certification of the design by the US Nuclear Regulatory Commission preparatory to use of reactor modules built to this standard design in licensed commercial plants

  13. Activities of passive cooling applications and simulation of innovative nuclear power plant design

    International Nuclear Information System (INIS)

    Aglar, F.; Tanrykut, A.

    2002-01-01

    This paper gives a general insight on activities of the Turkish Atomic Energy Authority (TAEA) concerning passive cooling applications and simulation of innovative nuclear power plant design. The condensation mode of heat transfer plays an important role for the passive heat removal application in advanced water-cooled reactor systems. But it is well understood that the presence of noncondesable (NC) gases can greatly inhibit the condensation process due to the build up of NC gas concentration at the liquid/gas interface. The isolation condenser of passive containment cooling system of the simplified boiling water reactors is a typical application area of in-tube condensation in the presence of NC. The test matrix of the experimental investigation undertaken at the METU-CTF test facility (Middle East Technical University, Ankara) covers the range of parameters; Pn (system pressure) : 2-6 bar, Rev (vapor Reynolds number): 45,000-94,000, and Xi (air mass fraction): 0-52%. This experimental study is supplemented by a theoretical investigation concerning the effect of mixture flow rate on film turbulence and air mass diffusion concepts. Recently, TAEA participated to an international standard problem (OECD ISP-42) which covers a set of simulation of PANDA test facility (Paul Scherrer Institut-Switzerland) for six different phases including different natural circulation modes. The concept of condensation in the presence of air plays an important role for performance of heat exchangers, designed for passive containment cooling, which in turn affect the natural circulation behaviour in PANDA systems. (author)

  14. Identification and characterization of passive safety system and inherent safety feature building blocks for advanced light-water reactors

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    1989-01-01

    Oak Ridge National Laboratory (ORNL) is investigating passive and inherent safety options for Advanced Light-Water Reactors (ALWRs). A major activity in 1989 includes identification and characterization of passive safety system and inherent safety feature building blocks, both existing and proposed, for ALWRs. Preliminary results of this work are reported herein. This activity is part of a larger effort by the US Department of Energy, reactor vendors, utilities, and others in the United States to develop improved LWRs. The Advanced Boiling Water Reactor (ABWR) program and the Advanced Pressurized Water Reactor (APWR) program have as goals improved, commercially available LWRs in the early 1990s. The Advanced Simplified Boiling Water Reactor (ASBWR) program and the AP-600 program are developing more advanced reactors with increased use of passive safety systems. It is planned that these reactors will become commercially available in the mid 1990s. The ORNL program is an exploratory research program for LWRs beyond the year 2000. Desired long-term goals for such reactors include: (1) use of only passive and inherent safety, (2) foolproof against operator errors, (3) malevolence resistance against internal sabotage and external assault and (4) walkaway safety. The acronym ''PRIME'' [Passive safety, Resilient operation, Inherent safety, Malevolence resistance, and Extended (walkaway) safety] is used to summarize these desired characteristics. Existing passive and inherent safety options are discussed in this document

  15. A concept of JAERI passive safety light water reactor system (JPSR)

    Energy Technology Data Exchange (ETDEWEB)

    Murao, Y.; Araya, F.; Iwamura, T. [Japan Atomic Energy Research Institute, Tokai-mura (Japan)

    1995-09-01

    The Japan Atomic Energy Research Institute (JAERI) proposed a passive safety reactor system concept, JPSR, which was developed for reducing manpower in operation and maintenance and influence of human errors on reactor safety. In the concept the system was extremely simplified. The inherent matching nature of core generation and heat removal rate within a small volume change of the primary coolant is introduced by eliminating chemical shim and adopting in-vessel control rod drive mechanism units, a low power density core and once-through steam generators. In order to simplify the system, a large pressurizer, canned pumps, passive engineered-safety-features-system (residual heat removal system and coolant injection system) are adopted and the total system can be significantly simplified. The residual heat removal system is completely passively actuated in non-LOCAs and is also used for depressurization of the primary coolant system to actuate accumulators in small break LOCAs and reactor shutdown cooling system in normal operation. All of systems for nuclear steam supply system are built in the containment except for the air coolers as a the final heat sink of the passive residual heat removal system. Accordingly the reliability of the safety system and the normal operation system is improved, since most of residual heat removal system is always working and a heat sink for normal operation system is {open_quotes}safety class{close_quotes}. In the passive coolant injection system, depressurization of the primary cooling system by residual heat removal system initiates injection from accumulators designed for the MS-600 in medium pressure and initiates injection from the gravity driven coolant injection pool at low pressure. Analysis with RETRAN-02/MOD3 code demonstrated the capability of passive load-following, self-power-controllability, cooling and depressurization.

  16. Plant maintenance and advanced reactors issue, 2008

    Energy Technology Data Exchange (ETDEWEB)

    Agnihotri, Newal [ed.

    2009-09-15

    The focus of the September-October issue is on plant maintenance and advanced reactors. Major articles/reports in this issue include: Technologies of national importance, by Tsutomu Ohkubo, Japan Atomic Energy Agency, Japan; Modeling and simulation advances brighten future nuclear power, by Hussein Khalil, Argonne National Laboratory, Energy and desalination projects, by Ratan Kumar Sinha, Bhabha Atomic Research Centre, India; A plant with simplified design, by John Higgins, GE Hitachi Nuclear Energy; A forward thinking design, by Ray Ganthner, AREVA; A passively safe design, by Ed Cummins, Westinghouse Electric Company; A market-ready design, by Ken Petrunik, Atomic Energy of Canada Limited, Canada; Generation IV Advanced Nuclear Energy Systems, by Jacques Bouchard, French Commissariat a l'Energie Atomique, France, and Ralph Bennett, Idaho National Laboratory; Innovative reactor designs, a report by IAEA, Vienna, Austria; Guidance for new vendors, by John Nakoski, U.S. Nuclear Regulatory Commission; Road map for future energy, by John Cleveland, International Atomic Energy Agency, Vienna, Austria; and, Vermont's largest source of electricity, by Tyler Lamberts, Entergy Nuclear Operations, Inc. The Industry Innovation article is titled Intelligent monitoring technology, by Chris Demars, Exelon Nuclear.

  17. Passive safety systems reliability and integration of these systems in nuclear power plant PSA

    International Nuclear Information System (INIS)

    La Lumia, V.; Mercier, S.; Marques, M.; Pignatel, J.F.

    2004-01-01

    Innovative nuclear reactor concepts could lead to use passive safety features in combination with active safety systems. A passive system does not need active component, external energy, signal or human interaction to operate. These are attractive advantages for safety nuclear plant improvements and economic competitiveness. But specific reliability problems, linked to physical phenomena, can conduct to stop the physical process. In this context, the European Commission (EC) starts the RMPS (Reliability Methods for Passive Safety functions) program. In this RMPS program, a quantitative reliability evaluation of the RP2 system (Residual Passive heat Removal system on the Primary circuit) has been realised, and the results introduced in a simplified PSA (Probabilistic Safety Assessment). The scope is to get out experience of definition of characteristic parameters for reliability evaluation and PSA including passive systems. The simplified PSA, using event tree method, is carried out for the total loss of power supplies initiating event leading to a severe core damage. Are taken into account: failures of components but also failures of the physical process involved (e.g. natural convection) by a specific method. The physical process failure probabilities are assessed through uncertainty analyses based on supposed probability density functions for the characteristic parameters of the RP2 system. The probabilities are calculated by MONTE CARLO simulation coupled to the CATHARE thermalhydraulic code. The yearly frequency of the severe core damage is evaluated for each accident sequence. This analysis has identified the influence of the passive system RP2 and propose a re-dimensioning of the RP2 system in order to satisfy the safety probabilistic objectives for reactor core severe damage. (authors)

  18. Passive cooling applications for nuclear power plants using pulsating steam-water heat pipes

    International Nuclear Information System (INIS)

    Aparna, J.; Chandraker, D.K.

    2015-01-01

    Gen IV reactors incorporate passive principles in their system design as an important safety philosophy. Passive safety systems use inherent physical phenomena for delivering the desired safe action without any external inputs or intrusion. The accidents in Fukushima have renewed the focus on passive self-manageable systems capable of unattended operation, for long hours even in extended station blackout (SBO) and severe accident conditions. Generally, advanced reactors use water or atmospheric air as their ultimate heat sink and employ passive principles in design for enhanced safety. This paper would be discussing the experimental results on pulsating steam water heat-pipe devices and their applications in passive cooling. (author)

  19. Results of a Demonstration Assessment of Passive System Reliability Utilizing the Reliability Method for Passive Systems (RMPS)

    Energy Technology Data Exchange (ETDEWEB)

    Bucknor, Matthew; Grabaskas, David; Brunett, Acacia; Grelle, Austin

    2015-04-26

    Advanced small modular reactor designs include many advantageous design features such as passively driven safety systems that are arguably more reliable and cost effective relative to conventional active systems. Despite their attractiveness, a reliability assessment of passive systems can be difficult using conventional reliability methods due to the nature of passive systems. Simple deviations in boundary conditions can induce functional failures in a passive system, and intermediate or unexpected operating modes can also occur. As part of an ongoing project, Argonne National Laboratory is investigating various methodologies to address passive system reliability. The Reliability Method for Passive Systems (RMPS), a systematic approach for examining reliability, is one technique chosen for this analysis. This methodology is combined with the Risk-Informed Safety Margin Characterization (RISMC) approach to assess the reliability of a passive system and the impact of its associated uncertainties. For this demonstration problem, an integrated plant model of an advanced small modular pool-type sodium fast reactor with a passive reactor cavity cooling system is subjected to a station blackout using RELAP5-3D. This paper discusses important aspects of the reliability assessment, including deployment of the methodology, the uncertainty identification and quantification process, and identification of key risk metrics.

  20. Selection of power plant elements for future reactor space electric power systems

    International Nuclear Information System (INIS)

    Buden, D.; Bennett, G.A.; Copper, K.

    1979-09-01

    Various types of reactor designs, electric power conversion equipment, and reject-heat systems to be used in nuclear reactor power plants for future space missions were studied. The designs included gas-cooled, liquid-cooled, and heat-pipe reactors. For the power converters, passive types such as thermoelectric and thermionic converters and dynamic types such as Brayton, potassium Rankine, and Stirling cycles were considered. For the radiators, heat pipes for transfer and radiating surface, pumped fluid for heat transfer with fins as the radiating surface, and pumped fluid for heat transfer with heat pipes as the radiating surface were considered. After careful consideration of weights, sizes, reliabilities, safety, and development cost and time, a heat-pipe reactor design, thermoelectric converters, and a heat-pipe radiator for an experimental program were selected

  1. Study on dual plant concept for the next generation boiling water reactors

    International Nuclear Information System (INIS)

    Sato, Takashi; Oikawa, Hirohide

    1999-01-01

    The paper presents the study results on the basic concept of dual BWRs. For the convenience, we call the concept here as Trial Study on BWR dual concept (TSBWR dual). The concept is general and applicable to all BWRs which have internal recirculation pumps (RIP). The TSBWR dual is a plant concept of dual BWRs contained in a same secondary containment building. The plant output is from 2 x l,350 MWe up to 2 x 1,700 MWe. This concept is mainly aiming at safety improvement and cost savings of the next generation BWRs. The TSBWR dual has two RPVs and two dry wells (DW). It has, however, only one wet well (WW) and only one R/B. The WW and the R/B are shared by the dual reactors. The operating floor is also shared by the two reactors. The TSBWR dual has both passive safety systems and active safety systems. They are also shared between the two reactors. A lot of sharing between the dual reactors enables significant cost savings accompanied by the power increase up to 3,400 MWe. Although the TSBWR dual consists of two reactors, the simplified cylindrical configuration of the key structures and reduction of the R/B height can minimize the plant construction period. The TSBWR dual provides a concept with which we can challenge to construct a dual BWR plant in the near future. (author)

  2. Inherently safe light water reactors

    International Nuclear Information System (INIS)

    Ise, Takeharu

    1987-01-01

    Today's large nuclear power reactors of world-wise use have been designed based on the philosophy. It seems that recent less electricity demand rates, higher capital cost and the TMI accident let us acknowledge relative small and simplified nuclear plants with safer features, and that Chernobyl accident in 1983 underlines the needs of intrinsic and passive safety characteristics. In such background, several inherently safe reactor concepts have been presented abroad and domestically. First describing 'Can inherently safe reactors be designed,' then I introduce representative reactor concepts of inherently safe LWRs advocated abroad so far. All of these innovative reactors employ intrinsic and passive features in their design, as follows: (1) PIUS, an acronym for Process Inherent Ultimate Safety, or an integral PWR with passive heat sink and passive shutdown mechanism, advocated by ASEA-ATOM of Sweden. (2) MAP(Minimum Attention Plant), or a self-pressurized, natural circulation integral PWR, promoted by CE Inc. of the U.S. (3) TPS(TRIGA Power System), or a compact PWR with passive heat sink and inherent fuel characteristics of large prompt temperature coefficient, prompted by GA Technologies Inc. of the U.S. (4) PIUS-BWR, or an inherently safe BWR employing passively actuated fluid valves, in competition with PIUS, prompted by ORNL of the U.S. Then, I will describe the domestic trends in Japan and the innovative inherently safe LWRs presented domestically so far. (author)

  3. Nuclear Power Plant with VK-300 Reactor. Simplicity and Passivity as a Way to Provide Enhanced Safety and Economic Competitive Ability

    International Nuclear Information System (INIS)

    Kuznetsov, Y. N.; Lisitsa, F. D.; Tokarev, Y. I.; Glazkov, O. M.

    1998-01-01

    NPP with simplified boiling reactors with coolant natural circulation is developed by RDIPE. removal of residual heat releases under any emergency conditions is supposed to be put into effect with maximum application of passive systems and devices. (author)

  4. Regulatory requirements for desalination plant coupled with nuclear reactor plant

    International Nuclear Information System (INIS)

    Yune, Young Gill; Kim, Woong Sik; Jo, Jong Chull; Kim, Hho Jung; Song, Jae Myung

    2005-01-01

    A small-to-medium sized reactor has been developed for multi-purposes such as seawater desalination, ship propulsion, and district heating since early 1990s in Korea. Now, the construction of its scaled-down research reactor, equipped with a seawater desalination plant, is planned to demonstrate the safety and performance of the design of the multi-purpose reactor. And the licensing application of the research reactor is expected in the near future. Therefore, a development of regulatory requirements/guides for a desalination plant coupled with a nuclear reactor plant is necessary for the preparation of the forthcoming licensing review of the research reactor. In this paper, the following contents are presented: the design of the desalination plant, domestic and foreign regulatory requirements relevant to desalination plants, and a draft of regulatory requirements/guides for a desalination plant coupled with a nuclear reactor plant

  5. Assessment of passive safety system of a Small Modular Reactor (SMR)

    International Nuclear Information System (INIS)

    Butt, Hassan Nawaz; Ilyas, Muhammad; Ahmad, Masroor; Aydogan, Fatih

    2016-01-01

    Highlights: • The MASLWR test facility has been modeled in RELAP5-SCDAP. The model is validated by comparing the simulation results with the experimental data. • Results obtained from various transients show that high pressure vent and sump recirculation lines provide natural circulation flow path for long term cooling of core. • New scenarios are considered in which the effect of vent and sump recirculation valves failure has been investigated. • It is found from the results that continuous loss of inventory occurs due to lack of recirculation. • It is concluded that the high pressure vent valves in the MASLWR safety system require more redundancy. - Abstract: Innovative SMRs are designed with enhanced safety features based on lessons learnt from past experience of plant operation. Reliance on natural circulation and addition of passive safety systems made them inherently safe and simple in design. It is required to study reliability assessment of passive safety systems during postulated transients prior to their deployment on commercial scale. Test facilities and best estimate system codes are playing significant role in assessment of passive safety systems as well as in design, certification and evaluation of these innovative types of reactors. RELAP5 code is widely used for thermal-hydraulic analysis of nuclear reactors. In this work, the passive safety systems of Multi-Application Small Light Water (MASLWR) have been assessed. The complete loop of the MASLWR test facility has been modeled in RELAP5-SCDAP Mod 4.0. The RELAP5 model is validated by comparing the simulation results with the experimental data. Results obtained for various transients show that high pressure vent and sump recirculation lines provide natural circulation flow path for long term cooling of core to avoid core heat up. Some of the components of passive safety system of MASLWR still rely on active power. Therefore, it was necessary to investigate their performance under failure

  6. SWR 1000: the main design features of the advanced boiling water reactor with passive safety systems

    International Nuclear Information System (INIS)

    Carsten, Pasler

    2007-01-01

    The SWR-1000 (1000 MW) is a boiling water reactor whose economic efficiency in comparison with large-capacity designs is achieved by deploying very simple passive safety equipment, simplified systems for plant operation, and a very simple plant configuration in which systems engineering is optimized and dependence on electrical and instrumentation and control systems is reduced. In addition, systems and components that require protection against natural and external man-made hazards are accommodated in such a way that as few buildings as possible have to be designed to withstand the loads from such events. The fuel assemblies have been enlarged from a 10*10 rod array to a 12*12 array. This reduces the total number of fuel assemblies in the core and thus also the number of control rods and control rod drives, as well as in-core neutron flux monitors. The design owes its competitiveness to the fact that investment costs, maintenance costs and fuel cycle costs are all lower. In addition, refueling outages are shorter, thanks to the reduced scope of outage activities. The larger fuel assemblies have been extensively and successfully tested, as have all of the other new components and systems incorporated into the plant design. As in existing plants, the forced coolant circulation method is deployed, ensuring problem-free startup, and enabling plant operators to adjust power rapidly in the high power range (70%-100%) without moving the control rods, as well as allowing spectral-shift and stretch-out operation. The plant safety concept is based on a combination of passive safety systems and a reduced number of active safety systems. All postulated accidents can be controlled using passive systems alone. Control of a postulated core melt accident is assured with considerable safety margins thanks to passive flooding of the containment for in-vessel melt retention. The SWR-1000 is compliant with international nuclear codes and standards, and is also designed to withstand

  7. Passive and engineered safety features of the prototype fast reactor (PFR), Dounreay

    International Nuclear Information System (INIS)

    Gregory, C.V.

    1991-01-01

    Prototype fast reactor (PFR) combines passive and engineered safety features. Natural convection, a strong negative power coefficient, the decay heat removal system, and a fuel design able to operate beyond failure are all inherent and passive safety features of the PFR. The reliable shutdown system and the protection provided against SGU leaks are example of engineered protection. Experience at PFR demonstrates the worth and potential of a range of passive and engineered safeguards

  8. Passive safety system of a super fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sutanto, E-mail: sutanto@fuji.waseda.jp [Cooperative Major in Nuclear Energy, Waseda University, Tokyo (Japan); Polytechnic Institute of Nuclear Technology—National Nuclear Energy Agency, Yogyakarta (Indonesia); Oka, Yoshiaki [The University of Tokyo, Tokyo (Japan)

    2015-08-15

    Highlights: • Passive safety system of a Super FR is proposed. • Total loss of feedwater flow and large LOCA are analyzed. • The criteria of MCST and core pressure are satisfied. - Abstract: Passive safety systems of a Super Fast Reactor are studied. The passive safety systems consist of isolation condenser (IC), automatic depressurization system (ADS), core make-up tank (CMT), gravity driven cooling system (GDCS), and passive containment cooling system (PCCS). Two accidents of total loss of feedwater flow and 100% cold-leg break large LOCA are analyzed by using the passive systems and the criteria of maximum cladding surface temperature (MCST) and maximum core pressure are satisfied. The isolation condenser can be used for mitigation of the accident of total loss of feedwater flow at both supercritical and subcritical pressures. The ADS is used for depressurization leading to a loss of coolant during line switching to operation of the isolation condenser at subcritical pressure. Use of CMT during line switching recovers the lost coolant. In case of large LOCA, GDCS can be used for core reflooding. Coolant vaporization in the core released to containment through the break is condensed by passive containment cooling system. The condensate flows to the GDCS pool by gravity force. The maximum cladding surface temperature (MCST) of the accident satisfies the criterion.

  9. Reactivity control in HTR power plants with respect to passive safety system. Summary

    Energy Technology Data Exchange (ETDEWEB)

    Barnert, H; Kugeler, K [Forschungszentrum Juelich GmbH (Germany). Inst. fuer Sicherheitsforschung und Reaktortechnik

    1996-12-01

    The R and D and Demonstration of the High Temperature Reactor (HTR) is described in overview. The HTR-MODULE power plant, as the most advanced concept, is taken for the description of the reactivity control in general. The idea of the ``modularization of the core`` of the HTR has been developed as the answer on the experiences of the core melt accident at Three Miles Island. The HTR module has two shutdown systems: The ``6 rods``-system for hot shutdown at the ``18 small absorber pebbles units`` - system for cold shutdown. With respect to the definition of ``Passive Systems`` of IAEA-TECDOC-626 the total reactivity control system of the HTR-MODULE is a passive system of category D, because it is an emergency reactor shutdown system based on gravity driven rods, and devices, activated by fail-safe trip logic. But reactivity control of the HTR does not only consist of these engineered safety system but does have a self-acting stabilization by the negative temperature coefficient of the reactivity, being rather effective in reactivity control. Examples from computer calculations are presented, and, in addition, experimental results from the ``Stuck Rod Experiment`` at the AVR reactor in Juelich. On the basis of this the proposal is made that ``self-acting stabilization as a quality of the function`` should be discussed as a new category in addition to the active and passive engineered safety systems, structures and components of IAEA-TECDOC-626. The requirements for a future ``catastrophe-free`` nuclear technology are presented. In the appendix the 7th amendment of the atomic energy act of the Federal Republic of Germany, effective 28 July 94, is given. (author).

  10. Trends vs. reactor size of passive reactivity shutdown and control performance

    International Nuclear Information System (INIS)

    Wade, D.C.; Fujita, E.K.

    1987-01-01

    For LMR concepts, the goal of passive reactivity shutdown has been approached in the US by designing the reactors for favorable relationships among the power, power/flow, and inlet temperature coefficients of reactivity, for high internal conversion ratio (yielding small burnup control swing), and for a primary pump coastdown time appropriately matched to the delayed neutron hold back of power decay upon negative reactivity input. The use of sodium bonded metallic fuel pins has facilitated the achievement of the massive shutdown design goals as a consequence of their high thermal conductivity and high effective heavy metal density. Alternately, core designs based on derated oxide pins may be able to achieve the passive shutdown features at the cost of larger core volume and increased initial fissile inventory. For LMR concepts, the passive decay heat removal goal of inherent safety has been approached in US designs by use of pool layouts, larger surface to volume ratio of the reactor vessel with natural draft air cooling of the vessel surface, elevations and redans which promote natural circulation through the core, and thermal mass of the pool contents sufficient to absorb that initial transient decay heat which exceeds the natural draft air cooling capacity. This paper describes current US ''inherently safe'' reactor design

  11. Design strategy for control of inherently safe reactors

    International Nuclear Information System (INIS)

    Chisholm, G.H.

    1984-01-01

    Reactor power plant safety is assured through a combination of engineered barriers to radiation release (e.g., reactor containment) in combination with active reactor safety systems to shut the reactor down and remove decay heat. While not specifically identified as safety systems, the control systems responsible for continuous operation of plant subsystems are the first line of defense for mitigating radiation releases and for plant protection. Inherently safe reactors take advantage of passive system features for decay-heat removal and reactor shutdown functions normally ascribed to active reactor safety systems. The advent of these reactors may permit restructuring of the present control system design strategy. This restructuring is based on the fact that authority for protection against unlikely accidents is, as much as practical, placed upon the passive features of the system instead of the traditional placement upon the PPS. Consequently, reactor control may be simplified, allowing the reliability of control systems to be improved and more easily defended

  12. A passive emergency heat sink for water cooled reactors with particular application to CANDU reg-sign reactors

    International Nuclear Information System (INIS)

    Spinks, N.J.

    1996-01-01

    Water in an overhead pool can serve as a general-purpose passive emergency heat sink for water-cooled reactors. It can be used for containment cooling, for emergency depressurization of the heat transport-system, or to receive any other emergency heat, such as that from the CANDU reg-sign moderator. The passive emergency water system provides in-containment depressurization of steam generators and no other provision is needed for supply of low-pressure emergency water to the steam generators. For containment cooling, the pool supplies water to the tube side of elevated tube banks inside containment. The elevation with respect to the reactor heat source maximizes heat transport, by natural convection, of hot containment gases. This effective heat transport combines with the large heat-transfer coefficients of tube banks, to reduce containment overpressure during accidents. Cooled air from the tube banks is directed past the break in the heat-transport system, to facilitate removal of hydrogen using passive catalytic recombiners

  13. New reactor programs from passive to pebble bed

    Energy Technology Data Exchange (ETDEWEB)

    Bruschi, H.J. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

    2002-07-01

    The market for new nuclear power plants is small and challenged by alternative means of electric power generation. Customers and countries may vary in their requirements for a new nuclear plant; but all have a common theme of seeking a design that possesses favorable economics. This paper sets forth the economic challenges a new nuclear plant must overcome. In particular, it delineates the capital cost, construction time, and generation cost required to compete with combined cycle gas electric power generation. The U.S. power generation market is used as a point of comparison. Following this, the portfolio of BNFL/ Westinghouse plant designs are described and the methods by which they will meet the economic challenges previously delineated will be discussed. The portfolio includes the family of passive plants originated by the AP600 Design Certification process in the U.S. These plants are marked by a high degree of safety and simplicity, short construction times, and superior economics. In addition, the effort to meet European requirements for passive plants will be described. Lastly, the paper explores some advanced nuclear designs that are not yet licensed, and the hope that they hold for meeting the industry challenge ahead. (author)

  14. New reactor programs from passive to pebble bed

    International Nuclear Information System (INIS)

    Bruschi, H.J.

    2002-01-01

    The market for new nuclear power plants is small and challenged by alternative means of electric power generation. Customers and countries may vary in their requirements for a new nuclear plant; but all have a common theme of seeking a design that possesses favorable economics. This paper sets forth the economic challenges a new nuclear plant must overcome. In particular, it delineates the capital cost, construction time, and generation cost required to compete with combined cycle gas electric power generation. The U.S. power generation market is used as a point of comparison. Following this, the portfolio of BNFL/ Westinghouse plant designs are described and the methods by which they will meet the economic challenges previously delineated will be discussed. The portfolio includes the family of passive plants originated by the AP600 Design Certification process in the U.S. These plants are marked by a high degree of safety and simplicity, short construction times, and superior economics. In addition, the effort to meet European requirements for passive plants will be described. Lastly, the paper explores some advanced nuclear designs that are not yet licensed, and the hope that they hold for meeting the industry challenge ahead. (author)

  15. Testing of the multi-application small light water reactor (MASLWR) passive safety systems

    International Nuclear Information System (INIS)

    Reyes, Jose N.; Groome, John; Woods, Brian G.; Young, Eric; Abel, Kent; Yao, You; Yoo, Yeon Jong

    2007-01-01

    Experimental thermal hydraulic research has been conducted at Oregon State University for the purpose of assessing the performance of a new reactor design concept, the multi-application small light water reactor (MASLWR). The MASLWR is a pressurized light water reactor design with a net output of 35 MWe that uses natural circulation in both normal and transient operation. Due to its small size, portability and modularity, the MASLWR design is well suited to help fill the potential need for grid appropriate reactor designs for smaller electricity grids as may be found in developing or remote regions. The purpose of the OSU MASLWR test facility is to assess the operation of the MASLWR under normal full operating pressure and full temperature conditions and to assess the passive safety systems under transient conditions. The data generated by the testing program will be used to assess computer code calculations and to provide a better understanding of the thermal-hydraulic phenomena in the design of the MASLWR NSSS. During this testing program, four tests were conducted at the OSU MASLWR test facility. These tests included one design basis accident and one beyond design basis accident. During the performance of these tests, plant operations to include start up, normal operation and shut down evolutions were demonstrated successfully

  16. Water treatment for the ISER [intrinsically safe and economical reactor] plant

    International Nuclear Information System (INIS)

    Sugawara, Ichiro.

    1985-01-01

    The ISER reactor assures inherent safety by causing the core, which is submerged in pool water containing a high boric acid concentration, to quickly shut down the nuclear reaction when overheating, pump trip or other problems occur. However, large quantities of pool water may cause difficulties in water quality control and waste management, resulting in higher costs. Therefore, the ISER as a total plant would not be publicly acceptable unless the water treatment and waste management system offer both safety balanced with reactor inherent safety, and economy counterbalanced by large quantities of pool water. This report clarifies the passive safety concept of possible waste treatment and management systems, and the ways to economically construct such facilities

  17. Passive containment system for a nuclear reactor

    International Nuclear Information System (INIS)

    Kleimola, F.W.

    1976-01-01

    A containment system is described that provides complete protection entirely by passive means for the loss of coolant accident in a nuclear power plant and wherein all stored energy released in the coolant blowdown is contained and absorbed while the nuclear fuel is continuously maintained submerged in liquid. The primary containment vessel is restored to a high subatmospheric pressure within a few minutes after accident initiation and the decay heat is safely transferred to the environment while radiolytic hydrogen is contained by passive means

  18. Reactor plant for Belene NPP completion

    International Nuclear Information System (INIS)

    Dragunov, Yu. G.; Ryzhov, S. B.; Ermakov, D. N.; Repin, A. I.

    2004-01-01

    Construction of 'Belene' NPP was started at the end of 80-ties using project U-87 with V-320 reactor plant, general designer of this plant is OKB 'Gidropress'. At the beginning of 90-ties, on completing the considerable number of deliveries and performance of civil engineering work at the site the NPP construction was suspended. Nowadays, considering the state of affairs at the site and the work performed by Bulgarian Party on preservation of the equipment delivered, the most perspective is supposed to be implementation of the following versions in completing 'Belene' NPP: for completion of Unit 1 - reactor plant VVER-1000 on the basis of V-320 reactor with the maximum use of the delivered equipment (V-320M) having the extended service life and safety improvement; for Unit 2 - advanced reactor plant VVER-1000. For the upgraded reactor plant V-230M the basic solutions and characteristics are presented, as well as the calculated justification of strength and safety analyses, design of the reactor core and fuel cycle, instrumentation and control systems, application of the 'leak-before break' in the project and implementation of safety measures. For the modernised reactor plant V-392M the main characteristics and basic changes are presented, concerning reactor pressure vessel, steam generator, reactor coolant pump set. Design of NPP with the modernized reactor plant V-320M meets the up-to-date requirements and can be licensed for completion and operation. In the design of NPP with the advanced reactor plant the basic solutions and the equipment are used that are similar to those used in standard reactor plant V-320 and new one with VVER-1000 under construction and completion in Russia, and abroad. Compliance of reactor design with the up-to-date international requirements, considering the extended service life of the main equipment, shows its rather high potential for implementation during completion of 'Belene' NPP

  19. The importance of carry out studies about the use of passive autocatalytic recombiners for hydrogen control in reactors type ESBWR; La importancia de realizar estudios sobre el uso de recombinadores autocataliticos pasivos para control de hidrogeno en reactores tipo ESBWR

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez J, J.; Morales S, J. B. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)], e-mail: jersonsanchez@gmail.com

    2009-10-15

    A way to satisfy and to guarantee the energy necessities in the future is increasing in a gradual way the creation of nuclear power plants, introducing advanced designs in its systems that contribute in way substantial in the security of the same nuclear plants. The tendency of new designs of these nuclear plants is the incorporation of systems more reliable and sure, and that the operation does not depend on external factors as the electric power, motors diesel or the action of the operator of nuclear plant, what is known as security passive systems. In this sense, the passive autocatalytic recombiners are a contribution toward the use of this type of systems. At the present time it is had studies of the incorporation of passive autocatalytic recombiners in nuclear plants in operation and that they have contributed to minimize the danger associated to hydrogen. The present work contains a first approach to the study of hydrogen recombiners incorporation in advanced nuclear plants, for this case in a nuclear power plant of ESBWR type. To achieve our objective it seeks to use specialized codes as RELAP/SCDAP to obtain simulations of passive autocatalytic recombiners behaviour and we can to estimate their operation inside the reactor contention, contemplating the possibility to use other codes like SCILAB and/or MATLAB for the simulation of a passive autocatalytic recombiner. (Author)

  20. Steady state and LOCA analysis of Kartini reactor using RELAP5/SCDAP code: The role of passive system

    Science.gov (United States)

    Antariksawan, Anhar R.; Wahyono, Puradwi I.; Taxwim

    2018-02-01

    Safety is the priority for nuclear installations, including research reactors. On the other hand, many studies have been done to validate the applicability of nuclear power plant based best estimate computer codes to the research reactor. This study aims to assess the applicability of the RELAP5/SCDAP code to Kartini research reactor. The model development, steady state and transient due to LOCA calculations have been conducted by using RELAP5/SCDAP. The calculation results are compared with available measurements data from Kartini research reactor. The results show that the RELAP5/SCDAP model steady state calculation agrees quite well with the available measurement data. While, in the case of LOCA transient simulations, the model could result in reasonable physical phenomena during the transient showing the characteristics and performances of the reactor against the LOCA transient. The role of siphon breaker hole and natural circulation in the reactor tank as passive system was important to keep reactor in safe condition. It concludes that the RELAP/SCDAP could be use as one of the tool to analyse the thermal-hydraulic safety of Kartini reactor. However, further assessment to improve the model is still needed.

  1. Passive containment cooling system performance in the simplified boiling water reactor

    International Nuclear Information System (INIS)

    Shiralkar, B.S.; Gamble, R.E.; Yadigaroglu, G.

    1997-01-01

    The Simplified Boiling Water Reactor (SBWR) incorporates a passive system for decay heat removal from the containment in the event of a postulated Loss-of-Coolant Accident (LOCA). Decay heat is removed by condensation of the steam discharged from the reactor pressure vessel (RPV) in three condensers which comprise the Passive Containment Cooling System (PCCS). These condensers are designed to carry the heat load while transporting a mixture of steam and noncondensible gas (primarily nitrogen) from the drywell to the suppression chamber. This paper describes the expected LOCA response of the SBWR with respect to the PCCS performance, based on analysis and test results. The results confirm that the PCCS has excess capacity for decay heat removal and that overall system performance is very robust. 12 refs., 8 figs

  2. ESBWR-an economical passive plant design

    International Nuclear Information System (INIS)

    Arnold, H.; Stoop, P.M.; Gonzales, A.; Rao, A.

    1996-01-01

    The ESBWR is a plant design that builds on the GKN Dodewaard natural-circulation reactor and the simplified boiling water reactor (SBWR) design. The major objective of the ESBWR program, which has been in place for the past 3 yr, is to develop a plant design with proven technology that improves the overall plant economics. It utilizes the experience and basic simplicity of the Dodewaard plant and 670-MW(electric) SBWR design features. The design is being developed by an international team of utilities, designers, and researchers. It is being designed to meet the utility and regulatory requirements of Europe. It also addresses the key economic challenges for future nuclear power stations

  3. Passive safe small reactor for distributed energy supply system sited in water filled pit at seaside

    International Nuclear Information System (INIS)

    Ishida, Toshihisa; Imayoshi, Shou

    2003-01-01

    Japan Atomic Energy Research Institute has developed a Passive Safe Small Reactor for Distributed Energy Supply System (PSRD) concept. The PSRD is an integrated-type PWR with reactor thermal power of 100 to 300 MW aimed at supplying electricity, district heating, etc. In design of the PSRD, high priority is laid on enhancement of safety as well as improvement of economy. Safety is enhanced by the following means: i) Extreme reduction of pipes penetrating the reactor vessel, by limiting to only those of the steam, the feed water and the safety valves, ii) Adoption of the water filled containment and the passive safety systems with fluid driven by natural circulation force, and iii) Adoption of the in-vessel type control rod drive mechanism, accompanying a passive reactor shut-down device. For improvement of economy, simplification of the reactor system and long operation of the core over five years without refueling with low enriched UO 2 fuel rods are achieved. To avoid releasing the radioactive materials to the circumstance even if a hypothetical accident, the containment is submerged in a pit filled with seawater at a seaside. Refueling or maintenance of the reactor can be conducted using an exclusive barge instead of the reactor building. (author)

  4. Thermal-hydraulic analysis code development and application to passive safety reactor at JAERI

    International Nuclear Information System (INIS)

    Araya, F.

    1995-01-01

    After a brief overview of safety assessment process, the author describes the LOCA analysis code system developed in JAERI. It comprises audit calculation code (WREM, WREM-J2, Japanese own code and BE codes (2D/3D, ICAP, ROSA). The codes are applied to development of Japanese passive safety reactor concept JPSR. Special attention is paid to the passive heat removal system and phenomena considered to occur under loss of heat sink event. Examples of LOCA analysis based on operation of JPSR for the cases of heat removal by upper RHR and heat removal from core to atmosphere are given. Experiments for multi-dimensional flow field in RPV and steam condensation in water pool are used for understanding the phenomena in passive safety reactors. The report is in a poster form only. 1 tab., 13 figs

  5. SWR 1000: The new boiling water reactor power plant concept

    International Nuclear Information System (INIS)

    Brettschuh, W.

    1999-01-01

    Siemens' Power Generation Group (KWU) is currently developing - on behalf of and in close co-operation with the German nuclear utilities and with support from various European partners - the boiling water reactor SWR 1000. This advanced design concept marks a new era in the successful tradition of boiling water reactor technology in Germany and is aimed, with an electric output of 1000 MW, at assuring competitive power generating costs compared to large-capacity nuclear power plants as well as coal-fired stations, while at the same time meeting the highest of safety standards, including control of a core melt accident. This objective is met by replacing active safety systems with passive safety equipment of diverse design for accident detection and control and by simplifying systems needed for normal plant operation on the basis of past operating experience. A short construction period, flexible fuel cycle lengths of between 12 and 24 months and a high fuel discharge burnup all contribute towards meeting this goal. The design concept fulfils international nuclear regulatory requirements and will reach commercial maturity by the year 2000. (author)

  6. Preservation of FFTF Data Related to Passive Safety Testing

    International Nuclear Information System (INIS)

    Wootan, David W.; Butner, R. Scott; Omberg, Ronald P.; Makenas, Bruce J.; Nielsen, Deborah L.

    2010-01-01

    One of the goals of the Fuel Cycle Research and Development Program (FCRD) is to preserve the knowledge that has been gained in the United States on Liquid Metal Reactors (LMR). A key area deserving special attention for preservation is the data relating to passive safety testing that was conducted in FFTF and EBR-II during the 1980's. Accidents at Unit 4 of the Chernobyl Station and Unit 2 at Three Mile Island changed the safety paradigm of the nuclear power industry. New emphasis was placed on assured safety based on intrinsic plant characteristics that protect not only the public, but the significant investment in the plant as well. Plants designated to perform in this manner are considered to be passively safe since no active sensor/alarm system or human intervention is required to bring the reactor to a safe shutdown condition. The liquid metal reactor (LMR) has several key characteristics needed for a passively safe reactor: reactor coolant with superior heat transfer capability and very high boiling point, low (atmospheric) system pressures, and reliable negative reactivity feedback. The credibility of the design for a passively safe LMR rests on two issues: the validity of analytic methods used to predict passive safety performance and the availability of relevant test data to calibrate design tools. Safety analysis methods used to analyze LMRs under the old safety paradigm were focused on calculating the source term for the Core Disruptive Accident. Passive safety design requires refined analysis methods for transient events because treatment of the detailed reactivity feedbacks is important in predicting the response of the reactor. Similarly, analytic tools should be calibrated against actual test experience in existing LMR facilities. The principal objectives of the combined FFTF natural circulation and Passive Safety Testing program were: (1) to verify natural circulation as a reliable means to safely remove decay heat, (2) to extend passive safety

  7. A small floating seawater desalination plant using a nuclear heating reactor coupled with the MED process

    International Nuclear Information System (INIS)

    Dong Duo; Wu Shaorong; Zhang Dafang; Wu Zongxin

    1997-01-01

    A small floating seawater desalination plant using a nuclear heating reactor coupled with a multi-effect distillation (MED) process was designed by the Institute of Nuclear Energy Technology, Tsinghua University of China. It was intended to supply potable water to remove coastal areas or islands where both fresh water and energy are severely lacking, and also to serve as a demonstration and training facility. The design of a small floating plant coupled two proven technologies in the cogeneration mode: a nuclear heating reactor (NHR-10), with inherent, passive safety features based on NHR-5 experience, and a low temperature MED process. The secondary loop was designed as a safety barrier between the primary loop and the steam loop. With a 10 MW(th) heating reactor, the floating plant could provide 4,000 m 3 /d of potable water and 750 kW of electricity. The design concept and parameters, safety features, coupling scheme and floating plant layout are presented in the paper. (author). 3 refs, 4 figs, 3 tabs

  8. Coupled analysis of passive safety injection and containment filtered venting for passive decay heat removal - 15140

    International Nuclear Information System (INIS)

    Kim, S.H.; Ham, J.H.; Jeong, Y.H.; Chang, S.H.

    2015-01-01

    Lots of interests for the safety of nuclear power plants have risen these days. The safety has to be continuously reviewed and enhanced in nuclear power plants currently operating as well as those designed and constructed in future. After the Fukushima accidents, many additional safety systems which can be applied to nuclear power plants in operation have been proposed. Those include alternating power source such as movable diesel generators and DC batteries in non-safety grade. Also, emergency preparedness for the prevention of a core damage accident was proposed to cope with the extended-SBO (station blackout) by using fire protection systems. In order to prevent the release of radioactive materials, safety systems for preserving the integrity of containment were proposed in two views of cooling and venting containment. Two approaches are effective for mitigating a severe accident. The design concept installing big water tanks besides containment at high level was proposed for various safety functions. One of the functions in the system is to inject the coolant from the elevated tank into a reactor vessel in the case of loss of coolant accident. When the pressure in reactor coolant system is sufficiently low, the coolant can be injected by gravity. If not, the depressurization in reactor vessel would be needed considering the containment pressure. Containment cooling in conventional pressurized water reactors is dependent on containment cooling pumps and sprays. Additional containment cooling systems cannot be simply and easily applied in the current nuclear power plants without major modifications. Therefore, for the operation of passive safety injection system, containment filtered venting system can be adopted for the depressurization of containment. In the design and operation of the passive safety injection system and the containment filtered venting system, main operating points related with open and close pressures in the filtered venting system were

  9. Relevance of passive safety testing at the fast flux test facility to advanced liquid metal reactors - 5127

    International Nuclear Information System (INIS)

    Wootan, D.W.; Omberg, R.P.

    2015-01-01

    Significant cost and safety improvements can be realized in advanced liquid metal reactor (LMR) designs by emphasizing inherent or passive safety through crediting the beneficial reactivity feedbacks associated with core and structural movement. This passive safety approach was adopted for the Fast Flux Test Facility (FFTF), and an experimental program was conducted to characterize the structural reactivity feedback. Testing at the Rapsodie and EBR-II reactors had demonstrated the beneficial effect of reactivity feedback caused by changes in fuel temperature and core geometry mechanisms in a liquid metal fast reactor in a holistic sense. The FFTF passive safety testing program was developed to examine how specific design elements influenced dynamic reactivity feedback in response to a reactivity input and to demonstrate the scalability of reactivity feedback results from smaller cores like Rapsodie and EBR-II to reactor cores that were more prototypic in scale to reactors of current interest. The U.S. Department of Energy, Office of Nuclear Energy Advanced Reactor Technology program is in the process of preserving, protecting, securing, and placing in electronic format information and data from the FFTF, including the core configurations and data collected during the passive safety tests. Evaluation of these actual test data could provide insight to improve analytical methods which may be used to support future licensing applications for LMRs. (authors)

  10. Experimental study of the passive flooding system in the WWER-1000 reactor

    International Nuclear Information System (INIS)

    Malyshev, A.B.; Efanov, A.D.; Kalyakin, S.G.

    2002-01-01

    The design solution of the passive flooding system in the WWER-1000 reactor core with the V-392 reactor facility and the scheme of the GE-2 large-scale thermohydraulic stand for substantiation of its functions are presented. The proposals, improving the efficiency of the system are developed on the basis of the experimental studies on the equipment input-output operational characteristics and the recommendations on the substantiation of the function of the reactor core flooding system are given [ru

  11. Passive decay heat removal from the core region

    International Nuclear Information System (INIS)

    Hichen, E.F.; Jaegers, H.

    2002-01-01

    The decay heat in commercial Light Water Reactors is commonly removed by active and redundant safety systems supported by emergency power. For advanced power plant designs passive safety systems using a natural circulation mode are proposed: several designs are discussed. New experimental data gained with the NOKO and PANDA facilities as well as operational data from the Dodewaard Nuclear Power Plant are presented and compared with new calculations by different codes. In summary, the effectiveness of these passive decay heat removal systems have been demonstrated: original geometries and materials and for the NOKO facility and the Dodewaard Reactor typical thermal-hydraulic inlet and boundary conditions have been used. With several codes a good agreement between calculations and experimental data was achieved. (author)

  12. Spanish program of advanced Nuclear Power Plants

    International Nuclear Information System (INIS)

    Marco, M.; Redon, R.

    1993-01-01

    The energy Spanish Plan is promoting some actions within the area of advanced reactors. Efforts are focussed onto the European Program of Advanced Reactors, the Program of Passive Plants (EPRI), European Fast Reactor Project and the APWR-1000 Program of INI. Electrical sector utilities and industrial partners supported by the Administration have organized an steering committee. The program of Passive Plants includes activities on Qualification, design and detailed engineering (Qualification project, SBWR project of G.E. and AP600 Project of Westinghouse. The european project on advanced plants has the following Spanish contribution: Analysis of alternative Dossier on European requisites (EUR) and Design of an European Reactor (EPR)

  13. Determination of reactor fuel burnup using passive neutron assay

    International Nuclear Information System (INIS)

    Kodeli, I.; Trkov, A.; Najzer, M.; Ertek, C.

    1988-01-01

    Passive neutron assay (PNA) method was developed to verify the fissile inventory of the irradiated reactor fuels. The characteristics of the method were studied at 'Jozef Stefan' Institute. The dependence of neutron source in the fuel on burnup, cooling time, initial enrichment and specific power were investigated and the accuracy of the method, using available computer codes was estimated. (author)

  14. Field test and evaluation of the passive neutron coincidence collar for prototype fast reactor fuel subassemblies

    International Nuclear Information System (INIS)

    Menlove, H.O.; Keddar, A.

    1982-08-01

    The passive neutron Coincidence Collar, which was developed for the verification of plutonium content in fast reactor fuel subassemblies, has been field tested using Prototype Fast Reactor fuel. For passive applications, the system measures the 240 Pu-effective mass from the spontaneous fission rate, and in addition, a self-interrogation technique is used to determine the fissile content in the subassembly. Both the passive and active modes were evaluated at the Windscale Works in the United Kingdom. The results of the tests gave a standard deviation 0.75% for the passive count and 3 to 7% for the active measurement for a 1000-s counting time. The unit will be used in the future for the verification of plutonium in fresh fuel assemblies

  15. Reliability assessment of Passive Containment Cooling System of an Advanced Reactor using APSRA methodology

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Mukesh, E-mail: mukeshd@barc.gov.in [Reactor Engineering Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Chakravarty, Aranyak [School of Nuclear Studies and Application, Jadavpur University, Kolkata 700032 (India); Nayak, A.K. [Reactor Engineering Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Prasad, Hari; Gopika, V. [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2014-10-15

    Highlights: • The paper deals with the reliability assessment of Passive Containment Cooling System of Advanced Heavy Water Reactor. • Assessment of Passive System ReliAbility (APSRA) methodology is used for reliability assessment. • Performance assessment of the PCCS is initially performed during a postulated design basis LOCA. • The parameters affecting the system performance are then identified and considered for further analysis. • The failure probabilities of the various components are assessed through a classical PSA treatment using generic data. - Abstract: Passive Systems are increasingly playing a prominent role in the advanced nuclear reactor systems and are being utilised in normal operations as well as safety systems of the reactors following an accident. The Passive Containment Cooling System (PCCS) is one of the several passive safety features in an Advanced Reactor (AHWR). In this paper, the APSRA methodology has been employed for reliability evaluation of the PCCS of AHWR. Performance assessment of the PCCS is initially performed during a postulated design basis LOCA using the best-estimate code RELAP5/Mod 3.2. The parameters affecting the system performance are then identified and considered for further analysis. Based on some pre-determined failure criterion, the failure surface for the system is predicted using the best-estimate code taking into account the deviations of the identified parameters from their nominal states as well as the model uncertainties inherent to the best estimate code. Root diagnosis is then carried out to determine the various failure causes, which occurs mainly due to malfunctioning of mechanical components. The failure probabilities of the various components are assessed through a classical PSA treatment using generic data. The reliability of the PCCS is then evaluated from the probability of availability of these components.

  16. Reliability assessment of Passive Containment Cooling System of an Advanced Reactor using APSRA methodology

    International Nuclear Information System (INIS)

    Kumar, Mukesh; Chakravarty, Aranyak; Nayak, A.K.; Prasad, Hari; Gopika, V.

    2014-01-01

    Highlights: • The paper deals with the reliability assessment of Passive Containment Cooling System of Advanced Heavy Water Reactor. • Assessment of Passive System ReliAbility (APSRA) methodology is used for reliability assessment. • Performance assessment of the PCCS is initially performed during a postulated design basis LOCA. • The parameters affecting the system performance are then identified and considered for further analysis. • The failure probabilities of the various components are assessed through a classical PSA treatment using generic data. - Abstract: Passive Systems are increasingly playing a prominent role in the advanced nuclear reactor systems and are being utilised in normal operations as well as safety systems of the reactors following an accident. The Passive Containment Cooling System (PCCS) is one of the several passive safety features in an Advanced Reactor (AHWR). In this paper, the APSRA methodology has been employed for reliability evaluation of the PCCS of AHWR. Performance assessment of the PCCS is initially performed during a postulated design basis LOCA using the best-estimate code RELAP5/Mod 3.2. The parameters affecting the system performance are then identified and considered for further analysis. Based on some pre-determined failure criterion, the failure surface for the system is predicted using the best-estimate code taking into account the deviations of the identified parameters from their nominal states as well as the model uncertainties inherent to the best estimate code. Root diagnosis is then carried out to determine the various failure causes, which occurs mainly due to malfunctioning of mechanical components. The failure probabilities of the various components are assessed through a classical PSA treatment using generic data. The reliability of the PCCS is then evaluated from the probability of availability of these components

  17. Passive containment system

    International Nuclear Information System (INIS)

    Kleimola, F.W.

    1977-01-01

    Disclosed is a containment system that provides complete protection entirely by passive means for the loss of coolant accident in a nuclear power plant and wherein all stored energy released in the coolant blowdown is contained and absorbed while the nuclear fuel is prevented from over-heating by a high containment back-pressure and a reactor vessel refill system. The primary containment vessel is restored to a high sub-atmospheric pressure within a few minutes after accident initiation and the decay heat is safely transferred to the environment while radiolytic hydrogen is contained by passive means. 20 claims, 14 figures

  18. Conceptual design of a passively safe thorium breeder Pebble Bed Reactor

    International Nuclear Information System (INIS)

    Wols, F.J.; Kloosterman, J.L.; Lathouwers, D.; Hagen, T.H.J.J. van der

    2015-01-01

    Highlights: • This work proposes three possible designs for a thorium Pebble Bed Reactor. • A high-conversion PBR (CR > 0.96), passively safe and within practical constraints. • A thorium breeder PBR (220 cm core) in practical regime, but not passively safe. • A passively safe breeder, requiring higher fuel reprocessing and recycling rates. - Abstract: More sustainable nuclear power generation might be achieved by combining the passive safety and high temperature applications of the Pebble Bed Reactor (PBR) design with the resource availability and favourable waste characteristics of the thorium fuel cycle. It has already been known that breeding can be achieved with the thorium fuel cycle inside a Pebble Bed Reactor if reprocessing is performed. This is also demonstrated in this work for a cylindrical core with a central driver zone, with 3 g heavy metal pebbles for enhanced fission, surrounded by a breeder zone containing 30 g thorium pebbles, for enhanced conversion. The main question of the present work is whether it is also possible to combine passive safety and breeding, within a practical operating regime, inside a thorium Pebble Bed Reactor. Therefore, the influence of several fuel design, core design and operational parameters upon the conversion ratio and passive safety is evaluated. A Depressurized Loss of Forced Cooling (DLOFC) is considered the worst safety scenario that can occur within a PBR. So, the response to a DLOFC with and without scram is evaluated for several breeder PBR designs using a coupled DALTON/THERMIX code scheme. With scram it is purely a heat transfer problem (THERMIX) demonstrating the decay heat removal capability of the design. In case control rods cannot be inserted, the temperature feedback of the core should also be able to counterbalance the reactivity insertion by the decaying xenon without fuel temperatures exceeding 1600 °C. Results show that high conversion ratios (CR > 0.96) and passive safety can be combined in

  19. Research and development on next generation reactor (phase I)

    International Nuclear Information System (INIS)

    Park, Jong Kyoon; Chang, Moon Heuy; Hwang, Yung Dong

    1994-10-01

    The objective of the study is to improve the volume of nuclear power plant which adopts passive safety system concept. The passive safety system reactor is characterized by excellent safety and reliability. But the volume of NSSS (Nuclear Steam Supply System) of the passive safety system reactor is so small that it should be upgraded for commercial operation. For volume upgrade, detailed analyses are performed as follows; core design, hydraulics, design and mechnical structures, and safety analysis. In addition to above analysis, some investigations must be supplied as follows: power density vs. DNB margin decrease, outlet temperature vs. EPRI-URD, additional tests for upgraded reactor, dynamic analysis of mechanical vibration according to expanded reactor vessel and expanded in-core structures, and Merit loss of passive safety system reactor according to design margin decrease. (Author)

  20. Research and development on next generation reactor (phase I)

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Kyoon; Chang, Moon Heuy; Hwang, Yung Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); and others

    1994-10-01

    The objective of the study is to improve the volume of nuclear power plant which adopts passive safety system concept. The passive safety system reactor is characterized by excellent safety and reliability. But the volume of NSSS (Nuclear Steam Supply System) of the passive safety system reactor is so small that it should be upgraded for commercial operation. For volume upgrade, detailed analyses are performed as follows; core design, hydraulics, design and mechnical structures, and safety analysis. In addition to above analysis, some investigations must be supplied as follows: power density vs. DNB margin decrease, outlet temperature vs. EPRI-URD, additional tests for upgraded reactor, dynamic analysis of mechanical vibration according to expanded reactor vessel and expanded in-core structures, and Merit loss of passive safety system reactor according to design margin decrease. (Author).

  1. Implications of passive safety based on historical industrial experience

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    1988-01-01

    In the past decade, there have been multiple proposals for applying different technologies to achieve passively safe light water reactors (LWRs). A key question for all such concepts is, ''What are the gains in safety, costs, and reliability for passive safety systems.'' Using several types of historical data, estimates have been made of gains from passive safety and operating systems, which are independent of technology. Proposals for passive safety in reactors usually have three characteristics: (1) Passive systems with no moving mechanical parts, (2) systems with far fewer components and (3) more stringent design criteria for safety-related and process systems. Each characteristic reduces the potential for an accident and may increase plant reliability. This paper addresses gains from items (1) and (2). Passive systems often allow adoption of more rigorous design criteria which would be either impossible or economically unfeasible for active systems. This important characteristic of passive safety systems cannot be easily addressed using historical industrial experience

  2. Deep underground reactor (passive heat removal of LWR with hard neutron energy spectrum)

    Energy Technology Data Exchange (ETDEWEB)

    Hiroshi, Takahashi [Brookhaven National Lab., Upton, NY (United States)

    2001-07-01

    To run a high conversion reactor with Pu-Th fueled tight fueled assembly which has a long burn-up of a fuel, the reactor should be sited deep underground. By putting the reactor deep underground heat can be removed passively not only during a steady-state run and also in an emergency case of loss of coolant and loss of on-site power; hence the safety of the reactor can be much improved. Also, the evacuation area around the reactor can be minimized, and the reactor placed near the consumer area. This approach reduces the cost of generating electricity by eliminating the container building and shortening transmission lines. (author)

  3. Deep underground reactor (passive heat removal of LWR with hard neutron energy spectrum)

    International Nuclear Information System (INIS)

    Hiroshi, Takahashi

    2001-01-01

    To run a high conversion reactor with Pu-Th fueled tight fueled assembly which has a long burn-up of a fuel, the reactor should be sited deep underground. By putting the reactor deep underground heat can be removed passively not only during a steady-state run and also in an emergency case of loss of coolant and loss of on-site power; hence the safety of the reactor can be much improved. Also, the evacuation area around the reactor can be minimized, and the reactor placed near the consumer area. This approach reduces the cost of generating electricity by eliminating the container building and shortening transmission lines. (author)

  4. Modular nuclear reactor for a land-based power plant and method for the fabrication installation and operation thereof

    International Nuclear Information System (INIS)

    Craig, E. R.; Blumberg, B. Jr.

    1985-01-01

    A self-contained modular nuclear reactor which can be prefabricated at a factory location, nuclear-certified at the factory, transported to a field location for final assembly and connection to a large-scale electric-power generating facility. The modular reactor includes a prefabricated nuclear heat supply module and a plurality of shell segments which can be assembled about the heat supply module and which provide a form for the pouring and curing of a cementatious biological shield about the heat supply module. The modular reactor includes passive shutdown heat removal systems sufficient to render the reactor safe in an emergency. A large-scale power plant arrangement is disclosed which incorporates a plurality of the modular reactors

  5. Reliability assurance for regulation of advanced reactors

    International Nuclear Information System (INIS)

    Fullwood, R.; Lofaro, R.; Samanta, P.

    1992-01-01

    The advanced nuclear power plants must achieve higher levels of safety than the first generation of plants. Showing that this is indeed true provides new challenges to reliability and risk assessment methods in the analysis of the designs employing passive and semi-passive protection. Reliability assurance of the advanced reactor systems is important for determining the safety of the design and for determining the plant operability. Safety is the primary concern, but operability is considered indicative of good and safe operation. this paper discusses several concerns for reliability assurance of the advanced design encompassing reliability determination, level of detail required in advanced reactor submittals, data for reliability assurance, systems interactions and common cause effects, passive component reliability, PRA-based configuration control system, and inspection, training, maintenance and test requirements. Suggested approaches are provided for addressing each of these topics

  6. Reliability assurance for regulation of advanced reactors

    International Nuclear Information System (INIS)

    Fullwood, R.; Lofaro, R.; Samanta, P.

    1991-01-01

    The advanced nuclear power plants must achieve higher levels of safety than the first generation of plants. Showing that this is indeed true provides new challenges to reliability and risk assessment methods in the analysis of the designs employing passive and semi-passive protection. Reliability assurance of the advanced reactor systems is important for determining the safety of the design and for determining the plant operability. Safety is the primary concern, but operability is considered indicative of good and safe operation. This paper discusses several concerns for reliability assurance of the advanced design encompassing reliability determination, level of detail required in advanced reactor submittals, data for reliability assurance, systems interactions and common cause effects, passive component reliability, PRA-based configuration control system, and inspection, training, maintenance and test requirements. Suggested approaches are provided for addressing each of these topics

  7. Advances in passive cooling design and performance analysis

    International Nuclear Information System (INIS)

    Woodcock, J.

    1994-01-01

    The Third International Conference on Containment Design and Operation continues the trend of rapidly extending the state of the art in containment methodology, joining other conferences, OECD-sponsored International Standard Problem exercises, and vendor licensing submittals. Methodology developed for use on plants with passive features is under increasing scrutiny for advanced designs, since the passive features are often the only deviation from existing operating base of the past 30 years of commercial nuclear power. This session, 'Containment Passive Safety Systems Design and Operation,' offers papers on a wide range of topics, with authors from six organizations from around the world, dealing with general passive containments, Westinghouse AP600, large (>1400 MWe) passive plants, and the AECL advanced CANDU reactor. This level and variety of participation underscores the high interest and accelerated methods development associated with advanced passive containment heat removal. The papers presented in this session demonstrate that significant contributions are being made to the advancement of technology necessary for building a new generation of safer, more economical nuclear plants. (author)

  8. Preliminary assessment of a combined passive safety system for typical 3-loop PWR CPR1000

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Zijiang; Shan, Jianqiang, E-mail: jqshan@mail.xjtu.edu.cn; Gou, Junli

    2017-03-15

    Highlights: • A combined passive safety system was placed on a typical 3-loop PWR CPR1000. • Three accident analyses show the three different accident mitigation methods of the passive safety system. • The three mitigation methods were proved to be useful. - Abstract: As the development of the nuclear industry, passive technology turns out to be a remarkable characteristic of advanced nuclear power plants. Since the 20th century, much effort has been given to the passive technology, and a number of evolutionary passive systems have developed. Thoughts have been given to upgrade the existing reactors with passive systems to meet stricter safety demands. In this paper, the CPR1000 plant, which is one kind of mature pressurized water reactor plants in China, is improved with some passive systems to enhance safety. The passive systems selected are as follows: (1) the reactor makeup tank (RMT); (2) the advanced accumulator (A-ACC); (3) the in-containment refueling water storage tank (IRWST); (4) the passive emergency feed water system (PEFS), which is installed on the secondary side of SGs; (5) the passive depressurization system (PDS). Although these passive components is based on the passive technology of some advanced reactors, their structural and trip designs are adjusted specifically so that it could be able to mitigate accidents of the CPR1000. Utilizing the RELAP5/MOD3.3 code, accident analyses (small break loss of coolant accident, large break loss of coolant accident, main feed water line break accident) of this improved CPR1000 plant were presented to demonstrate three different accident mitigation methods of the safety system and to test whether the passive safety system preformed its function well. In the SBLOCA, all components of the passive safety system were put into work sequentially, which prevented the core uncover. The LBLOCA analysis illustrates the contribution of the A-ACCs whose small-flow-rate injection can control the maximum cladding

  9. Flooding of a large, passive, pressure-tube light water reactor

    International Nuclear Information System (INIS)

    Hejzlar, P.; Todreas, N.E.; Driscoll, M.J.

    1997-01-01

    A reactor concept has been developed which can survive loss of coolant accidents without scram and without replenishing primary coolant inventory, while maintaining safe temperature limits on the fuel and pressure tubes. The proposed concept is a pressure tube type reactor of similar design to CANDU reactors, but differing in three key aspects. First, a solid SiC-coated graphite fuel matrix is used in place of fuel pin bundles to enable the dissipation of decay heat from the fuel in the absence of primary coolant. Second, the heavy water coolant in the pressure tubes is replaced by light water, which also serves as the moderator. Finally, the calandria tank, surrounded by a graphite reflector, contains a low pressure gas instead of heavy water moderator, and this normally-voided calandria is connected to a light water heat sink. The cover gas displaces the light water from the calandria during normal operation, while during loss of coolant or loss of heat sink accidents it allows passive calandria flooding. Calandria flooding also provides redundant and diverse reactor shutdown. This paper describes the thermal hydraulic characteristics of the passively initiated, gravity driven calandria flooding process. Flooding the calandria space with light water is a unique and very important feature of the proposed pressure-tube light water reactor (PTLWR) concept. The flooding of the top row of fuel channels must be accomplished fast enough so that in the total loss of coolant, none of the critical components of the fuel channel, i.e. the pressure tube, the calandria tube, the matrix and the fuel, exceed their design limits. The flooding process has been modeled and shown to be rapid enough to maintain all components within their design limits. (orig.)

  10. Plants for passive cooling. A preliminary investigation of the use of plants for passive cooling in temperate humid climates

    Energy Technology Data Exchange (ETDEWEB)

    Spirn, A W; Santos, A N; Johnson, D A; Harder, L B; Rios, M W

    1981-04-01

    The potential of vegetation for cooling small, detached residential and commercial structures in temperate, humid climates is discussed. The results of the research are documented, a critical review of the literature is given, and a brief review of energy transfer processes is presented. A checklist of design objectives for passive cooling, a demonstration of design applications, and a palette of selected plant species suitable for passive cooling are included.

  11. ESBWR-An economical passive plant design

    International Nuclear Information System (INIS)

    Gonzalez Lopez, A.; Rao, A.

    1996-01-01

    This paper provides an overview of the design features of the European Simplified Boiling Water Reactor (ESBWR) design. The ESBWR is a plant design that builds on the Simplified Boiling Water Reactor (SBWR) design described in Reference 1 and 2. The major objective of the ESBWR programme is to develop a plant design that utilizes the basic simplicity of the SBWR design features to improve overall economics as discussed in Reference 3. The design is being developed by an international team of utilities, designers and researchers, with the objective of applying it to the utility and regulatory requirements of Europe. (Author)

  12. Study on plant concept for gas cooled fast reactor

    International Nuclear Information System (INIS)

    Moribe, Takeshi; Kubo, Shigenobu; Saigusa, Toshiie; Konomura, Mamoru

    2003-05-01

    In 'Feasibility Study on Commercialized Fast Reactor Cycle System', technological options including various coolant (sodium, heavy metal, gas, water, etc.), fuel type (MOX, metal, nitride) and output power are considered and classified, and commercialized FBR that have economical cost equal to LWR are pursued. In conceptual study on gas cooled FBR in FY 2002, to identify the prospect of the technical materialization of the helium cooled FBR using coated particle fuel which is an attractive concept extracted in the year of FY2001, the preliminary conceptual design of the core and entire plant was performed. This report summarizes the results of the plant design study in FY2002. The results of study is as follows. 1) For the passive core shutdown equipment, the curie point magnet type self-actuated device was selected and the device concept was set up. 2) For the reactor block, the concept of the core supporting structure, insulators and liners was set up. For the material of the heat resistant structure, SiC was selected as a candidate. 3) For the seismic design of the plant, it was identified that a design concept with three-dimensional base isolation could be feasible taking the severe seismic condition into account. 4) For the core catcher, an estimation of possible event sequences under severe core damage condition was made. A core catcher concept which may suit the estimation was proposed. 5) The construction cost was roughly estimated based on the amount of materials and its dependency on the plant output power was evaluated. The value for a small sized plant exceeds the target construction cost about 20%. (author)

  13. Engineered safeguards and passive safety features (safety analysis detailed report no. 6)

    Energy Technology Data Exchange (ETDEWEB)

    1988-01-15

    The Safety-Analysis Summary lists the reactor's safety aspects for passive and active prevention of severe accidents and mitigation of accident consequences, i.e., intrinsic and passive protections of the plant; intrinsic and passive protections of the core; inherent decay-heat removal systems; rapid-shutdown systems; four physical containment barriers. This report goes into further details regarding some of this aspects.

  14. Study for Safeguards Challenges to the Most Probably First Indonesian Future Power Plant of the Pebble Bed Modular Reactor

    International Nuclear Information System (INIS)

    Susilowati, E.

    2015-01-01

    In the near future Indonesia, the fourth most populous country, plans to build a small size power plant most probably a Pebble Bed Modular Reactor PBMR. This first nuclear power plant (NPP) is aimed to provide clear picture to the society in regard to performance and safety of nuclear power plant operation. Selection to the PBMR based on several factor including the combination of small size of the reactor and type of fuel allowing the use of passive safety systems, resulting in essential advantages in nuclear plant design and less dependence on plant operators for safety. In the light of safeguards perspective this typical reactor is also quite difference with previous light water reactor (LWR) design. From the fact that there are a small size large number of elements present in the reactor produced without individual serial numbers combine to on-line refueling same as the CANDU reactor, enforcing a new challenge to safeguards approach for this typical reactor. This paper discusses a bunch of safeguards measures have to be prepared by facility operator to support successfully international nuclear material and facility verification including elements of design relevant to safeguards need to be accomplished in consultation to the regulatory body, supplier or designer and the Agency/IAEA such as nuclear material balance area and key measurement point; possible diversion scenarios and safeguards strategy; and design features relevant to the IAEA equipment have to be installed at the reactor facility. It is deemed that result of discussion will alleviate and support the Agency approaching safeguards measure that may be applied to the purpose Indonesian first power plant of PBMR construction and operation. (author)

  15. Passive Decay Heat Removal System Options for S-CO2 Cooled Micro Modular Reactor

    International Nuclear Information System (INIS)

    Moon, Jangsik; Jeong, Yong Hoon; Lee, Jeong Ik

    2014-01-01

    To achieve modularization of whole reactor system, Micro Modular Reactor (MMR) which has been being developed in KAIST took S-CO 2 Brayton power cycle. The S-CO 2 power cycle is suitable for SMR due to high cycle efficiency, simple layout, small turbine and small heat exchanger. These characteristics of S-CO 2 power cycle enable modular reactor system and make reduced system size. The reduced size and modular system motived MMR to have mobility by large trailer. Due to minimized on-site construction by modular system, MMR can be deployed in any electricity demand, even in isolated area. To achieve the objective, fully passive safety systems of MMR were designed to have high reliability when any offsite power is unavailable. In this research, the basic concept about MMR and Passive Decay Heat Removal (PDHR) system options for MMR are presented. LOCA, LOFA, LOHS and SBO are considered as DBAs of MMR. To cope with the DBAs, passive decay heat removal system is designed. Water cooled PDHR system shows simple layout, but has CCF with reactor systems and cannot cover all DBAs. On the other hand, air cooled PDHR system with two-phase closed thermosyphon shows high reliability due to minimized CCF and is able to cope with all DBAs. Therefore, the PDHR system of MMR will follows the air-cooled PDHR system and the air cooled system will be explored

  16. 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)

  17. The passive response of the Integral Fast Reactor concept to the chilled inlet accident

    International Nuclear Information System (INIS)

    Vilim, R.B.

    1990-01-01

    Simple methods are described for bounding the passive response of a metal fueled liquid-metal cooled reactor to the chilled inlet accident. Calculation of these bounds for a prototype of the Integral Fast Reactor concept shows that failure limits --- eutectic melting, sodium boiling and fuel pin failure --- are not exceeded. 2 refs., 1 fig., 2 tabs

  18. Technological status of reactor coolant pumps in generation III+ pressurized nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Brecht, Bernhard; Bross, Stephan [KSB Aktiengesellschaft, Frankenthal (Germany)

    2016-05-15

    KSB has been developing and producing pumps for thermal power plants for nearly 90 years. Consequently, KSB also started to develop and manufacture pumps for all kinds of nuclear power plants from the very beginning of the civil use of nuclear energy. This is especially true for reactor coolant pumps for pressurized water reactors. For the generation of advanced evolutionary reactors (Generation III+ reactors), KSB developed an advanced shaft seal system which is also able to fulfill the requirements of station blackout conditions. The tests in the KSB test rigs, which were successfully completed in December 2015, proved the full functionality of the new design. For generation III+ passive plant reactors KSB developed a new reactor coolant pump type called RUV, which is based on the experience of classic reactor coolant pumps and reactor internal pumps. It is a very compact, hermetically sealed vertical pump-motor unit with a wet winding motor. A full scale prototype successfully passed the 1st stage qualification test program in October 2015.

  19. Analyses for passive safety of fusion reactor during ex-vessel loss of coolant accident

    International Nuclear Information System (INIS)

    Honda, Takuro; Okazaki, Takashi; Maki, Koichi; Uda, Tatuhiko; Seki, Yasushi; Aoki, Isao; Kunugi, Tomoaki.

    1995-01-01

    Passive safety of nuclear fusion reactors during ex-vessel Loss-of-Coolant Accidents (LOCAs) in the divertor cooling system has been investigated using a hybrid code, which can treat the interaction of the plasma and plasma facing components (PFCs). The code has been modified to include the impurity emission from PFCs with a diffusion model at the edge plasma. We assumed an ex-vessel LOCA of the divertor cooling system during the ignited operation in International Thermonuclear Experimental Reactor (ITER), in which a carbon-copper brazed divertor plate was employed in the Conceptual Design Activity (CDA). When a double-ended break occurs at the cold leg of the divertor cooling system, the impurity density in the main plasma becomes about twice within 2s after the LOCA due to radiation enhanced sublimation of graphite PFCs. The copper cooling tube of the divertor begins to melt at about 3s after the LOCA, even though the plasma is passively shut down at about 4s due to the impurity accumulation. It is necessary to apply other PFC materials, which can shorten the time period for passive shutdown, or an active shutdown system to keep the reactor structures intact for such rapid transient accident. (author)

  20. Passive deca-heat removal in the fixed bed nuclear reactor (FBNR) - 15551

    International Nuclear Information System (INIS)

    Solano Diaz, E.C.; Luna Aguilera, G.M.; Santos, R.A.; Vaca, D.E.

    2015-01-01

    The Fixed Bed Nuclear Reactor (FBNR) is a Generation IV small reactor concept, where the spherical elements contain Triso-type microspheres with UO 2 , which serves as nuclear fuel. In the event that adverse operation conditions occur, the water pump is automatically shut off and the fuel pebbles fall back by gravity into the fuel chamber. Since the FBNR relies on passive security systems, the removal of the decay heat in the fuel chamber is achieved by contact with quiescent water. In the present paper, a mathematical simulation of the passive cooling of the system was conducted in SOLIDWORKS so as to obtain a temperature profile in the body during the decay heat removal process. Homogenization techniques were employed to smooth out spatial variations across the multiphase system and to derive expression for the effective thermophysical properties that are valid through the macroscopic entry (the chamber). The simulation showed that the chamber's temperature rose from 573 K to its maximum temperature, 1234 K, in the first hour. Afterwards, the temperature fluctuated, but stayed under 552 K. Since the temperature of the system was always kept under the value of the safety parameter (1200 C. degrees) the simulation confirmed that an effective passive cooling of the fuel chamber is indeed feasible. (authors)

  1. Thermal limits for passive safety of fusion reactors

    International Nuclear Information System (INIS)

    Kazimi, M.S.; Massidda, J.E.; Oshima, M.

    1989-01-01

    The thermal response of the first wall and blanket due to power/cooling mismatch in the absence of operation action is examined. The analyses of coolant and power transients are carried out on six reference blanket designs representing a broad range of fusion first wall and blanket technology. It is concluded that the requirement of plant protection will impose sufficiently stringent peak neutron wall loading limits to avoid a serious threat to the public. It is found that for the D-T design,s the operating wall loading may have to be limited to 3 - 8 MW/m/sup 2/ for passive plant protection, depending on the plant design

  2. Passive cooling of condensate chambers as retrofitting measure in boiling water reactors; Passive Kuehlung der Kondensationskammern in Siedewasserreaktoren als Nachruestmassnahme

    Energy Technology Data Exchange (ETDEWEB)

    Freis, Daniel; Nachtrodt, Frederik; Sporn, Michael; Tietsch, Wolfgang; Sassen, Felix [Westinghouse Electric Germany GmbH, Mannheim (Germany)

    2012-11-01

    Westinghouse Electric Germany GmbH has developed a concept for passive cooling of condensate chambers of BWR-type reactors. Due to its compactness the system is feasible as retrofitting measure. The passive condensate chamber cooling system is based on a cooling module with ascending and down pipe that are connected with the evaporation condenser to form a cooling circuit. Based on the consequent use of high-effective heat transport mechanisms, as boiling, condensation without non-condensable gases and mass transport a high cooling performance and compact construction is possible. The system is completely passive and completely diverse to existing active cooling systems. In the frame of a true-scale experiment the significant cooling performance was demonstrated. RELAP5 calculations confirmed the functionality of the cooling module.

  3. Westinghouse Small Modular Reactor (SMR) Programe

    International Nuclear Information System (INIS)

    Shulyak, Nick

    2014-01-01

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (> 225 MWe) integral pressurized water reactor (iPWR) in which all primarycomponents associated with the nuclear steam supply system, including the steam generator and the pressurizer, are housed within the reactor vessel. The Westinghouse SMR utilizes passive safety systems and proven components from the AP1000 plant design with a compact containment that houses the integral reactor vessel and the passive safety systems. This paper describes the design and functionality of the Westinghouse SMR, the key drivers influencing the design of the Westinghouse SMR and the unique passive safety features of the Westinghouse SMR. Several critical motivators contributed to the development and integration of the Westinghouse SMR design. These design drivers include safety, economics, reactor expertise and experience, research and development requirements, functionality of systems and components, size of the systems and vessels, simplicity of design, and licensing requirements. The Westinghouse SMR safety system design is passive, is based largely on the passive safety systems used in the AP1000 reactor, and provides mitigation of all design basis accidents without the need for offsite AC electrical power for a period of seven days. The economics of the Westinghouse SMR challenges the established approach of large Light Water Reactors (LWR) that utilized the economies of scale to reach economic competiveness. To serve the market expectation of smaller capital investment and cost competitive energy, a modular design approach is implemented within the Westinghouse SMR. The Westinghouse SMR building layout integrates the three basic design constraints of modularization; transportation, handling and module-joining technology. The integral Westinghouse SMR design eliminates large loop piping, which significantly reduces the flow area of postulated loss of coolant accidents (LOCAs). The Westinghouse SMR containment is a high

  4. Survey and evaluation of inherent safety characteristics and passive safety systems for use in probabilistic safety analyses

    International Nuclear Information System (INIS)

    Wetzel, N.; Scharfe, A.

    1998-01-01

    The present report examines the possibilities and limits of a probabilistic safety analysis to evaluate passive safety systems and inherent safety characteristics. The inherent safety characteristics are based on physical principles, that together with the safety system lead to no damage. A probabilistic evaluation of the inherent safety characteristic is not made. An inventory of passive safety systems of accomplished nuclear power plant types in the Federal Republic of Germany was drawn up. The evaluation of the passive safety system in the analysis of the accomplished nuclear power plant types was examined. The analysis showed that the passive manner of working was always assumed to be successful. A probabilistic evaluation was not performed. The unavailability of the passive safety system was determined by the failure of active components which are necessary in order to activate the passive safety system. To evaluate the passive safety features in new concepts of nuclear power plants the AP600 from Westinghouse, the SBWR from General Electric and the SWR 600 from Siemens, were selected. Under these three reactor concepts, the SWR 600 is specially attractive because the safety features need no energy sources and instrumentation in this concept. First approaches for the assessment of the reliability of passively operating systems are summarized. Generally it can be established that the core melt frequency for the passive concepts AP600 and SBWR is advantageous in comparison to the probabilistic objectives from the European Pressurized Water Reactor (EPR). Under the passive concepts is the SWR 600 particularly interesting. In this concept the passive systems need no energy sources and instrumentation, and has active operational systems and active safety equipment. Siemens argues that with this concept the frequency of a core melt will be two orders of magnitude lower than for the conventional reactors. (orig.) [de

  5. Passive safety design characteristics of the KALIMER-600 burner reactor

    International Nuclear Information System (INIS)

    Kwon, Young-Min; Jeong, Hae-Yong; Cho, Chung-Ho; Ha, Ki-Seok; Kim, Sang-Ji

    2009-01-01

    The Korea Atomic Energy Research Institute (KAERI) has recently studied several burner core designs for a transuranics (TRU) transmutation based on the breakeven core geometry of KALIMER-600. The KALIMER-600 is a net electrical rating of 600MWe, sodium-cooled, metallic-fueled, pool-type reactor. For the burner core concept selected for the present analysis, the smearing fractions of the fuel rods in three fuel zones are changed while maintaining the cladding outer diameter and cladding thickness. The resulting fuel slug smearing fractions of the inner, middle, and outer core zones are 36%, 40%, and 48%, respectively. The TRU conversion ratio is 0.57 and the TRU enrichment of the driver fuel is set to 30.0 w/o because of the current practical limitation of the U-TRU-10%Zr metal fuel database. The purpose of this paper is to evaluate the safety performance characteristics provided by the passive safety design features in the KALIMER-600 burner reactor by using a system-wide safety analysis code. The present scoping analysis focuses on an assessment of the enhanced safety design features that provide passive and self-regulating responses to transient conditions and an evaluation of the safety margin during unprotected overpower, unprotected loss of flow, and unprotected loss of heat sink events. The analysis results show that the KALIMER-600 burner reactor provides larger safety margins with respect to the sodium boiling, fuel rod integrity, and structural integrity. The overall inherent safety can be enhanced by accounting for the reactivity feedback mechanisms in the design process. (author)

  6. The european passive plant (EPP) design: compliance with the european utilities requirements (EUR)

    International Nuclear Information System (INIS)

    Noviello, L.; Oyarzabal, M.

    1996-01-01

    Back 1986, most of the European firms have participated to the American program called the Advanced Light Water Reactors (ALWR) including the development of the Utilities Requirements as well as four projects as for instance AP600. Later, in the year 1990, seven European firms have begun to develop the European Utilities Requirements. This development is justified by the fact that the lessons learned by the nuclear power plants designs programs of the years 1980 can be incorporated and the European specific conditions can be taken into consideration. Thus, in 1994, eight European firms - Westinghouse and their industrial partners - have decided to launch a multiphase program in order to check the AP600 compliance with the European Utilities Requirements (EUR) and to develop the required alterations. Today, the phase I of the EPP (European Passive Plant) program has been completed. In this phase, the main important objectives have been reached. (O.M.)

  7. NRC review of passive reactor design certification testing programs: Overview and regulatory perspective

    International Nuclear Information System (INIS)

    Levin, A.E.

    1993-01-01

    Reactor vendors are developing new designs for future deployment, including open-quotes passiveclose quotes light water reactors (LWRs), such as General Electric's (G.E.'s) simplified boiling water reactor (SBWR) and Westinghouse's AP600, which depend primarily on inherent processes, such as national convection and gravity feed, for safety injection and emergency core cooling. The U.S. Nuclear Regulatory Commission (NRC) has implemented a new process, certification of standardized reactor designs, for licensing these Plants. Part 52 of Title 10 of the Code of Federal Regulations (10CFR52) contains the requirements that vendors must meet for design certification. One important section, 10CFR52.47, reads open-quotes Certification of a standard design which . . . utilizes simplified, inherent, passive, or other innovative means to accomplish its safety functions will be granted only if: (1) The performance of each safety feature of the design has been demonstrated through either analysis, appropriate test programs, experience, or a combination thereof; (2) Interdependent effects among the safety features have been found acceptable by analysis, appropriate test programs, experience, or a combination thereof; and (3) Sufficient data exist on the safety features of the design to assess the analytical tools used for safety analyses. . . . The vendors have initiated programs to test innovative features of their designs and to develop data bases needed to validate their analytical codes, as required by the design certification rule. Accordingly, the NRC is reviewing and evaluating the vendors programs to ensure that they address adequately key issues concerning safety system performance. This paper provides an overview of the NRC's review process and regulatory perspective

  8. Plant control impact on IFR power plant passive safety response

    International Nuclear Information System (INIS)

    Vilim, R.B.

    1993-01-01

    A method is described for optimizing the closed-loop plant control strategy with respect to safety margins sustained in the unprotected upset response of a liquid metal reactor. The optimization is performed subject to the normal requirements for reactor startup, load change and compensation for reactivity changes over the cycle. The method provides a formal approach to the process of exploiting the innate self-regulating property of a metal fueled reactor to make it less dependent on operator action and less vulnerable to automatic control system fault and/or operator error

  9. Westinghouse Small Modular Reactor balance of plant and supporting systems design

    Energy Technology Data Exchange (ETDEWEB)

    Memmott, M. J.; Stansbury, C.; Taylor, C. [Westinghouse Electric Company LLC, 600 Cranberry Woods Drive, Cranberry Twp. PA 16066 (United States)

    2012-07-01

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the second in a series of four papers which describe the design and functionality of the Westinghouse SMR. It focuses, in particular, upon the supporting systems and the balance of plant (BOP) designs of the Westinghouse SMR. Several Westinghouse SMR systems are classified as safety, and are critical to the safe operation of the Westinghouse SMR. These include the protection and monitoring system (PMS), the passive core cooling system (PXS), and the spent fuel cooling system (SFS) including pools, valves, and piping. The Westinghouse SMR safety related systems include the instrumentation and controls (I and C) as well as redundant and physically separated safety trains with batteries, electrical systems, and switch gears. Several other incorporated systems are non-safety related, but provide functions for plant operations including defense-in-depth functions. These include the chemical volume control system (CVS), heating, ventilation and cooling (HVAC) systems, component cooling water system (CCS), normal residual heat removal system (RNS) and service water system (SWS). The integrated performance of the safety-related and non-safety related systems ensures the safe and efficient operation of the Westinghouse SMR through various conditions and transients. The turbine island consists of the turbine, electric generator, feedwater and steam systems, moisture separation systems, and the condensers. The BOP is designed to minimize assembly time, shipping challenges, and on-site testing requirements for all structures, systems, and components. (authors)

  10. Westinghouse Small Modular Reactor balance of plant and supporting systems design

    International Nuclear Information System (INIS)

    Memmott, M. J.; Stansbury, C.; Taylor, C.

    2012-01-01

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the second in a series of four papers which describe the design and functionality of the Westinghouse SMR. It focuses, in particular, upon the supporting systems and the balance of plant (BOP) designs of the Westinghouse SMR. Several Westinghouse SMR systems are classified as safety, and are critical to the safe operation of the Westinghouse SMR. These include the protection and monitoring system (PMS), the passive core cooling system (PXS), and the spent fuel cooling system (SFS) including pools, valves, and piping. The Westinghouse SMR safety related systems include the instrumentation and controls (I and C) as well as redundant and physically separated safety trains with batteries, electrical systems, and switch gears. Several other incorporated systems are non-safety related, but provide functions for plant operations including defense-in-depth functions. These include the chemical volume control system (CVS), heating, ventilation and cooling (HVAC) systems, component cooling water system (CCS), normal residual heat removal system (RNS) and service water system (SWS). The integrated performance of the safety-related and non-safety related systems ensures the safe and efficient operation of the Westinghouse SMR through various conditions and transients. The turbine island consists of the turbine, electric generator, feedwater and steam systems, moisture separation systems, and the condensers. The BOP is designed to minimize assembly time, shipping challenges, and on-site testing requirements for all structures, systems, and components. (authors)

  11. Natural Circulation Phenomena and Modelling for Advanced Water Cooled Reactors

    International Nuclear Information System (INIS)

    2012-03-01

    The role of natural circulation in advanced water cooled reactor design has been extended with the adoption of passive safety systems. Some designs utilize natural circulation to remove core heat during normal operation. Most passive safety systems used in evolutionary and innovative water cooled reactor designs are driven by natural circulation. The use of passive systems based on natural circulation can eliminate the costs associated with the installation, maintenance and operation of active systems that require multiple pumps with independent and redundant electric power supplies. However, considering the weak driving forces of passive systems based on natural circulation, careful design and analysis methods must be employed to ensure that the systems perform their intended functions. Several IAEA Member States with advanced reactor development programmes are actively conducting investigations of natural circulation to support the development of advanced water cooled reactor designs with passive safety systems. To foster international collaboration on the enabling technology of passive systems that utilize natural circulation, in 2004 the IAEA initiated a coordinated research project (CRP) on Natural Circulation Phenomena, Modelling and Reliability of Passive Systems that Utilize Natural Circulation. Three reports were published within the framework of this CRP. The first report (IAEA-TECDOC-1474) contains the material developed for the first IAEA training course on natural circulation in water cooled nuclear power plants. The second report (IAEA-TECDOC-1624) describes passive safety systems in a wide range of advanced water cooled nuclear power plant designs, with the goal of gaining insights into system design, operation and reliability. This third, and last, report summarizes the research studies completed by participating institutes during the CRP period.

  12. Comprehensive safety analysis code system for nuclear fusion reactors III: Ex-vessel LOCA analyses considering passive safety

    International Nuclear Information System (INIS)

    Honda, T.; Okazaki, T.; Maki, K.; Uda, T.; Seki, Y.; Aoki, I.; Kunugi, T.

    1996-01-01

    Ex-vessel loss-of-coolant accidents (LOCAs) in a fusion reactor have been analyzed to investigate the possibility of passive plasma shutdown. For this purpose, a hybrid code of the plasma dynamics and thermal characteristics of the reactor structures, which has been modified to include the impurity emission from plasma-facing components (PFCs), has been developed. Ex-vessel LOCAs of the cooling system during the ignition operation in the International Thermonuclear Experimental Reactor (ITER), in which graphite PFCs were employed in conceptual design activity, were assumed. When double-ended break occurs at the cold leg of the divertor cooling system, the copper cooling tube begins to melt within 3 s after the LOCA, even though the plasma is passively shut down at nearly 4 s. An active plasma shutdown system will be needed for such rapid transient accidents. On the other hand, when a small (1%) break LOCA occurs there, the plasma is passively shut down at nearly 36 s, which happens before the copper cooling tube begins to melt. When the double-ended break LOCA occurs at the cold leg of the first-wall cooling system, there is enough time (nearly 100 s) to shut down the plasma with a controllable method before the reactor structures are damaged. 21 refs., 8 figs

  13. Feasible reactor power cutback logic development for an integral reactor

    International Nuclear Information System (INIS)

    Han, Soon-Kyoo; Lee, Chung-Chan; Choi, Suhn; Kang, Han-Ok

    2013-01-01

    Major features of integral reactors that have been developed around the world recently are simplified operating systems and passive safety systems. Even though highly simplified control system and very reliable components are utilized in the integral reactor, the possibility of major component malfunction cannot be ruled out. So, feasible reactor power cutback logic is required to cope with the malfunction of components without inducing reactor trip. Simplified reactor power cutback logic has been developed on the basis of the real component data and operational parameters of plant in this study. Due to the relatively high rod worth of the integral reactor the control rod assembly drop method which had been adapted for large nuclear power plants was not desirable for reactor power cutback of the integral reactor. Instead another method, the control rod assembly control logic of reactor regulating system controls the control rod assembly movements, was chosen as an alternative. Sensitivity analyses and feasibility evaluations were performed for the selected method by varying the control rod assembly driving speed. In the results, sensitivity study showed that the performance goal of reactor power cutback system could be achieved with the limited range of control rod assembly driving speed. (orig.)

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  15. Specialists' meeting on passive and active safety features of LMFRs

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1991-07-01

    The objective of the meeting was to discuss and exchange information on passive and active safety concepts and to find some reasonable coupling of these concept, aiming at firmer establishment of plant safety and at the same time of plant cost reduction. The following main topical areas were discussed by delegates: (1) Overview - review of national status on the safety design approaches of LMFRs (2) Safety characteristics of decay heat removal system (DHRS) (3) Safety characteristics of reactor protection system (RPS) and reactor shutdown system (RSS) (4) Core safety characteristics.

  16. Specialists' meeting on passive and active safety features of LMFRs

    International Nuclear Information System (INIS)

    1991-01-01

    The objective of the meeting was to discuss and exchange information on passive and active safety concepts and to find some reasonable coupling of these concept, aiming at firmer establishment of plant safety and at the same time of plant cost reduction. The following main topical areas were discussed by delegates: (1) Overview - review of national status on the safety design approaches of LMFRs (2) Safety characteristics of decay heat removal system (DHRS) (3) Safety characteristics of reactor protection system (RPS) and reactor shutdown system (RSS) (4) Core safety characteristics

  17. NucleDyne's passive containment system

    International Nuclear Information System (INIS)

    Falls, O.B. Jr.; Kleimola, F.W.

    1987-01-01

    A simple definition of the passive containment system is that it is a total safeguards system for light water reactors designed to prevent and contain any accidental release of radioactivity. Its passive features utilize the natural laws of physics and thermodynamics. The system encompasses three basic containments constructed as one integrated structure on the reactor building foundation. The primary containment encloses the reactor pressure vessel and coolant system and passive engineered safety systems and components. Auxiliary containment enclosures house auxiliary systems and components. Secondary containment (the reactor building), housing the primary and auxiliary containment structures, provides a second containment barrier as added defense-in-depth against leakage of radioactivity for all accidents assumed by the industry. The generic features of the passive containment system are applicable to both the boiling water reactors and the pressurized water reactors as standardized features for all power ranges. These features provide for a zero source term, the industry's ultimate safety goal. This paper relates to a four-loop pressurized water reactor

  18. Integration of the functional reliability of two passive safety systems to mitigate a SBLOCA+BO in a CAREM-like reactor PSA

    Energy Technology Data Exchange (ETDEWEB)

    Mezio, Federico, E-mail: federico.mezio@cab.cnea.gov.ar [CNEA, Sede Central, Av. Del Libertador 8250, CABA (Argentina); Grinberg, Mariela [CNEA, Centro Atómico Bariloche, S.C. de Bariloche, Río Negro (Argentina); Lorenzo, Gabriel [CNEA, Sede Central, Av. Del Libertador 8250, CABA (Argentina); Giménez, Marcelo [CNEA, Centro Atómico Bariloche, S.C. de Bariloche, Río Negro (Argentina)

    2014-04-01

    Highlights: • An estimation of the Functional Unreliability was performed using RMPS methodology. • The methodology uses an improved response surface in order to estimate the FU. • The FU may become relevant to be analyzed in the Passive Safety Systems. • There were proposed two ways to incorporate the FU into an APS. - Abstract: This paper describes a case study of a methodological approach for assessing the functional reliability of passive safety systems (PSS) and its treatment within a probabilistic safety assessment (PSA). The functional unreliability (FU) can be understood as the failure probability of PSS to fulfill its mission due to the impairment of the related passive safety function. The safety function accomplishment is characterized and quantified by a performance indicator (PI), which is a measure of how far the system is from verifying its mission. PI uncertainties are estimated from uncertainty propagation of selected parameters. A methodology based on the reliability methodology for passive system (RMPS) one is used to estimate the FU associated to the isolation condensers (ICs) in combination with the accumulators (medium pressure injection system) of a CAREM-like integral advanced reactor. A small break loss of coolant accident with black-out is selected as an evaluation case. This implies success of reactor shut-down (inherent) and failure of residual heat removal by active systems. The safety function to accomplish is to refill the reactor pressure vessel (RPV) in order to avoid core damage. For this case, to allow the discharge of accumulators into RPV, the pressure must be reduced by the IC. The methodology for passive safety function assessment considers uncertainties in code parameters, besides uncertainties in engineering parameters (design, construction, operation and maintenance), in order to perform Monte Carlo simulations based on best estimate (B-E) plant model. Then, response surfaces based on PI are used for improving the

  19. Enhanced Passive Cooling for Waterless-Power Production Technologies

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez, Salvador B. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-06-14

    Recent advances in the literature and at SNL indicate the strong potential for passive, specialized surfaces to significantly enhance power production output. Our exploratory computational and experimental research indicates that fractal and swirl surfaces can help enable waterless-power production by increasing the amount of heat transfer and turbulence, when compared with conventional surfaces. Small modular reactors, advanced reactors, and non-nuclear plants (e.g., solar and coal) are ideally suited for sCO2 coolant loops. The sCO2 loop converts the thermal heat into electricity, while the specialized surfaces passively and securely reject the waste process heat in an environmentally benign manner. The resultant, integrated energy systems are highly suitable for small grids, rural areas, and arid regions.

  20. Realization of safety culture into a reactor plant-4S (super safe, small and simple) LMR

    International Nuclear Information System (INIS)

    Hattori, S.; Ikemoto, I.; Minato, A.

    1998-01-01

    International Nuclear Safety Advisory Group (INSAG) defines Safety Culture as the following; Safety Culture is that assembly of characteristics and attitudes in organization and individuals which establishes that, as an overriding priority, nuclear plant safety issues receive the attention warranted by their significance. While such Safety Culture is certainly a critical element of nuclear safety assurance, it is important to design nuclear power facilities as friendly to operators as possible with minimum dependence on human factors. From the viewpoint of ensuring supply in our global society, it will be necessary to have multiple approaches to further promote the use of nuclear energy worldwide despite various social and cultural restrictions. It should then be considered, as prospective options, to disperse small nuclear power plants throughout the world under technical, social and cultural conditions. Under this circumstance, we have quested for and now propose a a scheme of assuring sheer safety of nuclear power plants by implementing operator-friendly nuclear reactors virtually free from human errors. The scheme specifically includes the measures for improving reliability through fabricating more compact reactors with a continuous in-factory production system, simplify maintenance and inspection of the reactor system using passive systems and further relieving operators of burden of labors. (author)

  1. Passive safety features of low sodium void worth metal fueled cores in a bottom supported reactor vessel

    International Nuclear Information System (INIS)

    Chang, Y.I.; Marchaterre, J.F.; Wade, D.C.; Wigeland, R.A.; Kumaoka, Yoshio; Suzuki, Masao; Endo, Hiroshi; Nakagawa, Hiroshi

    1991-01-01

    A study has been performed on the passive safety features of low-sodium-void-worth metallic-fueled reactors with emphasis on using a bottom-supported reactor vessel design. The reactor core designs included self-sufficient types as well as actinide burners. The analyses covered the reactor response to the unprotected, i.e. unscrammed, transient overpower accident and the loss-of-flow accident. Results are given demonstrating the safety margins that were attained. 4 refs., 4 figs., 2 tabs

  2. Passive safety systems for decay heat removal of MRX

    Energy Technology Data Exchange (ETDEWEB)

    Ochiai, M; Iida, H; Hoshi, T [Japan Atomic Energy Research Inst., Ibaraki (Japan). Nuclear Ship System Lab.

    1996-12-01

    The MRX (marine Reactor X) is an advanced marine reactor, its design has been studied in Japan Atomic Energy Research Institute. It is characterized by four features, integral type PWR, in-vessel type control rod drive mechanisms, water-filled containment vessel and passive decay heat removal system. A water-filled containment vessel is of great advantage since it ensures compactness of a reactor plant by realizing compact radiation shielding. The containment vessel also yields passive safety of MRX in the event of a LOCA by passively maintaining core flooding without any emergency water injection. Natural circulation of water in the vessels (reactor and containment vessels) is one of key factors of passive decay heat removal systems of MRX, since decay heat is transferred from fuel rods to atmosphere by natural circulation of the primary water, water in the containment vessel and thermal medium in heat pipe system for the containment vessel water cooling in case of long terms cooling after a LOCA as well as after reactor scram. Thus, the ideal of water-filled containment vessel is considered to be very profitable and significant in safety and economical point of view. This idea is, however, not so familiar for a conventional nuclear system, so experimental and analytical efforts are carried out for evaluation of hydrothermal behaviours in the reactor pressure vessel and in the containment vessel in the event of a LOCA. The results show the effectiveness of the new design concept. Additional work will also be conducted to investigate the practical maintenance of instruments in the containment vessel. (author). 4 refs, 9 figs, 2 tabs.

  3. Simplified simulation of an experimental fast reactor plant

    International Nuclear Information System (INIS)

    Fujii, Masaaki; Fujita, Minoru.

    1978-01-01

    Purposes of the simulation are to study the dynamic behavior of a liquid metal-cooled experimental fast breeder reactor plant and to design the control system of the reactor plant by modified-RAPID (Reactor and Plant Integrated Dynamics) computer program. As for the plant model, the Japan Experimental Fast Reactor ''Joyo'' was referred to approximately. This computer program is designed for the calculation of steady-state and transient temperatures in a FBR plant; which is described by a model consisting of the core, upper and lower plenums, an intermediate heat exchanger, an air dump heat exchanger, primary-secondary and tertiary coolant systems and connecting pipes. The basic equations are solved numerically by finite difference approximation. The mathematical model for an experimental FBR plant is useful for the design of the control system of FBR plants. The results of numerical simulation showed that the proportional change in the flow rates of the primary and secondary coolant loops provides good performance in relation to the stepped change in the power level. (J.P.N.)

  4. Study on operational aspect of natural circulation HLMC reactor (1)

    International Nuclear Information System (INIS)

    Sienicki, J.J.; Cahalan, J.E.; Spencer, B.W.

    2000-08-01

    The concept of a heavy liquid metal cooled fast reactor that achieves 100% natural circulation heat removal from the core has the potential to attain improved cost competitiveness through extreme simplification, proliferation resistance, and heightened passive safety. The concept offers the potential for simplifications in plant control strategies wherein inherent reactor feedbacks may restore balance between energy release and heat removal from the reactor during operation as well as providing passive reactivity shutdown in the event of transients involving failure to scram. This study was initiated to evaluate the operational characteristics of the 100% natural circulation reactor under normal and transient states using a plant dynamics analysis computer code and to seek design and operational optimization of the concept. In the current Phase I of the project, the stage for the overall study has been prepared. A coupled thermal hydraulics-kinetics plant dynamics analysis code has been developed/modified that has the capabilities to calculate operational and accident transients. Code input has been prepared for the heavy liquid metal cooled natural circulation reactor concept. A preliminary analysis using the plant dynamics code and its input to calculate three illustrative cases relevant to initial startup, shutdown following long-term operation, and change in turbine load demonstrates the capability to analyze typical transient cases. (author)

  5. The passive safety systems of the Swr 1000

    International Nuclear Information System (INIS)

    Neumann, D.

    2001-01-01

    In recent years, a new boiling water reactor (BWR) plant called the SWR 1000 has been developed by Siemens on behalf of Germany's electric utilities. This new plant design concept incorporates the wide range of operating experience gained with German BWRs. The main objective behind developing the SWR 1000 was to design a plant with a rated electric output of approximately 1000 MW which would not only have a lower capital cost and lower power generating costs but would also provide a much higher level of nuclear safety compared to plants currently in operation. This safety-related goal has been met through, for example, the use of passive safety equipment. Passive systems make a significant contribution towards increasing the over-all level of plant safety due to the way in which they operate. They function solely accord-ing to basic laws of nature, such as gravity, and perform their designated functions with-out any need for electric power or other sources of external energy, or signals from instrumentation and control (I and C) equipment. The passive safety systems have been designed such that design basis accidents can be controlled using just these systems alone. However, the design concept of the SWR 1000 is nevertheless still based on the provision of active safety systems in addition to passive systems. (author)

  6. Passive pH adjustment of nuclear reactor containment flood water

    International Nuclear Information System (INIS)

    Gerlowski, T.J.

    1986-01-01

    A method is described of automatically and passively adjusting the pH of the recirculating liquid used to flood the containment structure of a nuclear reactor upon the occurence of an accident in order to cool the reactor core, wherein the containment structure has a concrete floor which is provided with at least one sump from which the liquid is withdrawn for recirculation via at least one outlet pipe. The method consists of: prior to flooding and during or prior to normal operation of the reactor, providing at least one perforated basket within at least one sump with the basket containing crystals of a pH adjusting chemical which is soluble in the liquid, and covering each basket with a plastic coating which is likewise soluble in the liquid, whereby upon flooding of the containment structure the liquid in the sump will reach the level of the baskets, causing the coating and the crystals to be dissolved and the chemical to mix with the recirculating liquid to adjust the pH

  7. ANALISIS TRANSIEN PADA PASSIVE COMPACT MOLTEN SALT REACTOR (PCMSR

    Directory of Open Access Journals (Sweden)

    M. Makrus Imron

    2015-04-01

    Full Text Available Penggunaan bahan bakar cair berupa garam LiF-BeF2-ThF4-UF4 pada Passive Compact Molten Salt Reactor (PCMSR meyebabkan pengendalian daya pada PCMSR dapat dilakukan dengan mengendalikan laju aliran bahan bakar dan pendingin. Sedangkan dari sistem keselamatan, penggunaan bahan bakar cair menjadikan PCMSR memiliki karakter keselamatan melekat (inherent safety yang baik. Pada penelitian ini telah dilakukan analisis transien PCMSR pada tiga kondisi, yaitu: ketika terjadi perubahan laju aliran bahan bakar, ketika terjadi perubahan laju aliran pendingin dan ketika terdapat kegagalan pada sistem pelepasan panas (loss of heat sink. Penelitian dilakukan dengan memodelkan reaktor pada kondisi tunak menggunakan paket program. Standart Reactor Analysis Code (SRAC. Selanjutnya dari keluaran paket program SRAC diperoleh data data yang meliputi fluks netron,konstanta grup, kontanta peluran prekusor netron, fraksi netron kasip untuk perhitungan transien. Penelitian ini menunjukkan bahwa penurunan laju aliran bahan bakar sebesar 50 % dari laju bahan bakar sebelumnya, menyebabkan daya pada PCMSR turun menjadi 78 % dari daya sebelumnya. Dan penurunan laju aliran pendingin sebesar 50 % dari laju pendingin sebelumnya, menyebabkan daya pada PCMSR turun menjadi 63 % dari daya sebelumnya. Sedangkan pada saat terjadi loss of heat sink daya PCMSR menunjukkan penurunan. Kata kunci: PCMSR, transien, daya, laju aliran.   The use of liquid fuels in the form of molten salts LiF-BeF2-ThF4-UF4 in Passive Compact Molten Salt Reactor (PCMSR makes power control at PCMSR can be done by controlling the flow rate of fuel and coolant. In addition, from safety systems aspect, the use of liquid fuels makes PCMSR has good inherent safety characteristics. In this study transient analysis has been carried out on three conditions of PCMSR, namely when the fuel flow rate is changing, when the coolant flow rate is changing and when there is loss of heat sink condition. This research is

  8. FAST and SAFE Passive Safety Devices for Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hartanto, Donny; Kim, Chihyung; Kim, In-Hyung; Kim, Yonghee [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    The major factor is the impact of the neutron spectral hardening. The second factor that affects the CVR is reduced capture by the coolant when the coolant voiding occurs. To improve the CVR, many ideas and concepts have been proposed, which include introduction of an internal blanket, spectrum softening, or increasing the neutron leakage. These ideas may reduce the CVR, but they deteriorate the neutron economy. Another potential solution is to adopt a passive safety injection device such as the ARC (autonomous reactivity control) system, which is still under development. In this paper, two new concepts of passive safety devices are proposed. The devices are called FAST (Floating Absorber for Safety at Transient) and SAFE (Static Absorber Feedback Equipment). Their purpose is to enhance the negative reactivity feedback originating from the coolant in fast reactors. SAFE is derived to balance the positive reactivity feedback due to sodium coolant temperature increases. It has been demonstrated that SAFE allows a low-leakage SFR to achieve a self-shutdown and self-controllability even though the generic coolant temperature coefficient is quite positive and the coolant void reactivity can be largely managed by the new FAST device. It is concluded that both FAST and SAFE devices will improve substantially the fast reactor safety and they deserve more detailed investigations.

  9. Passive BWR integral LOCA testing at the Karlstein test facility INKA

    Energy Technology Data Exchange (ETDEWEB)

    Drescher, Robert [AREVA GmbH, Erlangen (Germany); Wagner, Thomas [AREVA GmbH, Karlstein am Main (Germany); Leyer, Stephan [TH University of Applied Sciences, Deggendorf (Germany)

    2014-05-15

    KERENA is an innovative AREVA GmbH boiling water reactor (BWR) with passive safety systems (Generation III+). In order to verify the functionality of the reactor design an experimental validation program was executed. Therefore the INKA (Integral Teststand Karlstein) test facility was designed and erected. It is a mockup of the BWR containment, with integrated pressure suppression system. While the scaling of the passive components and the levels match the original values, the volume scaling of the containment compartments is approximately 1:24. The storage capacity of the test facility pressure vessel corresponds to approximately 1/6 of the KERENA RPV and is supplied by a benson boiler with a thermal power of 22 MW. In March 2013 the first integral test - Main Steam Line Break (MSLB) - was executed. The test measured the combined response of the passive safety systems to the postulated initiating event. The main goal was to demonstrate the ability of the passive systems to ensure core coverage, decay heat removal and to maintain the containment within defined limits. The results of the test showed that the passive safety systems are capable to bring the plant to stable conditions meeting all required safety targets with sufficient margins. Therefore the test verified the function of those components and the interplay between them. The test proved that INKA is an unique test facility, capable to perform integral tests of passive safety concepts under plant-like conditions. (orig.)

  10. A Simple Fully Passive Safety Option for SMART SBLOCA

    International Nuclear Information System (INIS)

    Lee, Won Jae

    2012-01-01

    SMART reactor, an integral pressurized water reactor (iPWR), is developed by KAERI and now under standard design licensing review. Integral reactor design of the SMART has small diameter penetrations below 2 inches at upper parts of reactor pressure vessel (RPV) and the core is located at very lower part. Amount of reactor coolant inventory is around 0.55tons/MWth during normal operations, which is seven times more than that of conventional PWRs. Such intrinsic safety features of the SMART can provide prolonged core cooling during a small-break loss-of-coolant accident (SBLOCA). As an engineered safety feature for SBLOCA, electrically two-train and mechanically four-train active safety injection (SI) systems are provided to refill the RPV, whose safety been proven through safety analysis and experiments. In addition, four-train passive residual heat removal systems (PRHRSs) are provided to remove core decay heat by natural circulation in the secondary side of steam generators during transient and accident conditions. After Fukushima disaster, a passive safety of nuclear power plants has become more emphasized than conventional active safety, even though there are still debates whether it can really insure the realistic safety. Passive safety is defined such that the core safety is ensured for 72 hours after accidents without any active safety systems and operator actions. In light of this, a simple fully passive safety option for SBLOCA is proposed: low-pressure safety injection tanks (SITs) and heat pipes submerged in the PRHRS emergency coolant tanks (ECTs). Post-LOCA long-term cooling after 72 hours is provided by sump recirculation using shutdown cooling system. Realistic analysis method using MARS3.1 is used to derive fully passive safety option, and then to screen design and operating parameters and to demonstrate the safety performance of SITs. SI line break is selected as a reference SBLOCA scenario

  11. Final report-passive safety optimization in liquid sodium-cooled reactors

    International Nuclear Information System (INIS)

    Cahalana, J. E.; Hahn, D.

    2007-01-01

    This report summarizes the results of a three-year collaboration between Argonne National Laboratory (ANL) and the Korea Atomic Energy Research Institute (KAERI) to identify and quantify the performance of innovative design features in metallic-fueled, sodium-cooled fast reactor designs. The objective of the work was to establish the reliability and safety margin enhancements provided by design innovations offering significant potential for construction, maintenance, and operating cost reductions. The project goal was accomplished with a combination of advanced model development (Task 1), analysis of innovative design and safety features (Tasks 2 and 3), and planning of key safety experiments (Task 4). Task 1--Computational Methods for Analysis of Passive Safety Design Features: An advanced three-dimensional subassembly thermal-hydraulic model was developed jointly and implemented in ANL and KAERI computer codes. The objective of the model development effort was to provide a high-accuracy capability to predict fuel, cladding, coolant, and structural temperatures in reactor fuel subassemblies, and thereby reduce the uncertainties associated with lower fidelity models previously used for safety and design analysis. The project included model formulation, implementation, and verification by application to available reactor tests performed at EBR-II. Task 2--Comparative Analysis and Evaluation of Innovative Design Features: Integrated safety assessments of innovative liquid metal reactor designs were performed to quantify the performance of inherent safety features. The objective of the analysis effort was to identify the potential safety margin enhancements possible in a sodium-cooled, metal-fueled reactor design by use of passive safety mechanisms to mitigate low-probability accident consequences. The project included baseline analyses using state-of-the-art computational models and advanced analyses using the new model developed in Task 1. Task 3--Safety

  12. Study on thermal-hydraulic phenomena identification of passive heat removal facilities

    International Nuclear Information System (INIS)

    Park, J. Y.

    2011-01-01

    Recently, passive heat removal facilities have been integral features of new generation or future reactor designs worldwide. This is because the passive heat removal facilities depending on a natural force such as buoyancy can give much higher operational reliability compared to active heat removal facilities depending on pumped fluid flow and as a result they can decrease core damage frequency of a nuclear power plant drastically ever achievable before. Keeping pace with this global trend, SMART and APR+ reactors also have introduced passive heat removal features such as a passive residual heat removal system (PRHRS) and a passive auxiliary feed water system (PAFS) in their designs. Since many thermal-hydraulic (T-H) phenomena including steam condensation are involved during operation of the passive heat removal facilities, they ought to be properly simulated by T-H codes such as MARS-KS and RELAP5 in order to guarantee reliable safety analysis by these codes. Unfortunately, however, these T-H codes are not well validated with respect to phenomena related to passive heat removal mechanism because previous focus on these codes validation was mainly on the LB LOCA and resulting phenomena. To resolve this gap, Korea Institute of Nuclear Safety has initiated a research program on the development of safety analysis technology for passive heat removal facilities. The main target of this program is PRHRS and PAFS in SMART and APR+ reactors and through this program, validation of capability of existing T-H codes and improvement of codes regarding passive facilities analysis are to be sought. In part of this research, T-H phenomena important to passive heat removal facilities (PRHRS and PAFS) are investigated in the present study

  13. Nuclear power reactors of new generation

    International Nuclear Information System (INIS)

    Ponomarev-Stepnoi, N.N.; Slesarev, I.S.

    1988-01-01

    The paper presents discussions on the following topics: fuel supply for nuclear power; expansion of the sphere of nuclear power applications, such as district heating; comparative estimates of power reactor efficiencies; safety philosophy of advanced nuclear plants, including passive protection and inherent safety concepts; nuclear power unit of enhanced safety for the new generation of nuclear power plants. The emphasis is that designers of new generation reactors face a complicated but technically solvable task of developing highly safe, efficient, and economical nuclear power sources having a wide sphere of application

  14. Probabilistic Analysis of Passive Safety System Reliability in Advanced Small Modular Reactors: Methodologies and Lessons Learned

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, David; Bucknor, Matthew; Brunett, Acacia; Grelle, Austin

    2015-06-28

    Many advanced small modular reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended due to deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize with a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper describes the most promising options: mechanistic techniques, which share qualities with conventional probabilistic methods, and simulation-based techniques, which explicitly account for time-dependent processes. The primary intention of this paper is to describe the strengths and weaknesses of each methodology and highlight the lessons learned while applying the two techniques while providing high-level results. This includes the global benefits and deficiencies of the methods and practical problems encountered during the implementation of each technique.

  15. A reliability assessment methodology for the VHTR passive safety system

    International Nuclear Information System (INIS)

    Lee, Hyungsuk; Jae, Moosung

    2014-01-01

    The passive safety system of a VHTR (Very High Temperature Reactor), which has recently attracted worldwide attention, is currently being considered for the design of safety improvements for the next generation of nuclear power plants in Korea. The functionality of the passive system does not rely on an external source of an electrical support system, but on the intelligent use of natural phenomena. Its function involves an ultimate heat sink for a passive secondary auxiliary cooling system, especially during a station blackout such as the case of the Fukushima Daiichi reactor accidents. However, it is not easy to quantitatively evaluate the reliability of passive safety for the purpose of risk analysis, considering the existing active system failure since the classical reliability assessment method cannot be applied. Therefore, we present a new methodology to quantify the reliability based on reliability physics models. This evaluation framework is then applied to of the conceptually designed VHTR in Korea. The Response Surface Method (RSM) is also utilized for evaluating the uncertainty of the maximum temperature of nuclear fuel. The proposed method could contribute to evaluating accident sequence frequency and designing new innovative nuclear systems, such as the reactor cavity cooling system (RCCS) in VHTR to be designed and constructed in Korea.

  16. U.S. Nuclear Power Reactor Plant Status

    Data.gov (United States)

    Nuclear Regulatory Commission — Demographic data on U.S. commercial nuclear power reactors, including: plant name/unit number, docket number, location, licensee, reactor/containment type, nuclear...

  17. 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

  18. Fast reactor system factors affecting reprocessing plant design

    International Nuclear Information System (INIS)

    Allardice, R.H.; Pugh, O.

    1982-01-01

    The introduction of a commercial fast reactor electricity generating system is very dependent on the availability of an efficient nuclear fuel cycle. Selection of fuel element constructional materials, the fuel element design approach and the reactor operation have a significant influence on the technical feasibility and efficiency of the reprocessing and waste management plants. Therefore the fast reactor processing plant requires liaison between many design teams -reactor, fuel design, reprocessing and waste management -often with different disciplines and conflicting objectives if taken in isolation and an optimised approach to determining several key parameters. A number of these parameters are identified and the design approach discussed in the context of the reprocessing plant. Radiological safety and its impact on design is also briefly discussed. (author)

  19. Nuclear power plant with several reactors

    Energy Technology Data Exchange (ETDEWEB)

    Grishanin, E I; Ilyunin, V G; Kuznetsov, I A; Murogov, V M; Shmelev, A N

    1972-05-10

    A design of a nuclear power plant suggested involves several reactors consequently transmitting heat to a gaseous coolant in the joint thermodynamical circuit. In order to increase the power and the rate of fuel reproduction the low temperature section of the thermodynamical circuit involves a fast nuclear reactor, whereas a thermal nuclear reactor is employed in the high temperature section of the circuit for intermediate heating and for over-heating of the working body. Between the fast nuclear and the thermal nuclear reactors there is a turbine providing for the necessary ratio between pressures in the reactors. Each reactor may employ its own coolant.

  20. Passive containment cooling system with drywell pressure regulation for boiling water reactor

    Science.gov (United States)

    Hill, P.R.

    1994-12-27

    A boiling water reactor is described having a regulating valve for placing the wetwell in flow communication with an intake duct of the passive containment cooling system. This subsystem can be adjusted to maintain the drywell pressure at (or slightly below or above) wetwell pressure after the initial reactor blowdown transient is over. This addition to the PCCS design has the benefit of eliminating or minimizing steam leakage from the drywell to the wetwell in the longer-term post-LOCA time period and also minimizes the temperature difference between drywell and wetwell. This in turn reduces the rate of long-term pressure buildup of the containment, thereby extending the time to reach the design pressure limit. 4 figures.

  1. Design and computational analysis of passive siphon breaker for 49-2 swimming pool reactor

    International Nuclear Information System (INIS)

    Yue Zhiting; Song Yunpeng; Liu Xingmin; Zou Yao; Wu Yuanyuan

    2014-01-01

    Based on safety considerations, a passive siphon breaker will be added to the primary cooling system of 49-2 Swimming Pool Reactor (SPR). With the breaker location determined, the capability of siphon breakers with diameters of 1.5 cm and 2.0 cm was calculated and analyzed respectively by RELAP5/MOD3.3 code. The results show that in the condition of large break loss of coolant accident these two sizes of siphon breakers are able to break the siphon phenomena, and maintain the pool water level above the reactor core when the reactor and the pump are shutdown. In the end, to be conservative, the siphon breaker with diameter of 2.0 cm is adopted. (authors)

  2. Establishment of design concept of large capacity passive reactor KP1000 and performance evaluation of safety system for LBLOCA

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seong O.; Hwang, Young Dong; Kim, Young In; Chang, Moon Hee

    1997-03-01

    This study was performed to establish the design concepts and to evaluate the performance of safety features of large capacity passive reactor (1000 MWe grade). The design concepts of the large capacity passive reactor `KP1000` were established to generate 1000 MW electric power based on the AP600 of Westinghouse by increasing the number of reactor coolant loop and by increasing the size of reactor internals/core. To implement the analysis of the LBLOCA for KP1000, various kinds of computer codes being considered, it was concluded that RELAP5 was the most appropriate one in availability and operations in present situation. By the analysis of the computer code `RELAP5/Mod3.2.1.2`, following conclusions were derived as described below. First, by spectrum analysis of the discharge factor of the berak part, the most conservative discharge factor C{sub D}=1.2 and the PCT value of KP1000 was 1254F, which is slightly higher than the value of AP600 but is much less than the existing active reactor `Kori 3 and 4` where blowdown PCT value is 1693.4 deg F and reflooding PCT is 1918.4 deg F. Second, after the 200 seconds from the initiation of LBLOCA, IRWST water was supplied in a stable state and the maximum temperature of clad were maintained in a saturated condition. Therefore, it was concluded that the passive safety features of KP1000 keep reactor core from being damaged for large break LOCA. (author). 11 refs., 28 tabs., 37 figs.

  3. Fuel management of mixed reactor type power plant systems

    International Nuclear Information System (INIS)

    Csom, Gyula

    1988-01-01

    In equilibrium symbiotic power plant system containing both thermal reactors and fast breeders, excess plutonium produced by the fast breeders is used to enrich the fuel of the thermal reactors. In plutonium deficient symbiotic power plant system plutonium is supplied both by thermal plants and fast breeders. Mathematical models were constructed and different equations solved to characterize the fuel utilization of both systems if they contain only a single thermal type and a single fast type reactor. The more plutonium is produced in the system, the higher output ratio of thermal to fast reactors is achieved in equilibrium symbiotic power plant system. Mathematical equations were derived to calculate the doubling time and the breeding gain of the equilibrium symbiotic system. (V.N.) 2 figs.; 2 tabs

  4. Active or passive systems? The EPR approach

    International Nuclear Information System (INIS)

    Bonhomme, N.; Py, J.P.

    1996-01-01

    In attempting to review how EPR is contemplated to meet requirements applicable to future nuclear power plants, the authors indicate where they see the markets and the corresponding unit sizes for the EPR which is a generic key factor for competitiveness. There are no reason in industrialized countries, other than USA (where the investment and amortizing practices under control by Public Utility Commission are quite particular), not to build future plants in the 1000 to 1500 MWe range. Standardization, which has been actively applied all along the French program and for the Konvoi plants, does not prevent evolution and allows to concentrate large engineering effort in smooth realization of plants and achieve actual construction and commissioning without significant delays. In order to contribute to public trust renewal, a next generation of power reactors should be fundamentally less likely to incur serious accidents. To reach this goal the best of passive and active systems must be considered without forgetting that the most important source of knowledge is construction and operating experience. Criteria to assess passive systems investigated for possible implementation in the EPR, such as simplicity of design, impact on plant operation, safety and cost, are discussed. Examples of the principal passive systems investigated are described and reasons why they have been dropped after screening through the criteria are given. (author). 11 figs

  5. Active or passive systems? The EPR approach

    Energy Technology Data Exchange (ETDEWEB)

    Bonhomme, N [Nuclear Power International, Cedex (France); Py, J P [FRAMATOME, Cedex (France)

    1996-12-01

    In attempting to review how EPR is contemplated to meet requirements applicable to future nuclear power plants, the authors indicate where they see the markets and the corresponding unit sizes for the EPR which is a generic key factor for competitiveness. There are no reason in industrialized countries, other than USA (where the investment and amortizing practices under control by Public Utility Commission are quite particular), not to build future plants in the 1000 to 1500 MWe range. Standardization, which has been actively applied all along the French program and for the Konvoi plants, does not prevent evolution and allows to concentrate large engineering effort in smooth realization of plants and achieve actual construction and commissioning without significant delays. In order to contribute to public trust renewal, a next generation of power reactors should be fundamentally less likely to incur serious accidents. To reach this goal the best of passive and active systems must be considered without forgetting that the most important source of knowledge is construction and operating experience. Criteria to assess passive systems investigated for possible implementation in the EPR, such as simplicity of design, impact on plant operation, safety and cost, are discussed. Examples of the principal passive systems investigated are described and reasons why they have been dropped after screening through the criteria are given. (author). 11 figs.

  6. Condition Based Prognostics of Passive Components - A New Era for Nuclear Power Plant Life Management

    International Nuclear Information System (INIS)

    Bakhtiari, S.; Mohanty, S.; Prokofiev, I.; Tregoning, R.

    2012-01-01

    As part of a research project sponsored by the U.S. NRC, Argonne National Laboratory (ANL) conducted scoping studies to identify viable and promising sensors and techniques for in-situ inspection and real-time monitoring of degradation in nuclear power plant (NPP) systems, structures, and components (SSC). Significant advances have been made over the past two decades toward development of online monitoring (OLM) techniques for detection, diagnostics, and prognostics of degradation in active nuclear power plant (NPP) components (e.g., pumps, valves). However, early detection of damage and degradation in safety-critical passive components, (e.g. piping, tubing pressure vessel), is challenging, and will likely remain so for the foreseeable future. Ensuring the structural integrity of the reactor pressure vessel (RPV) and piping systems in particular is a prerequisite to long term safe operation of NPPs. The current practice is to implement inservice inspection (ISI) and preventive maintenance programs. While these programs have generally been successful, they are limited in that information is only obtained during plant outages. Additionally, these inspections, often the critical path in the outage schedule, are costly, time consuming, and involve potentially high dose to nondestructive examination/evaluation (NDE) personnel. A viable plant-wide on-line structural health monitoring program for continuous and automatic monitoring of critical SSCs could be a more effective approach for guarding against unexpected failures. Specifically, OLM information about the current condition of the SSCs could be input to an online prognostics (OLP) system to forecast their remaining useful life in real time. This paper provides an overview of scoping studies performed at ANL on assessing the viability of OLM and OLP systems for real time and automated monitoring and remaining of condition and the remaining useful life of passive components in NPPs. (author)

  7. Invariant methods for an ensemble-based sensitivity analysis of a passive containment cooling system of an AP1000 nuclear power plant

    International Nuclear Information System (INIS)

    Di Maio, Francesco; Nicola, Giancarlo; Borgonovo, Emanuele; Zio, Enrico

    2016-01-01

    Sensitivity Analysis (SA) is performed to gain fundamental insights on a system behavior that is usually reproduced by a model and to identify the most relevant input variables whose variations affect the system model functional response. For the reliability analysis of passive safety systems of Nuclear Power Plants (NPPs), models are Best Estimate (BE) Thermal Hydraulic (TH) codes, that predict the system functional response in normal and accidental conditions and, in this paper, an ensemble of three alternative invariant SA methods is innovatively set up for a SA on the TH code input variables. The ensemble aggregates the input variables raking orders provided by Pearson correlation ratio, Delta method and Beta method. The capability of the ensemble is shown on a BE–TH code of the Passive Containment Cooling System (PCCS) of an Advanced Pressurized water reactor AP1000, during a Loss Of Coolant Accident (LOCA), whose output probability density function (pdf) is approximated by a Finite Mixture Model (FMM), on the basis of a limited number of simulations. - Highlights: • We perform the reliability analysis of a passive safety system of Nuclear Power Plant (NPP). • We use a Thermal Hydraulic (TH) code for predicting the NPP response to accidents. • We propose an ensemble of Invariant Methods for the sensitivity analysis of the TH code • The ensemble aggregates the rankings of Pearson correlation, Delta and Beta methods. • The approach is tested on a Passive Containment Cooling System of an AP1000 NPP.

  8. Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems

    Science.gov (United States)

    McDermott, Daniel J.; Schrader, Kenneth J.; Schulz, Terry L.

    1994-01-01

    The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

  9. Improving plant availability by predicting reactor trips

    International Nuclear Information System (INIS)

    Frank, M.V.; Epstein, S.A.

    1986-01-01

    Management Ahnalysis Company (MAC) has developed and applied two complementary software packages called RiTSE and RAMSES. Together they provide an mini-computer workstation for maintenance and operations personnel to dramatically reduce inadvertent reactor trips. They are intended to be used by those responsible at the plant for authorizing work during operation (such as a clearance coordinator or shift foreman in U.S. plants). They discover and represent all components, processes, and their interactions that could case a trip. They predict if future activities at the plant would cause a reactor trip, provide a reactor trip warning system and aid in post-trip cause analysis. RAMSES is a general reliability engineering software package that uses concepts of artificial intelligence to provide unique capabilities on personal and mini-computers

  10. Removal of FePO4 and Fe3(PO4)2 crystals on the surface of passive fillers in Fe0/GAC reactor using the acclimated bacteria

    International Nuclear Information System (INIS)

    Lai, Bo; Zhou, Yuexi; Yang, Ping; Wang, Juling; Yang, Jinghui; Li, Huiqiang

    2012-01-01

    Highlights: ► Fe 3 (PO 4 ) 2 and FePO 4 crystals would weaken treatment efficiency of Fe 0 /GAC reactor. ► Fe 3 (PO 4 ) 2 and FePO 4 crystals could be removed by the acclimated bacteria. ► FeS and sulfur in the passive film would be removed by the sulfur-oxidizing bacteria. ► Develop a cost-effective bio-regeneration technology for the passive fillers. - Abstract: As past studies presented, there is obvious defect that the fillers in the Fe 0 /GAC reactor begin to be passive after about 60 d continuous running, although the complicated, toxic and refractory ABS resin wastewater can be pretreated efficiently by the Fe 0 /GAC reactor. During the process, the Fe 3 (PO 4 ) 2 and FePO 4 crystals with high density in the passive film are formed by the reaction between PO 4 3− and Fe 2+ /Fe 3+ . Meanwhile, they obstruct the formation of macroscopic galvanic cells between Fe 0 and GAC, which will lower the wastewater treatment efficiency of Fe 0 /GAC reactor. In this study, in order to remove the Fe 3 (PO 4 ) 2 and FePO 4 crystals on the surface of the passive fillers, the bacteria were acclimated in the passive Fe 0 /GAC reactor. According to the results, it can be concluded that the Fe 3 (PO 4 ) 2 and FePO 4 crystals with high density in the passive film could be decomposed or removed by the joint action between the typical propionic acid type fermentation bacteria and sulfate reducing bacteria (SRB), whereas the PO 4 3− ions from the decomposition of the Fe 3 (PO 4 ) 2 and FePO 4 crystals were released into aqueous solution which would be discharged from the passive Fe 0 /GAC reactor. Furthermore, the remained FeS and sulfur (S) in the passive film also can be decomposed or removed easily by the oxidation of the sulfur-oxidizing bacteria. This study provides some theoretical references for the further study of a cost-effective bio-regeneration technology to solve the passive problems of the fillers in the zero-valent iron (ZVI) or Fe 0 /GAC reactor.

  11. Discharges from a fast reactor reprocessing plant

    International Nuclear Information System (INIS)

    Barnes, D.S.

    1987-01-01

    The purpose of this paper is to assess the environmental impact of the calculated routine discharges from a fast reactor fuel reprocessing plant. These assessments have been carried out during the early stages of an evolving in-depth study which culminated in the design for a European demonstration reprocessing plant (EDRP). This plant would be capable of reprocessing irradiated fuel from a series of European fast reactors. Cost-benefit analysis has then been used to assess whether further reductions in the currently predicted routine discharges would be economically justified

  12. Study of passive residual heat removal system of a modular small PWR reactor; Estudo do sistema passivo de remoção de calor residual de um reator PWR pequeno modular

    Energy Technology Data Exchange (ETDEWEB)

    Araujo, Nathália N., E-mail: nathalianunes@poli.ufrj.br [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Departamento de Engenharia Nuclear; Faccini, José L.H., E-mail: faccini@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil); Su, Jian, E-mail: sujian@lasme.coppe.ufrj.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2017-07-01

    This paper presents a study on the passive residual heat removal system (PRHRS) of a small modular nuclear reactor (SMR) of 75MW. More advanced nuclear reactors, such as generation III + and IV, have passive safety systems that automatically go into action in order to prevent accidents. The purpose of the PRHRS is to transfer the decay heat from the reactor's nuclear fuel, keeping the core cooled after the plant has shut down. It starts operating in the event of fall of power supply to the nuclear station, or in the event of an unavailability of the steam generator water supply system. Removal of decay heat from the core of the reactor is accomplished by the flow of the primary refrigerant by natural circulation through heat exchangers located in a pool filled with water located above the core. The natural circulation is caused by the density gradient between the reactor core and the pool. A thermal and comparative analysis of the PRHRS was performed consisting of the resolution of the mass conservation equations, amount of movement and energy and using incompressible fluid approximations with the Boussinesq approximation. Calculations were performed with the aid of Mathematica software. A design of the heat exchanger and the cooling water tank was done so that the core of the reactor remained cooled for 72 hours using only the PRHRS.

  13. The water desalination complex based on ABV-type reactor plant

    International Nuclear Information System (INIS)

    Panov, Yu.K.; Fadeev, Yu.P.; Vorobiev, V.M.; Baranaev, Yu.D.

    1997-01-01

    A floating nuclear desalination complex with two barges, one for ABV type reactor plant, with twin reactor 2 x 6 MW(e), and one for reverse osmosis desalination plant, was described. The principal specifications of the ABV type reactor plant and desalination barge were given. The ABV type reactor has a traditional two-circuit layout using an integral type reactor vessel with all mode natural convection of primary coolant. The desalted water cost was estimated to be around US $0.86 per cubic meter. R and D work has been performed and preparations for commercial production are under way. (author)

  14. Nuclear reactor kinetics and plant control

    CERN Document Server

    Oka, Yoshiaki

    2013-01-01

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

  15. Passivity Enhancement in RES Based Power Plant with Paralleled Grid-Connected Inverters

    DEFF Research Database (Denmark)

    Bai, Haofeng; Wang, Xiongfei; Blaabjerg, Frede

    2016-01-01

    Harmonic instability is threatening the operation of power plants with multiple grid connected converters in parallel. To analyze and improve the stability of the grid connected converters, the passivity of the output admittance converters is first analyzed in this paper. It is shown that the non-passivity...

  16. General Aspects of CAREM-25 Reactor

    International Nuclear Information System (INIS)

    Delmastro, Dario; Gomez, S.; Mazzi, R.; Gomez de Soler, S.; Santecchia, A.; Ishida, V.

    2000-01-01

    CAREM project consists on the development and design of an advanced Nuclear Power Plant. In order to verify its innovative features the construction of a prototype is planned. In this paper the main technical characteristics of CAREM-25 prototype reactor are presented. This is a very low power innovative reactor (100MWth) conceived with new generation design solutions. Based on an indirect cycle integrated light water reactor using enriched uranium, CAREM has some distinctive features that greatly simplify the reactor and also contribute to a high level of safety: integrated primary system, primary system cooling by natural convection, selfpressurization, and passive safety systems

  17. Adapting a reactor safety assessment system for specific plants

    International Nuclear Information System (INIS)

    Ballard, T.L.; Cordes, G.A.

    1991-01-01

    The Reactor Safety Assessment System (RSAS) is an expert system being developed by the Idaho National Engineering Laboratory, the University of Maryland (UofM) and US Nuclear Regulatory Commission (NRC) for use in the NRC Operations center. RSAS is designed to help the Reactor Safety Team monitor and project core status during an emergency at a licensed nuclear power plant. Analysis uses a hierarchical plant model based on equipment availability and automatically input parametric plant information. There are 3 families of designs of pressurized water reactors and 75 plants using modified versions of the basic design. In order to make an RSAS model for each power plant, a generic model for a given plant type is used with differences being specified by plant specific files. Graphical displays of this knowledge are flexible enough to handle any plant configuration. A variety of tools have been implemented to make it easy to modify a design to fit a given plant while minimizing chance for error. 3 refs., 4 figs

  18. European BWR R and D cluster for innovative passive safety systems

    International Nuclear Information System (INIS)

    Hicken, E.F.; Lensa, W. von

    1996-01-01

    The main technological innovation trends for future nuclear power plants tend towards a broader use of passive safety systems for the prevention, mitigation and managing of severe accident scenarios. Several approaches have been undertaken in a number of European countries to study and demonstrate the feasibility and charateristics of innovative passive safety systems. The European BWR R and D Cluster combines those experimental and analytical efforts that are mainly directed to the introduction of passive safety systems into boiling water reactor technology. The Cluster is grouped around thermohydraulic test facilities in Europe for the qualification of innovative BWR safety systems, also taking into account especially the operating experience of the nuclear power plant Dodewaard and other BWRs, which already incorporated some passive safety features. The background, the objectives, the structure of the project and the work programme are presented in this paper as well as an outline of the significance of the expected results. (orig.) [de

  19. European passive plant program preliminary safety analyses to support system design

    International Nuclear Information System (INIS)

    Saiu, Gianfranco; Barucca, Luciana; King, K.J.

    1999-01-01

    In 1994, a group of European Utilities, together with Westinghouse and its Industrial Partner GENESI (an Italian consortium including ANSALDO and FIAT), initiated a program designated EPP (European Passive Plant) to evaluate Westinghouse Passive Nuclear Plant Technology for application in Europe. In the Phase 1 of the European Passive Plant Program which was completed in 1996, a 1000 MWe passive plant reference design (EP1000) was established which conforms to the European Utility Requirements (EUR) and is expected to meet the European Safety Authorities requirements. Phase 2 of the program was initiated in 1997 with the objective of developing the Nuclear Island design details and performing supporting analyses to start development of Safety Case Report (SCR) for submittal to European Licensing Authorities. The first part of Phase 2, 'Design Definition' phase (Phase 2A) was completed at the end of 1998, the main efforts being design definition of key systems and structures, development of the Nuclear Island layout, and performing preliminary safety analyses to support design efforts. Incorporation of the EUR has been a key design requirement for the EP1000 form the beginning of the program. Detailed design solutions to meet the EUR have been defined and the safety approach has also been developed based on the EUR guidelines. The present paper describes the EP1000 approach to safety analysis and, in particular, to the Design Extension Conditions that, according to the EUR, represent the preferred method for giving consideration to the Complex Sequences and Severe Accidents at the design stage without including them in the design bases conditions. Preliminary results of some DEC analyses and an overview of the probabilistic safety assessment (PSA) are also presented. (author)

  20. Advanced designs of VVER reactor plant

    International Nuclear Information System (INIS)

    Mokhov, V.A.

    2010-01-01

    The history of VVER reactors, current challenges and approaches to the challenges are highlighted. The VVER-1200 reactor of 3+ generation for AES-2006 units are under construction at the Leningrad 2 nuclear power plant (LNPP-2). The main parameters are listed and details are presented of the vessel, steam generator, and improved fuel. The issue of the NPP safety is discussed. Additional topics include the MIR-1200 reactor unit, VVER-600, and VVER-SCP (Generation 4). (P.A.)

  1. Performance and safety design of the advanced liquid metal reactor

    International Nuclear Information System (INIS)

    Berglund, R.C.; Magee, P.M.; Boardman, C.E.; Gyorey, G.L.

    1991-01-01

    The Advanced Liquid Metal Reactor (ALMR) program led by General Electric is developing, under U.S. Department of Energy sponsorship, a conceptual design for an advanced sodium-cooled liquid metal reactor plant. This design is intended to improve the already excellent level of plant safety achieved by the nuclear power industry while at the same time providing significant reductions in plant construction and operating costs. In this paper, the plant design and performance are reviewed, with emphasis on the ALMR's unique passive design safety features and its capability to utilize as fuel the actinides in LWR spent fuel

  2. Gas core reactor power plants designed for low proliferation potential

    International Nuclear Information System (INIS)

    Lowry, L.L.

    1977-09-01

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

  3. Power maximization method for land-transportable fully passive lead–bismuth cooled small modular reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jaehyun, E-mail: chojh@kaeri.re.kr [Korea Atomic Energy Research Institute, 1405 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Shin, Yong-Hoon; Hwang, Il Soon [Seoul National University, Sillim-dong, Gwanak-gu, Seoul 151-742 (Korea, Republic of)

    2015-08-15

    Highlights: • The power maximization method for LBE natural circulation cooled SMRs was developed. • The two powers in view of neutronics and thermal-hydraulics were considered. • The limitations for designing of LBE natural circulation cooled SMRs were summarized. • The necessary conditions for safety shutdown in accidents were developed. • The maximized power in the case study is 206 MW thermal. - Abstract: Although current pressurized water reactors (PWRs) have significantly contributed to global energy supply, PWR technology has not been considered a trustworthy energy solution owing to its problems of spent nuclear fuels (SNFs), nuclear safety, and nuclear economy. In order to overcome these problems, a lead–bismuth eutectic (LBE) fully passive cooling small modular reactor (SMR) system is suggested. This technology can not only provide the solution for the problems of SNFs through the transmutation feature of the LBE coolant, but also strengthen safety and economy through the concept of natural circulation cooling SMRs. It is necessary to maximize the advantages, namely safety and economy, of this type of nuclear power plants for broader applications in the future. Accordingly, the objective of this study is to maximize the reactor core power while satisfying the limitations of shipping size, materials endurance, and criticality of a long-burning core as well as safety under beyond design basis events. To achieve these objectives, the design limitations of natural circulating LBE-cooling SMRs are derived. Then, the power maximization method is developed based on obtaining the design limitations. The results of this study are expected to contribute to the effectiveness of the reactor design stage by providing insights to designers, as well as by formulating methods for the power maximization of other types of SMRs.

  4. The safety characteristics of the HTR 500 reactor plant

    International Nuclear Information System (INIS)

    Wachholz, W.

    1987-01-01

    The HTR is a reactor having a passive safety. It is equipped with the usual active engineered safety systems in simplified form. Due to its inherent safety characteristics and the burst-safe prestressed concrete reactor vessel activity containment is ensured even without the effect of active safety systems. Even in the event of extremely hypothetical accidents the effect on the environment is low enough so that evacuation or relocation of the population is not required. Therefore large-scale damage of agricultural land and industrially used areas is safely ruled out. Thus the site selection for this type of reactor is not restricted i.e. an HTR can be constructed near industrial and urban center. (author)

  5. 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.

  6. The status of work in the USSR on using inherent self-protection features of fast reactors, of passive and active means of shutdown and decay heat removal system

    International Nuclear Information System (INIS)

    Buksha, Yu.K.

    1991-01-01

    Extensive studies on fast reactor safety, aimed to increase intrinsic safety features and introduce passive safety means, are under way in the USSR. Design of the BN-800 reactor core with a close-to-zero sodium void effect of reactivity has been developed, complementary reactivity control means, based on passive principles are being implemented. As a whole, after the Chernobyl accident, the preference is given to the 'passive' full proof methods of safety. This approach may possibly seem excessive and may result in some losses concerning reactor economic characteristics

  7. PWR type reactor plant

    International Nuclear Information System (INIS)

    Matsuoka, Tsuyoshi.

    1993-01-01

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

  8. Preapplication safety evaluation report for the Power Reactor Innovative Small Module (PRISM) liquid-metal reactor

    International Nuclear Information System (INIS)

    Donoghue, J.E.; Donohew, J.N.; Golub, G.R.; Kenneally, R.M.; Moore, P.B.; Sands, S.P.; Throm, E.D.; Wetzel, B.A.

    1994-02-01

    This preapplication safety evaluation report (PSER) presents the results of the preapplication desip review for die Power Reactor Innovative Small Module (PRISM) liquid-mew (sodium)-cooled reactor, Nuclear Regulatory Commission (NRC) Project No. 674. The PRISM conceptual desip was submitted by the US Department of Energy in accordance with the NRC's ''Statement of Policy for the Regulation of Advanced Nuclear Power Plants'' (51 Federal Register 24643). This policy provides for the early Commission review and interaction with designers and licensees. The PRISM reactor desip is a small, modular, pool-type, liquid-mew (sodium)-cooled reactor. The standard plant design consists of dim identical power blocks with a total electrical output rating of 1395 MWe- Each power block comprises three reactor modules, each with a thermal rating of 471 MWt. Each module is located in its own below-grade silo and is co to its own intermediate heat transport system and steam generator system. The reactors utilize a metallic-type fuel, a ternary alloy of U-Pu-Zr. The design includes passive reactor shutdown and passive decay heat removal features. The PSER is the NRC's preliminary evaluation of the safety features in the PRISM design, including the projected research and development programs required to support the design and the proposed testing needs. Because the NRC review was based on a conceptual design, the PSER did not result in an approval of the design. Instead it identified certain key safety issues, provided some guidance on applicable licensing criteria, assessed the adequacy of the preapplicant's research and development programs, and concluded that no obvious impediments to licensing the PRISM design had been identified

  9. Analysis of reactor trips involving balance-of-plant failures

    International Nuclear Information System (INIS)

    Seth, S.; Skinner, L.; Ettlinger, L.; Lay, R.

    1986-01-01

    The relatively high frequency of plant transients leading to reactor trips at nuclear power plants in the US is of economic and safety concern to the industry. A majority of such transients is due to failures in the balance-of-plant (BOP) systems. As a part of a study conducted for the US Nuclear Regulatory Commission, Mitre has carried out a further analysis of the BOP failures associated with reactor trips. The major objectives of the analysis were to examine plant-to-plant variations in BOP-related trips, to understand the causes of failures, and to determine the extent of any associated safety system challenges. The analysis was based on the Licensee Event Reports submitted on all commercial light water reactors during the 2-yr period, 1984-1985

  10. SBWR: A simplified boiling water reactor

    International Nuclear Information System (INIS)

    Duncan, J.D.; Sawyer, C.D.; Lagache, M.P.

    1987-01-01

    An advanced light water reactor concept is being developed for possible application in the 1990's. The concept, known as SBWR is a boiling water reactor which uses natural circulation to provide flow to the reactor core. In an emergency, a gravity driven core cooling system is used. The reactor is depressurized and water from an elevated suppression pool flows by gravity to the reactor vessel to keep the reactor core covered. The concept also features a passive containment cooling system in which water flows by gravity to cool the suppression pool wall. No operator action is required for a period of at least three days. Use of these and other passive systems allows the elimination of emergency diesel generators, core cooling pumps and heat removal pumps which is expected to simplify the plant design, reduce costs and simplify licensing. The concept is being developed by General Electric, Bechtel and the Massachusetts Institute of Technology supported by the Electric Power Research Institute and the United States Department of Energy in the United States. In Japan, The Japan Atomic Power Company has a great interest in this concept

  11. Enhanced thermal expansion control rod drive lines for improving passive safety of fast reactors

    International Nuclear Information System (INIS)

    Edelmann, M.; Baumann, W.; Kuechle, M.; Kussmaul, G.; Vaeth, W.; Bertram, A.

    1992-01-01

    The paper presents a device for increasing the thermal expansion effect of control rod drive lines on negative reactivity feedback in fast reactors. The enhanced thermal expansion of this device can be utilized for both passive rod drop and forced insertion of absorbers in unprotected transients, e.g. ULOF. In this way the reactor is automatically brought into a permanently subcritical state and temperatures are kept well below the boiling point of the coolant. A prototype of such a device called ATHENa (German: Shut-down by THermal Expansion of Na) is presently under construction and will be tested. The paper presents the principle, design features and thermal properties of ATHENs as well as results of reactor dynamics calculations of ULOF's for EFR with enhanced thermal expansion control rod drive lines. (author)

  12. The program of reactors and nuclear power plants

    International Nuclear Information System (INIS)

    Calabrese, Carlos R.

    2001-01-01

    Into de framework of the program of research reactors and nuclear power plants, the operating Argentine reactors are described. The uses of the research reactors in Argentina are summarized. The reactors installed by Argentina in other countries (Peru, Algeria, Egypt) are briefly described. The CAREM project for the design and construction of an innovator small power reactor (27 MWe) is also described in some detail. The next biennial research and development program for reactor is briefly outlined

  13. Progress on PRISM, an inherently safe, economic, and testable advanced liquid metal reactor

    International Nuclear Information System (INIS)

    Tippets, F.E.; Salerno, L.N.; Boardman, C.E.; Kwant, W.; Murata, R.E.; Snyder, C.R.

    1987-01-01

    This paper reports progress on the design of PRISM (Power Reactor Inherently Safe Module) under the DOE-sponsored innovative reactor program now in its third year at General Electric. The purpose of this program is to develop a design for an inherently safe, reliable, and marketable liquid metal fast reactor power plant. The PRISM design approach includes the following key elements: Compact sodium-cooled pool-type reactor modules that are sized to enable factory fabrication, economical shipment to inland as well as water-side sites, and economical full-scale prototype testing for design certification; Nuclear safety-related envelope limited to the reactor modules and their service systems; Inherent, passive shutdown heat removal for loss-of-cooling events; Inherent, passive reactivity shutdown for failure-to-scram events

  14. Interactive nuclear plant analyzer for VVER-440 reactor

    International Nuclear Information System (INIS)

    Shier, W.; Horak, W.; Kennett, R.

    1992-05-01

    This document discusses an interactive nuclear plant analyzer (NPA) which has been developed for a VVER-440, Model 213 reactor for use in the training of plant personnel, the development and verification of plant operating procedures, and in the analysis of various anticipated operational occurrences and accident scenarios. This NPA is operational on an IBM RISC-6000 workstation and utilizes the RELAP5/MOD2 computer code for the calculation of the VVER-440 reactor response to the interactive commands initiated by the NPA operator

  15. General aspects of CAREM-25 reactor

    International Nuclear Information System (INIS)

    Delmastro, Dario F.; Gomez, Silvia; Ishida, Viviana; Mazzi, Ruben; Santecchia, Alberto; Gomez de Soler, Susana M.

    2000-01-01

    CAREM project consists on the development and design of an advanced nuclear power plant. In order to verify its innovative features the construction of a prototype is planned. In this paper the main technical characteristics of CAREM-25 prototype reactor are presented. This is a very low power innovative reactor (100 M Wth) conceived with new generation design solutions. Based on an indirect cycle integrated light water reactor using enriched uranium, CAREM has some distinctive features that greatly simplify the reactor and also contribute to a high level of safety: -) Integrated primary system; -) Primary system cooling by natural convection; -) Self pressurization; -) and Passive safety systems. (author)

  16. Issues in risk analysis of passive LWR designs

    International Nuclear Information System (INIS)

    Youngblood, R.W.; Pratt, W.T.; Amico, P.J.; Gallagher, D.

    1992-01-01

    This paper discusses issues which bear on the question of how safety is to be demonstrated for ''simplified passive'' light water reactor (LWR) designs. First, a very simplified comparison is made between certain systems in today's plants. comparable systems in evolutionary designs, and comparable systems in the simplified passives. in order to introduce the issues. This discussion is not intended to describe the designs comprehensively, but is offered only to show why certain issues seem to be important in these particular designs. Next, an important class of accident sequences is described; finally, based on this discussion, some priorities in risk analysis are presented and discussed

  17. The program of reactors and nuclear power plants; Programa de reactores y centrales nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Calabrese, Carlos R [Comision Nacional de Energia Atomica, General San Martin (Argentina). Centro Atomico Constituyentes

    2001-07-01

    Into de framework of the program of research reactors and nuclear power plants, the operating Argentine reactors are described. The uses of the research reactors in Argentina are summarized. The reactors installed by Argentina in other countries (Peru, Algeria, Egypt) are briefly described. The CAREM project for the design and construction of an innovator small power reactor (27 MWe) is also described in some detail. The next biennial research and development program for reactor is briefly outlined.

  18. 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

  19. Application of the system engineering approach for reactor plants design

    International Nuclear Information System (INIS)

    Sitskiy, S. B.

    2010-01-01

    The main activities planned for to be implemented are: developing a data model of the reactor plant plus integration with the information model of the plant (3D model + P & ID); reengineering of processes, developing of electronic documents; description of the equipment for information management of the reactor plant lifecycle – according ISO15926

  20. The research activities on in-tube condensation in the presence of noncondensables for passive cooling applications

    Energy Technology Data Exchange (ETDEWEB)

    Tanrikut, A [Turkish Atomic Energy Authority, Ankara (Turkey)

    1996-12-01

    The introduction of nuclear power becomes an attractive solution to the problem of increasing demand for electricity power capacity in Turkey. Thus, Turkey is willing to follow the technological development trends in advanced reactor systems and to participate in joint research studies. The primary objectives of the passive design features are to simplify the design, which assures the minimized demand on operator, and to improve plant safety. To accomplish these features the operating principles of passive safety systems should be well understood by an experimental validation program. Such a validation program is also important for the assessment of advanced computer codes which are currently used for design and licensing procedures. The condensation mode of heat transfer plays an important role for the passive heat removal applications in the current nuclear power plants (e.g. decay heat removal via steam generators in case of loss of heat removal system) and advanced water-cooled reactor systems. But is well established that the presence of noncondensable gases can greatly inhibit the condensation process due to the build-up of noncondensable gas concentration at the liquid/gas interface. The isolation condenser of passive containment cooling system of the simplified boiling water reactors is a typical application area of in-tube condensation in the presence of noncondensable. This paper describes the research activities at the Turkish Atomic Energy Authority concerning condensation in the presence of air, as a noncondensable gas. (author). 9 refs, 6 figs.

  1. The research activities on in-tube condensation in the presence of noncondensables for passive cooling applications

    International Nuclear Information System (INIS)

    Tanrikut, A.

    1996-01-01

    The introduction of nuclear power becomes an attractive solution to the problem of increasing demand for electricity power capacity in Turkey. Thus, Turkey is willing to follow the technological development trends in advanced reactor systems and to participate in joint research studies. The primary objectives of the passive design features are to simplify the design, which assures the minimized demand on operator, and to improve plant safety. To accomplish these features the operating principles of passive safety systems should be well understood by an experimental validation program. Such a validation program is also important for the assessment of advanced computer codes which are currently used for design and licensing procedures. The condensation mode of heat transfer plays an important role for the passive heat removal applications in the current nuclear power plants (e.g. decay heat removal via steam generators in case of loss of heat removal system) and advanced water-cooled reactor systems. But is well established that the presence of noncondensable gases can greatly inhibit the condensation process due to the build-up of noncondensable gas concentration at the liquid/gas interface. The isolation condenser of passive containment cooling system of the simplified boiling water reactors is a typical application area of in-tube condensation in the presence of noncondensable. This paper describes the research activities at the Turkish Atomic Energy Authority concerning condensation in the presence of air, as a noncondensable gas. (author). 9 refs, 6 figs

  2. Soil Seed Bank and Plant Community Development in Passive Restoration of Degraded Sandy Grasslands

    Directory of Open Access Journals (Sweden)

    Renhui Miao

    2016-06-01

    Full Text Available To evaluate the efficacy of passive restoration on soil seed bank and vegetation recovery, we measured the species composition and density of the soil seed bank, as well as the species composition, density, coverage, and height of the extant vegetation in sites passively restored for 0, 4, 7, and 12 years (S0, S4, S7, and S12 in a degraded grassland in desert land. Compared with S0, three more species in the soil seed bank at depths of 0–30 cm and one more plant species in the community was detected in S12. Seed density within the topsoil (0–5 cm was five times higher in S12 than that in S0. Plant densities in S7 and S12 were triple and quadruple than that in S0. Plant coverage was increased by 1.5 times (S4, double (S7, and triple (S12 compared with S0. Sørensen’s index of similarity in species composition between the soil seed bank and the plant community were high (0.43–0.63, but it was lower in short-term restoration sites (S4 and S7 than that in no and long-term restoration sites (S0 and S12. The soil seed bank recovered more slowly than the plant community under passive restoration. Passive restoration is a useful method to recover the soil seed bank and vegetation in degraded grasslands.

  3. Preapplication safety evaluation report for the Power Reactor Innovative Small Module (PRISM) liquid-metal reactor. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Donoghue, J.E.; Donohew, J.N.; Golub, G.R.; Kenneally, R.M.; Moore, P.B.; Sands, S.P.; Throm, E.D.; Wetzel, B.A. [Nuclear Regulatory Commission, Washington, DC (United States). Associate Directorate for Advanced Reactors and License Renewal

    1994-02-01

    This preapplication safety evaluation report (PSER) presents the results of the preapplication desip review for die Power Reactor Innovative Small Module (PRISM) liquid-mew (sodium)-cooled reactor, Nuclear Regulatory Commission (NRC) Project No. 674. The PRISM conceptual desip was submitted by the US Department of Energy in accordance with the NRC`s ``Statement of Policy for the Regulation of Advanced Nuclear Power Plants`` (51 Federal Register 24643). This policy provides for the early Commission review and interaction with designers and licensees. The PRISM reactor desip is a small, modular, pool-type, liquid-mew (sodium)-cooled reactor. The standard plant design consists of dim identical power blocks with a total electrical output rating of 1395 MWe- Each power block comprises three reactor modules, each with a thermal rating of 471 MWt. Each module is located in its own below-grade silo and is co to its own intermediate heat transport system and steam generator system. The reactors utilize a metallic-type fuel, a ternary alloy of U-Pu-Zr. The design includes passive reactor shutdown and passive decay heat removal features. The PSER is the NRC`s preliminary evaluation of the safety features in the PRISM design, including the projected research and development programs required to support the design and the proposed testing needs. Because the NRC review was based on a conceptual design, the PSER did not result in an approval of the design. Instead it identified certain key safety issues, provided some guidance on applicable licensing criteria, assessed the adequacy of the preapplicant`s research and development programs, and concluded that no obvious impediments to licensing the PRISM design had been identified.

  4. Fuel management of mixed reactor type power plant systems

    International Nuclear Information System (INIS)

    Csom, Gyula

    1988-01-01

    Breeding gain in symbiotic nuclear power plant system consisting of both thermal and fast breeder reactors depends on the characteristics and the ratio of thermal and fast reactors. The composition of the symbiotic power plant systems was determined for equilibrium and plutonium deficient systems. According to natural uranium utilization, symbiotic power plant systems are not less efficient than the systems containing only fast breeders. Depleted uranium can be applied in both types of systems. Reprocessing demands of the symbiotic power plant sytems were determined. (V.N.) 23 figs.; 1 tab

  5. Comparing herbaceous plant communities in active and passive riparian restoration.

    Directory of Open Access Journals (Sweden)

    Elise S Gornish

    Full Text Available Understanding the efficacy of passive (reduction or cessation of environmental stress and active (typically involving planting or seeding restoration strategies is important for the design of successful revegetation of degraded riparian habitat, but studies explicitly comparing restoration outcomes are uncommon. We sampled the understory herbaceous plant community of 103 riparian sites varying in age since restoration (0 to 39 years and revegetation technique (active, passive, or none to compare the utility of different approaches on restoration success across sites. We found that landform type, percent shade, and summer flow helped explain differences in the understory functional community across all sites. In passively restored sites, grass and forb cover and richness were inversely related to site age, but in actively restored sites forb cover and richness were inversely related to site age. Native cover and richness were lower with passive restoration compared to active restoration. Invasive species cover and richness were not significantly different across sites. Although some of our results suggest that active restoration would best enhance native species in degraded riparian areas, this work also highlights some of the context-dependency that has been found to mediate restoration outcomes. For example, since the effects of passive restoration can be quite rapid, this approach might be more useful than active restoration in situations where rapid dominance of pioneer species is required to arrest major soil loss through erosion. As a result, we caution against labeling one restoration technique as better than another. Managers should identify ideal restoration outcomes in the context of historic and current site characteristics (as well as a range of acceptable alternative states and choose restoration approaches that best facilitate the achievement of revegetation goals.

  6. Balance of plant improvements for future reactor projects

    International Nuclear Information System (INIS)

    Hollingshaus, H.

    1987-01-01

    Many studies have shown that improvements in portions of the plant other than the reactor systems can yield large cost savings during the design, construction, and operation of future reactor power plants. This portion is defined as the Balance of Plant which includes virtually everything except the equipment furnished by the Nuclear Steam Supply System manufacturer. It normally includes the erection of the entire plant including the NSSS. Cost of BOP equipment, engineering and construction work is therefore most of the cost of the plant. Improvements in the BOP have been identified that will substantially reduce nuclear plant cost and construction time while at the same time increasing availability and operability and improving safety. Improvements achieved through standardizatoin, simplification, three-dimensional (3D) computer-aided design, modular construction, innovative construction techniques, and applications for Artificial Intelligence Systems are described. (author)

  7. Very High Efficiency Reactor (VHER) Concepts for Electrical Power Generation and Hydrogen Production

    International Nuclear Information System (INIS)

    PARMA JR, EDWARD J.; PICKARD, PAUL S.; SUO-ANTTILA, AHTI JORMA

    2003-01-01

    The goal of the Very High Efficiency Reactor study was to develop and analyze concepts for the next generation of nuclear power reactors. The next generation power reactor should be cost effective compared to current power generation plant, passively safe, and proliferation-resistant. High-temperature reactor systems allow higher electrical generating efficiencies and high-temperature process heat applications, such as thermo-chemical hydrogen production. The study focused on three concepts; one using molten salt coolant with a prismatic fuel-element geometry, the other two using high-pressure helium coolant with a prismatic fuel-element geometry and a fuel-pebble element design. Peak operating temperatures, passive-safety, decay heat removal, criticality, burnup, reactivity coefficients, and material issues were analyzed to determine the technical feasibility of each concept

  8. Use of passive systems to improve plant operation and maintenance

    International Nuclear Information System (INIS)

    Shah, D.

    2000-01-01

    In a deregulated future, a utility's strength will depend on its ability to be cost competitive in the marketplace. However, the competitive advantage of nuclear power will depend on each owner's ability to reduce Operating and Maintenance (O and M) costs without sacrificing nuclear safety. The use of passive systems (i.e., systems without any moving parts) can reduce plant O and M costs while increasing safety in nuclear power plants. (author)

  9. On reactor type comparisons for the next generation of reactors

    International Nuclear Information System (INIS)

    Alesso, H.P.; Majumdar, K.C.

    1991-01-01

    In this paper, we present a broad comparison of studies for a selected set of parameters for different nuclear reactor types including the next generation. This serves as an overview of key parameters which provide a semi-quantitative decision basis for selecting nuclear strategies. Out of a number of advanced reactor designs of the LWR type, gas cooled type, and FBR type, currently on the drawing board, the Advanced Light Water Reactors (ALWR) seem to have some edge over other types of the next generation of reactors for the near-term application. This is based on a number of attributes related to the benefit of the vast operating experience with LWRs coupled with an estimated low risk profile, economics of scale, degree of utilization of passive systems, simplification in the plant design and layout, modular fabrication and manufacturing. 32 refs., 1 fig., 3 tabs

  10. Overview of fast reactor safety research and development in the USA

    International Nuclear Information System (INIS)

    Neuhold, R.J.; Avery, R.; Marchaterre, J.F.

    1986-01-01

    The liquid metal reactor (LMR) safety R and D program in the U.S. is presently focused on support of two modular innovative reactor concepts: PRISM - the General Electric Power Reactor Inherently Safe Module and SAFR - the Rockwell International Sodium Advanced Fast Reactor. These reactor plant concepts accommodate the use of either oxide fuel or the metal fuel which is under development in the Argonne National Laboratory (ANL) Integral Fast Reactor (IFR) program. Both concepts emphasize prevention of accidents through enhancement of inherent and passive safety characteristics. Enhancement of these characteristics is expected to be a major factor in establishing new and improved safety criteria and licensing arrangements with regulatory authorities for advanced reactors. Limited work is also continuing on the Large Scale Prototype Breeder (LSPB), a large pool plant design. Major elements of the current and restructured safety program are discussed. (author)

  11. Staffing requirements for future small and medium reactors based on projections in the Russian Federation

    International Nuclear Information System (INIS)

    Antonovsky, G.M.; Kodochigov, N.G.; Kurachenkov, A.V.; Novikov, V.V.

    2001-01-01

    Experimental Design Bureau of Mechanical Engineering (OKBM) specializes in the development of small and medium power reactors having different purposes. They include reactor plants for NPHPP, nuclear district heating power plants and propulsion plants. Small and medium power plants have simpler processes of electricity and heat production, less systems, simpler control algorithms and considerably enhanced inherent safety properties. These plants are mainly equipped with passive safety systems. These properties are especially characteristic for reactor plants of nuclear district heating power plants and HTG reactor plants. The designs of small and medium power plants actually provide a high degree of control automation which considerably reduces workload on the personnel in both normal and abnormal operation conditions. All this allows the reduction in personnel for small and medium power reactors if compared to high capacity reactor plants. But due to objective reasons the specific number of personnel (man/MW) for average and especially small capacity reactors considerably exceeds the value for high capacity reactor plants. At the same time one can propose a set of organization - technical measures allowing the increase in this value in future. Safety requirements imposed for small and average capacity reactors are the same or more strict than those for high capacity reactors. That's why the requirements to the training of personnel for such reactor plants are not allowed to be lowered if compared to the requirements imposed to the personnel of high capacity reactors. (author)

  12. Large-scale experimental facility for emergency condition investigation of a new generation NPP WWER-640 reactor with passive safety systems

    Energy Technology Data Exchange (ETDEWEB)

    Aniskevich, Y.N.; Vasilenko, V.A.; Zasukha, V.K.; Migrov, Y.A.; Khabensky, V.B. [Research Inst. of Technology NITI (Russian Federation)

    1997-12-31

    The creation of the large-scale integral experimental facility (KMS) is specified by the programme of the experimental investigations to justify the engineering decisions on the safety of the design of the new generation NPP with the reactor WWER-640. The construction of KMS in a full volume will allow to conduct experimental investigations of all physical phenomena and processes, practically, occurring during the accidents on the NPPs with the reactor of WWER type and including the heat - mass exchange processes with low rates of the coolant, which is typical during the utilization of the passive safety systems, process during the accidents with a large leak, and also the complex intercommunicated processes in the reactor unit, passive safety systems and in the containment with the condition of long-term heat removal to the final absorber. KMS is being constructed at the Research Institute of Technology (NITI), Sosnovy Bor, Leningrad region, Russia. (orig.). 5 refs.

  13. Large-scale experimental facility for emergency condition investigation of a new generation NPP WWER-640 reactor with passive safety systems

    International Nuclear Information System (INIS)

    Aniskevich, Y.N.; Vasilenko, V.A.; Zasukha, V.K.; Migrov, Y.A.; Khabensky, V.B.

    1997-01-01

    The creation of the large-scale integral experimental facility (KMS) is specified by the programme of the experimental investigations to justify the engineering decisions on the safety of the design of the new generation NPP with the reactor WWER-640. The construction of KMS in a full volume will allow to conduct experimental investigations of all physical phenomena and processes, practically, occurring during the accidents on the NPPs with the reactor of WWER type and including the heat - mass exchange processes with low rates of the coolant, which is typical during the utilization of the passive safety systems, process during the accidents with a large leak, and also the complex intercommunicated processes in the reactor unit, passive safety systems and in the containment with the condition of long-term heat removal to the final absorber. KMS is being constructed at the Research Institute of Technology (NITI), Sosnovy Bor, Leningrad region, Russia. (orig.)

  14. Large-scale experimental facility for emergency condition investigation of a new generation NPP WWER-640 reactor with passive safety systems

    Energy Technology Data Exchange (ETDEWEB)

    Aniskevich, Y N; Vasilenko, V A; Zasukha, V K; Migrov, Y A; Khabensky, V B [Research Inst. of Technology NITI (Russian Federation)

    1998-12-31

    The creation of the large-scale integral experimental facility (KMS) is specified by the programme of the experimental investigations to justify the engineering decisions on the safety of the design of the new generation NPP with the reactor WWER-640. The construction of KMS in a full volume will allow to conduct experimental investigations of all physical phenomena and processes, practically, occurring during the accidents on the NPPs with the reactor of WWER type and including the heat - mass exchange processes with low rates of the coolant, which is typical during the utilization of the passive safety systems, process during the accidents with a large leak, and also the complex intercommunicated processes in the reactor unit, passive safety systems and in the containment with the condition of long-term heat removal to the final absorber. KMS is being constructed at the Research Institute of Technology (NITI), Sosnovy Bor, Leningrad region, Russia. (orig.). 5 refs.

  15. Passive devices of a reactor stop: classification of the characteristics and estimation of perfection degree

    International Nuclear Information System (INIS)

    Portyanoj, A.G.; Serdun', E.N.; Sorokin, A.P.; Egorov, V.S.; Shkarovskij, D.A.

    1998-01-01

    The perspective direction in NPP safety improvement connected with development of passive devices for nuclear reactor emergency shutdown (PDRS) is discussed. More than hundred devices which can fulfil the PDRS functions are suggested nowadays. The analysis of PDRS designing status as applicable for the fast reactors in the main which are based on the physical effect used in an element sensitive to temperature is made. The complex consisting of nine general characteristics including passive character, thresholdness, forces generation, inertia, multichannel design, stability towards operational factors, safety at failures, simplicity and visualisation, development conditions, is suggested for estimation of the quality of PDRS of different types. Basing on expert assessments realized using the complex of general characteristics it is shown that the types of PDRS may be separated into following three groups: linear expansion of solid bodies and thermoelectric ones (K ≅ 0.45); magnet ones with shape memory effect, liquid volume expansion (K ≅ 0.6); fusing ones (K ≅ 0.7). The conclusion is made that PDRS on the basis of fusing devices of the sulphon type with liofobic capillary-porous working body most completely satisfy the complex of general characteristics considered

  16. An Overview of the Safety Case for Small Modular Reactors

    International Nuclear Information System (INIS)

    Ingersoll, Daniel T.

    2011-01-01

    Several small modular reactor (SMR) designs emerged in the late 1970s and early 1980s in response to lessons learned from the many technical and operational challenges of the large Generation II light-water reactors. After the accident at the Three Mile Island plant in 1979, an ensuing reactor redesign effort spawned the term inherently safe designs, which later evolved into passively safe terminology. Several new designs were engineered to be deliberately small in order to fully exploit the benefits of passive safety. Today, new SMR designs are emerging with a similar philosophy of offering highly robust and resilient designs with increased safety margins. Additionally, because these contemporary designs are being developed subsequent to the September 11, 2001, terrorist attack, they incorporate a number of intrinsic design features to further strengthen their safety and security. Several SMR designs are being developed in the United States spanning the full spectrum of reactor technologies, including water-, gas-, and liquid-metal-cooled ones. Despite a number of design differences, most of these designs share a common set of design principles to enhance plant safety and robustness, such as eliminating plant design vulnerabilities where possible, reducing accident probabilities, and mitigating accident consequences. An important consequence of the added resilience provided by these design approaches is that the individual reactor units and the entire plant should be able to survive a broader range of extreme conditions. This will enable them to not only ensure the safety of the general public but also help protect the investment of the owner and continued availability of the power-generating asset. Examples of typical SMR design features and their implications for improved plant safety are given for specific SMR designs being developed in the United States.

  17. Nuclear reactor plant

    International Nuclear Information System (INIS)

    Schabert, H.P.; Laurer, E.

    1977-01-01

    The invention is concerned with a quick-closing valve on the main-steam pipe of a nuclear reactor plant. The quick-closing valve serves as isolating valve and as safety valve permitting depressurization in case of an accident. For normal operation a tube-shaped gate valve is provided as valve disc, enclosing an auxiliary valve disc to be used in case of accidents and which is opened at increased pressure to provide a smaller flow cross-section. The design features are described in detail. (RW) [de

  18. Potential of light water reactors for future nuclear power plants

    International Nuclear Information System (INIS)

    Gueldner, R.

    2003-01-01

    Energy consumption worldwide is going to increase further in the next few decades. Reliable supplies of electricity can be achieved only by centralized power plant structures. In this scenario, nuclear power plants are going to play a leading role as reliable and competitive plants, also under deregulated market conditions. Today, light water reactors have achieved a leading position, both technically and economically, contributing 85% to worldwide electricity generation in nuclear plants. They will continue to be a proven technology in power generation. In many countries, activities therefore are concentrated on extending the service life of plants beyond a period of forty years. New nuclear generating capacities are expected to be created and added from the end of this decade onward. Most of this capacity will be in light water reactors. The concepts of third-generation reactors will meet all economic and technical safety requirements of the 21st century and will offer considerable potential for further development. Probably some thirty years from now, fourth-generation nuclear power plants will be ready for commercial application. These plants will penetrate especially new sectors of the energy markets. Public acceptance of new nuclear power plants is not a matter of reactor lines, provided that safety requirements are met. The important issue is the management of radioactive waste. The construction of new nuclear power plants in Western Europe and North America mainly hinges on the ability to explain to the public that there is a need for new plants and that nuclear power is fundamental to assuring sustainable development. (orig.)

  19. A passive decay heat removal strategy of the integrated passive safety system (IPSS) for SBO combined with LOCA

    International Nuclear Information System (INIS)

    Kim, Sang Ho; Chang, Soon Heung; Choi, Yu Jung; Jeong, Yong Hoon

    2015-01-01

    Highlights: • A new PDHR strategy is proposed to cope with SBO-combined accidents. • The concept of integrated passive safety system (IPSS) is used in this strategy. • This strategy performs the functions of passive safety injection and SG gravity injection. • LOCAs in SBO are classified by the pressures in reactor coolant system for passive functions. • The strategy can be integrated with EOP and SAMG as a complementary strategy for ensuring safety. - Abstract: An integrated passive safety system (IPSS), to be achieved by the use of a large water tank placed at high elevation outside the containment, was proposed to achieve various passive functions. These include decay heat removal, safety injection, containment cooling, in-vessel retention through external reactor vessel cooling, and containment filtered venting. The purpose of the passive decay heat removal (PDHR) strategy using the IPSS is to cope with SBO and SBO-combined accidents under the assumption that existing engineered safety features have failed. In this paper, a PDHR strategy was developed based on the design and accident management strategy of Korean representative PWR, the OPR1000. The functions of a steam generator gravity injection and a passive safety injection system in the IPSS with safety depressurization systems were included in the PDHR strategy. Because the inadvertent opening of pressurizer valves and seal water leakage from RCPs could cause a loss of coolant in an SBO, LOCAs during a SBO were simulated to verify the performance of the strategy. The failure of active safety injection in LOCAs could also be covered by this strategy. Although LOCAs have generally been categorized according to their equivalent break diameters, the RCS pressure is used to classify the LOCAs during SBOs. The criteria values for categorization were determined from the proposed systems, which could maintain a reactor in a safe state by removing the decay heat for the SBO coping time of 8 h. The

  20. A passive decay heat removal strategy of the integrated passive safety system (IPSS) for SBO combined with LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sang Ho [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 291, Daehak-ro, Yuseong-gu, Daejeon 34141 (Korea, Republic of); Chang, Soon Heung [Handong Global University, 558, Handong-ro, Buk-gu, Pohang Gyeongbuk 37554 (Korea, Republic of); Choi, Yu Jung [Korea Hydro and Nuclear Power Co.—Central Research Institute, 70, 1312-gil, Yuseong-daero, Yuseong-gu, Daejeon 34101 (Korea, Republic of); Jeong, Yong Hoon, E-mail: jeongyh@kaist.ac.kr [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 291, Daehak-ro, Yuseong-gu, Daejeon 34141 (Korea, Republic of)

    2015-12-15

    Highlights: • A new PDHR strategy is proposed to cope with SBO-combined accidents. • The concept of integrated passive safety system (IPSS) is used in this strategy. • This strategy performs the functions of passive safety injection and SG gravity injection. • LOCAs in SBO are classified by the pressures in reactor coolant system for passive functions. • The strategy can be integrated with EOP and SAMG as a complementary strategy for ensuring safety. - Abstract: An integrated passive safety system (IPSS), to be achieved by the use of a large water tank placed at high elevation outside the containment, was proposed to achieve various passive functions. These include decay heat removal, safety injection, containment cooling, in-vessel retention through external reactor vessel cooling, and containment filtered venting. The purpose of the passive decay heat removal (PDHR) strategy using the IPSS is to cope with SBO and SBO-combined accidents under the assumption that existing engineered safety features have failed. In this paper, a PDHR strategy was developed based on the design and accident management strategy of Korean representative PWR, the OPR1000. The functions of a steam generator gravity injection and a passive safety injection system in the IPSS with safety depressurization systems were included in the PDHR strategy. Because the inadvertent opening of pressurizer valves and seal water leakage from RCPs could cause a loss of coolant in an SBO, LOCAs during a SBO were simulated to verify the performance of the strategy. The failure of active safety injection in LOCAs could also be covered by this strategy. Although LOCAs have generally been categorized according to their equivalent break diameters, the RCS pressure is used to classify the LOCAs during SBOs. The criteria values for categorization were determined from the proposed systems, which could maintain a reactor in a safe state by removing the decay heat for the SBO coping time of 8 h. The

  1. Argentinean integrated small reactor design and scale economy analysis of integrated reactor

    International Nuclear Information System (INIS)

    Florido, P. C.; Bergallo, J. E.; Ishida, M. V.

    2000-01-01

    This paper describes the design of CAREM, which is Argentinean integrated small reactor project and the scale economy analysis results of integrated reactor. CAREM project consists on the development, design and construction of a small nuclear power plant. CAREM is an advanced reactor conceived with new generation design solutions and standing on the large experience accumulated in the safe operation of Light Water Reactors. The CAREM is an indirect cycle reactor with some distinctive and characteristic features that greatly simplify the reactor and also contribute to a highly level of safety: integrated primary cooling system, self pressurized, primary cooling by natural circulation and safety system relying on passive features. For a fully doupled economic evaluation of integrated reactors done by IREP (Integrated Reactor Evaluation Program) code transferred to IAEA, CAREM have been used as a reference point. The results shows that integrated reactors become competitive with power larger than 200MWe with Argentinean cheapest electricity option. Due to reactor pressure vessel construction limit, low pressure drop steam generator are used to reach power output of 200MWe for natural circulation. For forced circulation, 300MWe can be achieved. (author)

  2. In-plant considerations for optimal offsite response to reactor accidents

    International Nuclear Information System (INIS)

    Burke, R.P.; Heising, C.D.; Aldrich, D.C.

    1982-11-01

    Offsite response decision-making methods based on in-plant conditions are developed for use during severe reactor-accident situations. Dose projections are used to eliminate all LWR plant systems except the reactor core and the spent-fuel storage pool from consideration for immediate offsite emergency response during accident situations. A simple plant information-management scheme is developed for use in offsite response decision-making. Detailed consequence calculations performed with the CRAC2 model are used to determine the appropriate timing of offsite-response implementation for a range of PWR accidents involving the reactor core. In-plant decision criteria for offsite-response implementation are defined. The definition of decision criteria is based on consideration of core-accident physical processes, in-plant accident monitoring information, and results of consequence calculations performed to determine the effectiveness of various public-protective measures. The benefits and negative aspects of the proposed response-implementation criteria are detailed

  3. General description of advanced heavy water reactor

    International Nuclear Information System (INIS)

    Kakodkar, A.; Sinha, R.K.; Dhawan, M.L.

    1999-01-01

    Advanced Heavy Water Reactor is a boiling light water cooled, heavy water moderated and vertical pressure tube type reactor with its design optimised for utilisation of thorium for power generation. The core consists of (Th-U 233 )O 2 and (Th-Pu)O 2 fuel with a discharge burn up of 20,000 MWd/Te. This reactor incorporates several features to simplify the design, which eliminate certain systems and components. AHWR design is also optimised for easy replaceability of coolant channels, facilitation of in-service inspection and maintenance and ease of erection. The AHWR design also incorporates several passive systems for performing safety-related functions in the event of an accident. In case of LOCA, emergency coolant is injected through 4 accumulators of 260 m 3 capacity directly into the core. Gravity driven water pool of capacity 6000 m 3 serves to cool the core for 3 days without operator's intervention. Core submergence, passive containment isolation and passive containment cooling are the added features in AHWR. The paper describes the various process systems, core and fuel design, primary components and safety concepts of AHWR. Plant layout and technical data are also presented. The conceptual design of the reactor has been completed, and the detailed design and development is scheduled for completion in the year 2002. (author)

  4. Gas Reactor International Cooperative program. Pebble bed reactor plant: screening evaluation. Volume 3. Appendix A. Equipment list

    International Nuclear Information System (INIS)

    1979-11-01

    This report consists of three volumes which describe the design concepts and screening evaluation for a 3000 MW(t) Pebble Bed Reactor Multiplex Plant (PBR-MX). The Multiplex plant produces both electricity and transportable chemical energy via the thermochemical pipeline (TCP). The evaluation was limited to a direct cycle plant which has the steam generators and steam reformers in the primary circuit. Volume 1 reports the overall plant and reactor system and was prepared by the General Electric Company. Core scoping studies were performed which evaluated the effects of annular and cylindrical core configurations, radial blanket zones, burnup, and ball heavy metal loadings. The reactor system, including the PCRV, was investigated for both the annular and cylindrical core configurations. Volume 3 is an Appendix containing the equipment list for the plant and was also prepared by United Engineers and Constructors, Inc. It tabulates the major components of the plant and describes each in terms of quantity, type, orientation, etc., to provide a basis for cost estimation

  5. PRISM: An innovative liquid metal fast breeder reactor

    International Nuclear Information System (INIS)

    Kruger, G.B.; Boardman, C.E.; Olich, E.E.; Switick, D.M.

    1986-01-01

    This paper describes an innovative sodium-cooled reactor concept employing small certified reactor modules coupled with a standardized steam generator system. The total plant employs nine PRISM reactors (power reactor inherently safe module) in three 415 MWe power blocks. The PRISM design concept utilizes inherent safety characteristics and modularity to improve licensability, reduce owner's risk, and reduce costs. The relatively small size of each reactor module facilitates the use of passive, inherent self-shutdown and shutdown heat removal features, which permit design simplification and reduction of safety-related systems. It is proposed that a single PRISM module be used in a full-scale integrated reactor safety test. Results from the test would be used to obtain NRC certification of the standard design

  6. Experimental development of power reactor intelligent control

    International Nuclear Information System (INIS)

    Edwards, R.M.; Garcia, H.E.; Lee, K.Y.

    1992-01-01

    The US nuclear utility industry initiated an ambitious program to modernize the control systems at a minimum of ten existing nuclear power plants by the year 2000. That program addresses urgent needs to replace obsolete instrumentation and analog controls with highly reliable state-of-the-art computer-based digital systems. Large increases in functionality that could theoretically be achieved in a distributed digital control system are not an initial priority in the industry program but could be logically considered in later phases. This paper discusses the initial development of an experimental sequence for developing, testing, and verifying intelligent fault-accommodating control for commercial nuclear power plant application. The sequence includes an ultra-safe university research reactor (TRIGA) and a passively safe experimental power plant (Experimental Breeder Reactor 2)

  7. The CEA/DRN innovative R and D programme: significant studies on passive systems

    International Nuclear Information System (INIS)

    Fiorini, G.L.; Magistris, F. de; Dumaz, P.; Gautier, G.M.; Pignatel, J.F.; Richard, P.

    1999-01-01

    The work on passive systems is an essential item of the R and D programme for future reactors; it is structured following four main guidelines: Research, and validation of innovative solutions for the safety functions achievement; An enlarged assessment of the performances of passive systems; Extension of the data base and of the tools qualification range; Assessment of new plant operation modes. After a recalling on the general framework, the paper describes, following these guidelines, the status of the art of the main corresponding programmes within CEA/DRN. (author)

  8. Advanced High Temperature Reactor Systems and Economic Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Peretz, Fred J [ORNL; Qualls, A L [ORNL

    2011-09-01

    The Advanced High Temperature Reactor (AHTR) is a design concept for a large-output [3400 MW(t)] fluoride-salt-cooled high-temperature reactor (FHR). FHRs, by definition, feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The AHTR's large thermal output enables direct comparison of its performance and requirements with other high output reactor concepts. As high-temperature plants, FHRs can support either high-efficiency electricity generation or industrial process heat production. The AHTR analysis presented in this report is limited to the electricity generation mission. FHRs, in principle, have the potential to be low-cost electricity producers while maintaining full passive safety. However, no FHR has been built, and no FHR design has reached the stage of maturity where realistic economic analysis can be performed. The system design effort described in this report represents early steps along the design path toward being able to predict the cost and performance characteristics of the AHTR as well as toward being able to identify the technology developments necessary to build an FHR power plant. While FHRs represent a distinct reactor class, they inherit desirable attributes from other thermal power plants whose characteristics can be studied to provide general guidance on plant configuration, anticipated performance, and costs. Molten salt reactors provide experience on the materials, procedures, and components necessary to use liquid fluoride salts. Liquid metal reactors provide design experience on using low-pressure liquid coolants, passive decay heat removal, and hot refueling. High temperature gas-cooled reactors provide experience with coated particle fuel and graphite components. Light water reactors (LWRs) show the potentials of transparent, high-heat capacity coolants with low chemical reactivity. Modern coal-fired power plants provide design experience

  9. Design and analysis of new prestressed concrete containment and its passive cooling system for nuclear power plants

    International Nuclear Information System (INIS)

    Tan Xiaoshi; Li Xiaowei; Li Xiaotian; He Shuyan

    2014-01-01

    A new nuclear power plant prestressed concrete containment and its passive cooling system design were proposed for CAP1700 nuclear power plant as an example. The thermal-hydraulic calculation method for the new passive containment cooling system of CAP1700 was introduced and the operating parameters in accident condition were obtained. The result shows that the design of passive containment cooling system for CAP1700 is feasible and can meet the cooling demand in accident condition. Reservoir capacity of tank has a big margin and can be further optimized by calculation. (authors)

  10. The role of passive and inherent safety properties in Siemens/KWU nuclear power plants

    International Nuclear Information System (INIS)

    Gremm, O.

    1990-01-01

    In Siemens/KWU Nuclear Power Plants the applied safety concept consist of a well balanced combination of active, passive use well is inherent safety measures. In principle it is not possible to realise a safety concept exclusively with inherent and/or passive safety properties. The respective measures and arguments will be explained in detail in the presentation. In addition the Siemens/KWU safety concept with examples of the role of inherent and passive safety measures will be illustrated. (author). 9 refs, 9 figs

  11. Advanced liquid metal reactor plant control system

    International Nuclear Information System (INIS)

    Dayal, Y.; Wagner, W.; Zizzo, D.; Carroll, D.

    1993-01-01

    The modular Advanced Liquid Metal Reactor (ALMR) power plant is controlled by an advanced state-of-the-art control system designed to facilitate plant operation, optimize availability, and protect plant investment. The control system features a high degree of automatic control and extensive amount of on-line diagnostics and operator aids. It can be built with today's control technology, and has the flexibility of adding new features that benefit plant operation and reduce O ampersand M costs as the technology matures

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

    International Nuclear Information System (INIS)

    Sarkulov, M.K.

    2015-01-01

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

  13. Advanced nuclear reactor safety design technology research in NPIC

    International Nuclear Information System (INIS)

    Yu, H.

    2014-01-01

    After the Fukushima accident happen, Nuclear Power Plants (NPPs) construction has been suspended in China for a time. Now the new regulatory rule has been proposed that the most advanced safety standard must be adopted for the new NPPs and practical elimination of large fission product release by design during the next five plans period. So the advanced reactor research is developing in China. NPIC is engaging on the ACP1000 and ACP100 (Small Module Reactor) design. The main design character will be introduced in this paper. The Passive Combined with Active (PCWA) design was adopted during the ACP1000 design to reduce the core damage frequency (CDF); the Cavity Injection System (CIS) is design to mitigation the consequence of the severe accident. Advance passive safety system was designed to ensure the long term residual heat removal during the Small Module Reactor (SMR). The SMR will be utilized to be the floating reactors, district heating reactor and so on. Besides, the Science and Technology on Reactor System Design Technology Laboratory (LRSDT) also engaged on the fundamental thermal-hydraulic characteristic research in support of the system validation. (author)

  14. Safety analysis of an expert reactor protection system in nuclear power plants

    International Nuclear Information System (INIS)

    EL-Kafas, A.E.A.E.

    1996-01-01

    the purpose of the dissertation is to develop a real time expert reactor protection system (ERPS) for operational safety of pressurized water reactor nuclear power plant. The system is developed to diagnose plant failures and for identification of plant transients (with and without scram). for this ERPS. probabilistic safety analysis techniques are used to check the availability and priority of the recommended safety system in case of plant accidents . the real- time information during transients and accidents can be obtained to asses the operator in his decision - making . Also, the ERPS is able to give advice for the reactor operator to take the appropriate corrective action during abnormal situations. The system model consists of the dynamic differential equations for reactor core, pressurizer, steam generator, turbine and generator, piping and plenums. The system of equations can be solved by appropriate codes also displayed directly from sensors of the plant. All scenarios of transients, accidents and fault tress for plant systems are learned to ERPS

  15. Advanced power reactors with improved safety characteristics

    International Nuclear Information System (INIS)

    Birkhofer, A.

    1994-01-01

    The primary objective of nuclear safety is the protection of individuals, society and environment against radiological hazards from accidental releases of radioactive materials contained in nuclear reactors. Hereto, these materials are enclosed by several successive barriers and the barriers protected against mishaps and accidents by a multi-level system of safety precautions. The evolution of reactor technology continuously improves this concept and its implementation. At a world-wide scale, several advanced reactor concepts are currently being considered, some of them already at a design stage. Essential safety objectives include both further strengthening the prevention of accidents and improving the containment of fission products should an accident occur. The proposed solutions differ considerably with regard to technical principles, plant size and time scales considered for industrial application. Two typical approaches can be distinguished: The first approach basically aims at an evolution of power reactors currently in use, taking into account the findings from safety research and from operation of current plants. This approach makes maximum use of proven technology and operating experience but may nevertheless include new safety features. The corresponding designs are often termed 'large evolutionary'. The second approach consists in more fundamental changes compared to present designs, often with strong emphasis on specific passive features protecting the fuel and fuel cladding barriers. Owing to the nature and capability of those passive features such 'innovative designs' are mostly smaller in power output. The paper describes the basic objectives of such developments and illustrates important technical concepts focusing on next generation plants, i.e. designs to be available for industrial application until the end of this decade. 1 tab. (author)

  16. Regulatory Considerations for the Long Term Cooling Safe Shutdown Requirements of the Passive Residual Heat Removal Systems in Advanced Reactors

    International Nuclear Information System (INIS)

    Sim, S. K.; Bae, S. H.; Kim, Y. S.; Hwang, Min Jeong; Bang, Young Seok; Hwang, Taesuk

    2016-01-01

    USNRC approved safe shutdown at 215.6 .deg. C for a safe and long term cooling state for the redundant passive RHRSs by SECY-94-084. USNRC issued COLA(Combined Construction and Operating License) for the Levy County NP Unit-1/2 for the AP1000 passive RHRSs in 2014. Korea Hydro and Nuclear Power(KHNP) is developing APR+ and adopted Passive Auxiliary Feedwater System(PAFS) as a new passive RHRS design. Korea Institute of Nuclear Safety(KINS) has been developing regulatory guides for the advanced safety design features of the advanced ALWRs which has plan to construct in near future in Korea[5]. Safety and regulatory issues as well as the safe shut down requirements of the passive RHRS are discussed and considerations in developing regulatory guides for the passive RHRS are presented herein. Passive RHRSs have been introduced as new safety design features for the advanced reactors under development in Korea. These passive RHRSs have potential advantages over existing active RHRS, however, their functions are limited due to inherent ability of passive heat removal processes. It is high time to evaluate the performance of the passive PRHRs and develop regulatory guides for the safety as well as the performance analyses of the passive RHRS

  17. Liquid Metal Fast Breeder Reactor plant maintenance and equipment design

    International Nuclear Information System (INIS)

    Swannack, D.L.

    1982-01-01

    This paper provides a summary of maintenance equipment considerations and actual plant handling experiences from operation of a sodium-cooled reactor, the Fast Flux Test Facility (FFTF). Equipment areas relating to design, repair techniques, in-cell handling, logistics and facility services are discussed. Plant design must make provisions for handling and replacement of components within containment or allow for transport to an ex-containment area for repair. The modular cask assemblies and transporter systems developed for FFTF can service major plant components as well as smaller units. The plant and equipment designs for the Clinch River Breeder Reactor (CRBR) plant have been patterned after successful FFTF equipment

  18. Impact of Passive Safety on FHR Instrumentation Systems Design and Classification

    International Nuclear Information System (INIS)

    Holcomb, David Eugene

    2015-01-01

    Fluoride salt-cooled high-temperature reactors (FHRs) will rely more extensively on passive safety than earlier reactor classes. 10CFR50 Appendix A, General Design Criteria for Nuclear Power Plants, establishes minimum design requirements to provide reasonable assurance of adequate safety. 10CFR50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors, provides guidance on how the safety significance of systems, structures, and components (SSCs) should be reflected in their regulatory treatment. The Nuclear Energy Institute (NEI) has provided 10 CFR 50.69 SSC Categorization Guideline (NEI-00-04) that factors in probabilistic risk assessment (PRA) model insights, as well as deterministic insights, through an integrated decision-making panel. Employing the PRA to inform deterministic requirements enables an appropriately balanced, technically sound categorization to be established. No FHR currently has an adequate PRA or set of design basis accidents to enable establishing the safety classification of its SSCs. While all SSCs used to comply with the general design criteria (GDCs) will be safety related, the intent is to limit the instrumentation risk significance through effective design and reliance on inherent passive safety characteristics. For example, FHRs have no safety-significant temperature threshold phenomena, thus enabling the primary and reserve reactivity control systems required by GDC 26 to be passively, thermally triggered at temperatures well below those for which core or primary coolant boundary damage would occur. Moreover, the passive thermal triggering of the primary and reserve shutdown systems may relegate the control rod drive motors to the control system, substantially decreasing the amount of safety-significant wiring needed. Similarly, FHR decay heat removal systems are intended to be running continuously to minimize the amount of safety-significant instrumentation needed to initiate

  19. Design study of plant system for the fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    Iida, Hiromasa; Kuroda, Hideo; Yamada, Masao; Suzuki, Tatsushi; Honda, Tsutomu; Ohmura, Hiroshi; Itoh, Shinichi.

    1986-11-01

    This report describes design study results of the FER plant system. The purpose of this study is to have an image of the FER plant system as a whole by designing major auxiliary systems, reactor building and maintenance and radwaste desposal systems. The major auxiliary systems include tritium, cooling, evacuation and fueling systems. For these each systems, flowdiagrams are studied and designs of devices and pipings are conducted. In the reactor building design, layout of the above auxiliary systems in the building is studied with careful zoning concept by the radiation level. Structural integrity of the reactor building is also studied including seismic analysis. In the design of the maintenance and radwaste system flowdiagram of failed reactor components is developed and transfer vehicles and buildings are designed. Finally assuming JAERI Naka site as the reactor site layout of the whole FER plant system is developed. (author)

  20. Integrated gasification gas combined cycle plant with membrane reactors: Technological and economical analysis

    International Nuclear Information System (INIS)

    Amelio, Mario; Morrone, Pietropaolo; Gallucci, Fausto; Basile, Angelo

    2007-01-01

    In the present work, the capture and storage of carbon dioxide from the fossil fuel power plant have been considered. The main objective was to analyze the thermodynamic performances and the technological aspects of two integrated gasification gas combined cycle plants (IGCC), as well as to give a forecast of the investment costs for the plants and the resulting energy consumptions. The first plant considered is an IGCC* plant (integrated gasification gas combined cycle plant with traditional shift reactors) characterized by the traditional water gas shift reactors and a CO 2 physical adsorption system followed by the power section. The second one is an IGCC M plant (integrated gasification gas combined cycle plant with membrane reactor) where the coal thermal input is the same as the first one, but the traditional shift reactors and the physical adsorption unit are replaced by catalytic palladium membrane reactors (CMR). In the present work, a mono-dimensional computational model of the membrane reactor was proposed to simulate and evaluate the capability of the IGCC M plant to capture carbon dioxide. The energetic performances, efficiency and net power of the IGCC* and IGCC M plants were, thus, compared, assuming as standard a traditional IGCC plant without carbon dioxide capture. The economical aspects of the three plants were compared through an economical analysis. Since the IGCC* and IGCC M plants have additional costs related to the capture and disposal of the carbon dioxide, a Carbon Tax (adopted in some countries like Sweden) proportional to the number of kilograms of carbon dioxide released in the environment was assumed. According to the economical analysis, the IGCC M plant proved to be more convenient than the IGCC* one

  1. Perspective of nuclear energy and advanced reactors

    International Nuclear Information System (INIS)

    Lopez Jimenez, J.; Cobian, J.

    2007-01-01

    Future nuclear energy growth will be the result of the contributions of every single plant being constructed or projected at present as it is connected to the grid. As per IAEA, there exists presently 34 nuclear power plants under construction 81 with the necessary permits and funding and 223 proposed, which are plants seriously pursuing permits and financing. This means that in a few decades the number of nuclear power plants in operation will have doubled. This growth rate is characterised by the incorporation of new countries to the nuclear club and the gradually increasing importance of Asian countries. During this expansive phase, generation III and III+designs are or will be used. These designs incorporate the experience from operating plants, and introduce innovations on rationalization design efficiency and safety, with emphasis on passive safety features. In a posterior phase, generation IV designs, presently under development, will be employed. Generation IV consists of several types of reactors (fast reactors, very high temperature reactors, etc), which will improve further sustain ability, economy, safety and reliability concepts. The described situation seems to lead to a renaissance of the nuclear energy to levels hardly thinkable several years ago. (Author)

  2. Chaotic behavior of water column oscillator simulating pressure balanced injection system in passive safety reactor

    International Nuclear Information System (INIS)

    Morimoto, Y.; Madarame, H.; Okamoto, K.

    2001-01-01

    Japan Atomic Energy Research Institute (JAERI) proposed a passive safety reactor called the System-integrated Pressurized Water Reactor (SPWR). In a loss of coolant accident, the Pressurizing Line (PL) and the Injection Line (IL) are passively opened. Vapor generated by residual heat pushes down the water level in the Reactor Vessel (RV). When the level is lower than the inlet of the PL, the vapor is ejected into the Containment Vessel (CV) through the PL. Then boronized water in the CV is injected into the RV through the IL by the static head. In an experiment using a simple apparatus, gas ejection and water injection were found to occur alternately under certain conditions. The gas ejection interval was observed to fluctuate considerably. Though stochastic noise affected the interval, the experimental results suggested that the large fluctuation was produced by an inherent character in the system. A set of piecewise linear differential equations was derived to describe the experimental result. The large fluctuation was reproduced in the analytical solution. Thus it was shown to occur even in a deterministic system without any source of stochastic noise. Though the derived equations simulated the experiment well, they had ten independent parameters governing the behavior of the solution. There appeared chaotic features and bifurcation, but the analytical model was too complicated to examine the features and mechanism of bifurcation. In this study, a new simple model is proposed which consists of a set of piecewise linear ordinary differential equations with only four independent parameters. (authors)

  3. CFD modeling and thermal-hydraulic analysis for the passive decay heat removal of a sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Hung, T.C.; Dhir, V.K.; Chang, J.C.; Wang, S.K.

    2011-01-01

    Research highlights: → The COOLOD/N2 and PARET/ANL codes were used for a steady-state thermal-hydraulic and safety analysis of the 2 MW TRIGA MARK II reactor located at the Nuclear Studies Center of Maamora (CENM), Morocco. → The main objective of this study is to ensure the safety margins of different safety related parameters by steady-state calculations at full power level (2 MW). → The most important conclusion is that all obtained values of DNBR, fuel center and surface temperature, cladding surface temperature and coolant temperature across the hottest channel are largely far to compromise safety of the reactor. - Abstract: In this study, a pool-typed design similar to sodium-cooled fast reactor (SFR) of the fourth generation reactors has been modeled using CFD simulations to investigate the characteristics of a passive mechanism of Shutdown Heat Removal System (SHRS). The main aim is to refine the reactor pool design in terms of temperature safety margin of the sodium pool. Thus, an appropriate protection mechanism is maintained in order to ensure the safety and integrity of the reactor system during a shutdown mode without using any active heat removal system. The impacts on the pool temperature are evaluated based on the following considerations: (1) the aspect ratio of pool diameter to depth, (2) the values of thermal emissivity of the surface materials of reactor and guard vessels, and (3) innerpool liner and core periphery structures. The computational results show that an optimal pool design in geometry can reduce the maximum pool temperature down to ∼551 o C which is substantially lower than ∼627 o C as calculated for the reference case. It is also concluded that the passive Reactor Air Cooling System (RACS) is effective in removing decay heat after shutdown. Furthermore, thermal radiation from the surface of the reactor vessel is found to be important; and thus, the selection of the vessel surface materials with a high emissivity would be a

  4. Technical feasibility and reliability of passive safety systems of AC600

    International Nuclear Information System (INIS)

    Niu, W.; Zeng, X.

    1996-01-01

    The first step conceptual design of the 600 MWe advanced PWR (AC-600) has been finished by the Nuclear Power Institute of China. Experiments on the passive system of AC-600 are being carried out, and are expected to be completed next year. The main research emphases of AC-600 conceptual design include the advanced core, the passive safety system and simplification. The design objective of AC-600 is that the safety, reliability, maintainability, operation cost and construction period are all improved upon compared to those of PWR plant. One of important means to achieve the objective is using a passive system, which has the following functions whenever its operation is required: providing the reactor core with enough coolant when others fail to make up the lost coolant; reactor residual heat removal; cooling and reducing pressure in the containment and preventing radioactive substances from being released into the environment after occurrence of accident (e.g. LOCA). The system should meet the single failure criterion, and keep operating when a single active component or passive component breaks down during the first 72 hour period after occurrence of accident, or in the long period following the 72 hour period. The passive safety system of AC-600 is composed of the primary safety injection system, the secondary emergency core residual heat removal system and the containment cooling system. The design of the system follows some relevant rules and criteria used by current PWR plant. The system has the ability to bear single failure, two complete separate subsystems are considered, each designed for 100% working capacity. Normal operation is separate from safety operation and avoids cross coupling and interference between systems, improves the reliability of components, and makes it easy to maintain, inspect and test the system. The paper discusses the technical feasibility and reliability of the passive safety system of AC-600, and some issues and test plans are also

  5. Technical feasibility and reliability of passive safety systems of AC600

    Energy Technology Data Exchange (ETDEWEB)

    Niu, W; Zeng, X [Nuclear Power Inst. of China, Chendu (China)

    1996-12-01

    The first step conceptual design of the 600 MWe advanced PWR (AC-600) has been finished. Experiments on the passive system of AC-600 are being carried out, and are expected to be completed next year. The main research emphases of AC-600 conceptual design include the advanced core, the passive safety system and simplification. The design objective of AC-600 is that the safety, reliability, maintainability, operation cost and construction period are all improved upon compared to those of PWR plant. One of important means to achieve the objective is using a passive system, which has the following functions whenever its operation is required: providing the reactor core with enough coolant when others fail to make up the lost coolant; reactor residual heat removal; cooling and reducing pressure in the containment and preventing radioactive substances from being released into the environment after occurrence of accident (e.g. LOCA). The system should meet the single failure criterion, and keep operating when a single active component or passive component breaks down during the first 72 hour period after occurrence of accident, or in the long period following the 72 hour period. The passive safety system of AC-600 is composed of the primary safety injection system, the secondary emergency core residual heat removal system and the containment cooling system. The design of the system follows some relevant rules and criteria used by current PWR plant. The system has the ability to bear single failure, two complete separate subsystems are considered, each designed for 100% working capacity. Normal operation is separate from safety operation and avoids cross coupling and interference between systems, improves the reliability of components, and makes it easy to maintain, inspect and test the system. The paper discusses the technical feasibility and reliability of the passive safety system of AC-600, and some issues and test plans are also involved. (author). 3 figs, 1 tab.

  6. Simulation of small break loss of coolant accident in pressurized water reactor (PWR)

    International Nuclear Information System (INIS)

    Abass, N. M. N.

    2012-02-01

    A major safety concern in pressurized-water-reactor (PWR) design is the loss-of-coolant accident (LOCA),in which a break in the primary coolant circuit leads to depressurization, boiling of the coolant, consequent reduced cooling of the reactor core, and , unless remedial measures are taken, overheating of the fuel rods. This concern has led to the development of several simulators for safety analysis. This study demonstrates how the passive and active safety systems in conventional and advanced PWR behave during the small break loss of Coolant Accident (SBLOCA). The consequences of SBOLOCA have been simulated using IAEA Generic pressurized Water Reactor Simulator (GPWRS) and personal Computer Transient analyzer (PCTRAN) . The results were presented and discussed. The study has confirmed the major safety advantage of passive plants versus conventional PWRs is that the passive safety systems provide long-term core cooling and decay heat removal without the need for operator actions and without reliance on active safety-related system. (Author)

  7. Effects of ship casualties on reactor safety and marine reactor design

    International Nuclear Information System (INIS)

    Agena, H.H.

    1978-01-01

    Ship casualties such as collision, grounding, fire and explosions, leakage and partial flooding may severely impair the safety of the nuclear reactor plant and result in nuclear hazards to the crew and the environment. Engineered safeguards are being discussed for protection against such consequences: manoeuvrability, structural (passive) collision and fire protection, protection against external fires and pressure waves, after heat transmission in case of a casualty, and after heat transmission out off the sunk wreck. Existing requirements will be discussed, shortly

  8. The characteristics of the prestressed concrete reactor vessel of the HHT demonstration plant

    International Nuclear Information System (INIS)

    Schoening, J.; Schwiers, H.G.

    1979-01-01

    The paper concentrates on the design studies of the HTGR prestressed concrete reactor vessel (PCRV) for the HHT Demonstration Plant. The multi-cavity reactor pressure vessel accommodates all components carrying primary gas, including heat exchangers and gas turbine. For reasons of economics and availability of the reactor plant, generic requirements are made for the PCRV. A short description of the power plant is also presented

  9. Gas Reactor International Cooperative program. Pebble bed reactor plant: screening evaluation. Volume 2. Conceptual balance of plant design

    Energy Technology Data Exchange (ETDEWEB)

    1979-11-01

    This report consists of three volumes which describe the design concepts and screening evaluation for a 3000 MW(t) Pebble Bed Reactor Multiplex Plant (PBR-MX). The Multiplex plant produces both electricity and transportable chemical energy via the thermochemical pipeline (TCP). The evaluation was limited to a direct cycle plant which has the steam generators and steam reformers in the primary circuit. This volume describes the conceptual balance-of-plant (BOP) design and was prepared by United Engineers and Constructors, Inc. of Philadelphia, Pennsylvania. The major emphasis of the BOP study was a preliminary design of an overall plant to provide a basis for future studies.

  10. Gas Reactor International Cooperative program. Pebble bed reactor plant: screening evaluation. Volume 2. Conceptual balance of plant design

    International Nuclear Information System (INIS)

    1979-11-01

    This report consists of three volumes which describe the design concepts and screening evaluation for a 3000 MW(t) Pebble Bed Reactor Multiplex Plant (PBR-MX). The Multiplex plant produces both electricity and transportable chemical energy via the thermochemical pipeline (TCP). The evaluation was limited to a direct cycle plant which has the steam generators and steam reformers in the primary circuit. This volume describes the conceptual balance-of-plant (BOP) design and was prepared by United Engineers and Constructors, Inc. of Philadelphia, Pennsylvania. The major emphasis of the BOP study was a preliminary design of an overall plant to provide a basis for future studies

  11. Passive decay heat removal by natural air convection after severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Erbacher, F.J.; Neitzel, H.J. [Forschungszentrum Karlsruhe Institut fur Angewandte Thermo- und Fluiddynamik, Karlsruhe (Germany); Cheng, X. [Technische Universitaet Karlsruhe Institut fur Stroemungslehre und Stroemungsmaschinen, Karlsruhe (Germany)

    1995-09-01

    The composite containment proposed by the Research Center Karlsruhe and the Technical University Karlsruhe is to cope with severe accidents. It pursues the goal to restrict the consequences of core meltdown accidents to the reactor plant. One essential of this new containment concept is its potential to remove the decay heat by natural air convection and thermal radiation in a passive way. To investigate the coolability of such a passive cooling system and the physical phenomena involved, experimental investigations are carried out at the PASCO test facility. Additionally, numerical calculations are performed by using different codes. A satisfying agreement between experimental data and numerical results is obtained.

  12. Knowledge and abilities catalog for nuclear power plant operators: boiling water reactors

    International Nuclear Information System (INIS)

    1986-09-01

    The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWR) (NUREG-1123) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog and Examiners' Handbook for Developing Operator Licensing Examinations (NUREG-1121) will cover those topics listed under Title 10, Code of Federal Regulations, Part 55. The BWR Catalog contains approximately 7000 knowledge and ability (K/A) statements for ROs and SROs at boiling water reactors. Each K/A statement has been rated for its importance to the safe operation of the plant in a manner ensuring personnel and public health and safety. The BWR K/A Catalog is organized into five major sections: Plant-wide Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Function, Emergency and Abnormal Plant Evolutions, Components, and Theory. The BWR Catalog represents a modification of the form and content of the K/A Catalog for Nuclear Power Plant Operators: Pressurized Water Reactors (NUREG-1122). First, categories of knowledge and ability statements have been redefined. Second, the scope of the definition of emergency and abnormal plant evolutions has been revised in line with a symptom-based approach. Third, K/As related to the operational applications of theory have been incorporated into the delineations for both plant systems and emergency and abnormal plant evolutions, while K/As pertaining to theory fundamental to plant operation have been delineated in a separate theory section. Finally, the components section has been revised

  13. Ensuring the operational safety of nuclear power plants with WWER reactors

    International Nuclear Information System (INIS)

    Shasharin, G.A.; Veretennikov, G.A.; Abagyan, A.A.; Lesnoj, S.A.

    1984-01-01

    At the start of 1983, 27 nuclear power producing units with reactor facilities of the WWER type were in operation in the Soviet Union and other countries. In 1982 the average load factor for nuclear power plants with WWER reactors was 73 per cent. There was not a single nuclear accident or even damage with any significant radiation consequences in the WWER reactors during the entire period of their operation. The most modern nuclear power plants with WWER-440 and WWER-1000 reactors meet all present-day international requirements. Safe operation of the plants is achieved by a variety of measures, the most important of which include: procedures for increasing the reliability of plant equipment and systems; ensuring exact compliance with plant operating instructions; ensuring reliable operation of plant safety systems; action directed towards maintaining the skills of plant personnel at a level adequate to ensure the taking of proper action during transient processes and accident situations. The paper discusses concrete steps for ensuring safe nuclear power plant operation along these lines. In particular, measures such as the following are described: the use of a system for collecting and processing information on equipment failures and defects; the development and introduction of methods of early defect diagnosis; the performance of complex testing of safety systems; the training of highly skilled personnel for nuclear power plants at educational combines and at teaching and training centres making use of simulators; arranging accident-prevention training and special instruction for personnel. (author)

  14. The development of reactor vessel internal heavy forging for 1000 MW pressurized-water reactor nuclear power plant

    International Nuclear Information System (INIS)

    Zhang Zhifeng; Chen Yongbo; Ding Xiuping; Zhang Lingfang

    2012-01-01

    This Paper introduced the development of Reactor Vessel Internal (RVI) heavy forgings for 1000 MW Pressurized Water Reactor (PWR) nuclear power plant, analyzed the manufacture difficulties and technical countermeasures. The testing result of the product indicated that the performance of RVI heavy forgings manufactured by Shanghai Heavy Machinery Plant Ld. (SHMP) is outstanding and entirely satisfy the technical requirements for RVI product. (authors)

  15. Passive Decay Heat Removal Strategy of Integrated Passive Safety System (IPSS) for SBO-combined Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sang Ho; Chang, Soon Heung; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-10-15

    The weak points of nuclear safety would be in outmoded nuclear power plants like the Fukushima reactors. One of the systems for the safety enhancement is integrated passive safety system (IPSS) proposed after the Fukushima accidents. It has the five functions for the prevention and mitigation of a severe accident. Passive decay heat removal (PDHR) strategy using IPSS is proposed for coping with SBO-combined accidents in this paper. The two systems for removing decay heat before core-melt were applied in the strategy. The accidents were simulated by MARS code. The reference reactor was OPR1000, specifically Ulchin-3 and 4. The accidents included loss-of-coolant accidents (LOCA) because the coolant losses could be occurred in the SBO condition. The examples were the stuck open of PSV, the abnormal open of SDV and the leakage of RCP seal water. Also, as LOCAs with the failure of active safety injection systems were considered, various LOCAs were simulated in SBO. Based on the thermal hydraulic analysis, the probabilistic safety analysis was carried out for the PDHR strategy to estimate the safety enhancement in terms of the variation of core damage frequency. AIMS-PSA developed by KAERI was used for calculating CDF of the plant. The IPSS was applied in the PDHR strategy which was developed in order to cope with the SBO-combined accidents. The estimation for initiating SGGI or PSIS was based on the pressure in RCS. The simulations for accidents showed that the decay heat could be removed for the safety duration time in SBO. The increase of safety duration time from the strategy provides the increase of time for the restoration of AC power.

  16. Review of advanced reactor transient analysis capabilities and applications for Savannah River Plant reactors

    International Nuclear Information System (INIS)

    Buckner, M.R.; Hostetler, D.E.; Anderson, M.M.; Dodds, H.L.

    1977-01-01

    GRASS is a three-dimensional, coupled neutronic and engineering code for analysis of the radioisotope production reactors at the Savannah River Plant. The capabilities of GRASS are reviewed with emphasis on recent additions to model accident conditions involving the transport of molten fuel material and to accurately characterize neutronic and engineering feedback. The general application of GRASS to the Savannah River reactors is discussed, and results are presented for the analyses of severla reactor transient calculations

  17. The investigation of Passive Accident Mitigation Scheme for advanced PWR NPP

    International Nuclear Information System (INIS)

    Shi, Er-bing; Fang, Cheng-yue; Wang, Chang; Xia, Geng-lei; Zhao, Cui-na

    2015-01-01

    Highlights: • We put forward a new PAMS and analyze its operation characteristics under SBO. • We conduct comparative analysis between PAMS and Traditional Secondary Side PHRS. • The PAMS could cope with SBO accident and maintain the plant in safe conditions. • PAMS could decrease heat removal capacity of PHRS. • PAMS has advantage in reducing cooling rate and PCCT temperature rising amplitude. - Abstract: To enhance inherent safety features of nuclear power plant, the advanced pressurized water reactors implement a series of passive safety systems. This paper puts forward and designs a new Passive Accident Mitigation Scheme (PAMS) to remove residual heat, which consists of two parts: the first part is Passive Auxiliary Feedwater System (PAFS), and the other part is Passive Heat Removal System (PHRS). This paper takes the Westinghouse-designed Advanced Passive PWR (AP1000) as research object and analyzes the operation characteristics of PAMS to cope with the Station Blackout Accident (SBO) by using RELAP5 code. Moreover, the comparative analysis is also conducted between PAMS and Traditional Secondary Circuit PHRS to derive the advantages of PAMS. The results show that the designed scheme can remove core residual heat significantly and maintain the plant in safe conditions; the first part of PAMS would stop after 120 min and the second part has to come into use simultaneously; the low pressurizer (PZR) pressure signal would be generated 109 min later caused by coolant volume shrinkage, which would actuate the Passive Safety Injection System (PSIS) to recovery the water level of pressurizer; the flow instability phenomenon would occur and last 21 min after the PHRS start-up; according to the comparative analysis, the coolant average temperature gradient and the Passive Condensate Cooling Tank (PCCT) water temperature rising amplitude of PAMS are lower than those of Traditional Secondary Circuit PHRS

  18. Development of Next-Generation LWR (Light Water Reactor) in Japan

    International Nuclear Information System (INIS)

    Yamamoto, T.; Kasai, S.

    2011-01-01

    The Next-Generation Light Water Reactor development program was launched in Japan in April 2008. The primary objective of the program is to cope with the need to replace existing nuclear power plants in Japan after 2030. The reactors to be developed are also expected to be a global standard design. Several innovative features are envisioned, including a reactor core system with uranium enrichment above 5%, a seismic isolation system, the use of long-life materials and innovative water chemistry, innovative construction techniques, safety systems with the best mix of passive and active concepts, and innovative digital technologies to further enhance reactor safety, reliability, economics, etc. In the first 3 years, a plant design concept with these innovative features is established and the effectiveness of the program is reevaluated. The major part of the program will be completed in 2015. (author)

  19. Active acoustic leak detection in steam generator units of fast reactors

    International Nuclear Information System (INIS)

    Oriol, L.; Journeau, Ch.

    1996-01-01

    Steam generators (SG) of Fast Reactors can be subject to water leakage into the sodium secondary circuit, causing an exothermic chemical reaction with potential serious damage to plant. Within the framework of the European Fast Reactor project, the CEA has developed an active acoustic detection technique which, when used in parallel with passive acoustic detection, will lead to effective leak detection results in terms of reliability and false alarm rates. Whilst the passive method is based on the increase in acoustic noise generated by the reaction, the active method takes advantage of the acoustic attenuation by the hydrogen bubbles produced. The method has been validated: in water, during laboratory testing at the Centre d'Etudes de Cadarache; in sodium, at the ASB loop at Bensberg (Germany) and at AEA Dounreay (Scotland). Full analysis of the tests carried out on the SG of the Prototype Fast Reactor in 1994 during end-of-life testing should lead to reactor validation on the method. (authors)

  20. Remote repairs and refurbishment of reactor internal structures of magnox plant

    International Nuclear Information System (INIS)

    Barnes, S.A.; Kelly, D.E.

    1992-01-01

    The original designers of the UK Magnox reactor plant made provision for the then perceived time dependent processes that could have influenced the operational life of the plant. Changes in graphite properties with irradiation, particularly dimensional change, were well understood and in-core samples were provided for subsequent laboratory examination to monitor the processes throughout plant life. The tendency towards embrittlement with irradiation of the steel of the reactor pressure vessels was also acknowledged and again in-core samples were provided for monitoring changes in materials properties in-service and thus provide data in support of structural analyses to sustain the reactor safety cases. (author)

  1. Westinghouse AP600 advanced nuclear plant design

    International Nuclear Information System (INIS)

    Gangloff, W.

    1999-01-01

    As part of the cooperative US Department of Energy (DOE) Advanced Light Water Reactor (ALWR) Program and the Electric Power Research Institute (EPRI), the Westinghouse AP600 team has developed a simplified, safe, and economic 600-megawatt plant to enter into a new era of nuclear power generation. Designed to satisfy the standards set by DOE and defined in the ALWR Utility Requirements Document (URD), the Westinghouse AP600 is an elegant combination of innovative safety systems that rely on dependable natural forces and proven technologies. The Westinghouse AP600 design simplifies plant systems and significant operation, inspections, maintenance, and quality assurance requirements by greatly reducing the amount of valves, pumps, piping, HVAC ducting, and other complex components. The AP600 safety systems are predominantly passive, depending on the reliable natural forces of gravity, circulation, convection, evaporation, and condensation, instead of AC power supplies and motor-driven components. The AP600 provides a high degree of public safety and licensing certainty. It draws upon 40 years of experience in light water reactor components and technology, so no demonstration plant is required. During the AP600 design program, a comprehensive test program was carried out to verify plant components, passive safety systems components, and containment behavior. When the test program was completed at the end of 1994, the AP600 became the most thoroughly tested advanced reactor design ever reviewed by the US Nuclear Regulatory Commission (NRC). The test results confirmed the exceptional behavior of the passive systems and have been instrumental in facilitating code validations. Westinghouse received Final Design Approval from the NRC in September 1998. (author)

  2. An overview of future sustainable nuclear power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Poullikkas, Andreas [Electricity Authority of Cyprus, P.O. Box 24506, 1399 Nicosia (Cyprus)

    2013-07-01

    In this paper an overview of the current and future nuclear power reactor technologies is carried out. In particular, the nuclear technology is described and the classification of the current and future nuclear reactors according to their generation is provided. The analysis has shown that generation II reactors currently in operation all around the world lack significantly in safety precautions and are prone to loss of coolant accident (LOCA). In contrast, generation III reactors, which are an evolution of generation II reactors, incorporate passive or inherent safety features that require no active controls or operational intervention to avoid accidents in the event of malfunction, and may rely on gravity, natural convection or resistance to high temperatures. Today, partly due to the high capital cost of large power reactors generating electricity and partly due to the consideration of public perception, there is a shift towards the development of smaller units. These may be built independently or as modules in a larger complex, with capacity added incrementally as required. Small reactors most importantly benefit from reduced capital costs, simpler units and the ability to produce power away from main grid systems. These factors combined with the ability of a nuclear power plant to use process heat for co-generation, make the small reactors an attractive option. Generally, modern small reactors for power generation are expected to have greater simplicity of design, economy of mass production and reduced installation costs. Many are also designed for a high level of passive or inherent safety in the event of malfunction. Generation III+ designs are generally extensions of the generation III concept, which include advanced passive safety features. These designs can maintain the safe state without the use of any active control components. Generation IV reactors, which are future designs that are currently under research and development, will tend to have closed

  3. Status of the IAEA coordinated research project on natural circulation phenomena, modelling, and reliability of passive systems that utilize natural circulation

    International Nuclear Information System (INIS)

    Reyes, J.N. Jr.; Cleveland, J.; Aksan, N.

    2004-01-01

    The International Atomic Energy Agency (IAEA) has established a Coordinated Research Project (CRP) titled ''Natural Circulation Phenomena, Modelling and Reliability of Passive Safety Systems that Utilize Natural Circulation. '' This work has been organized within the framework of the IAEA Department of Nuclear Energy's Technical Working Groups for Advanced Technologies for Light Water Reactors and Heavy Water Reactors (the TWG-LWR and the TWG-HWR). This CRP is part of IAEA's effort to foster international collaborations that strive to improve the economic performance of future water-cooled nuclear power plants while meeting stringent safety requirements. Thus far, IAEA has established 12 research agreements with organizations from industrialized Member States and 3 research contracts with organizations from developing Member States. The objective of the CRP is to enhance our understanding of natural circulation phenomena in water-cooled reactors and passive safety systems. The CRP participants are particularly interested in establishing a natural circulation and passive safety system thermal hydraulic database that can be used to benchmark computer codes for advanced reactor systems design and safety analysis. An important aspect of this CRP relates to developing methodologies to assess the reliability of passive safety systems in advanced reactor designs. This paper describes the motivation and objectives of the CRP, the research plan, and the role of each of the participating organizations. (author)

  4. The passive system for reflooding of the VVER reactor core from the second-stage hydro-accumulators: design and basic design solutions

    International Nuclear Information System (INIS)

    Alexandr D Efanov; Sergey G Kalyakin; Andrey V Morozov; Oleg V Remizov; Vladimir M Berkovich; Victor N Krushelnitskiy; Vladimir G Peresadko; Yuri G Dragunov; Alexey K Podshibyakin; Sergey I Zaitcev

    2005-01-01

    Full text of publication follows: The fundamental difference in the safety assurance of the operating NPPs and those under design implies that the safety in the existing NPPs is achieved by energy-dependent (active) systems and depends on the proficiency of attending personnel. To provide safety, the new NPP designs use the physical processes proceeding in the facility without power supply; and they are unaffected by human errors. As to the safety level, the design of the new generation nuclear power plant NPP-92 relates to the class of the improved NPPs; and it applies a principle of diversity in the structure of systems responsible for critical safety functions. In accordance with the above-mentioned safety concept, the design development required a complex of experimental investigations and numerical modeling to be conducted. Among the passive safety systems of the NPP with RP-392 is the system of the second stage hydro-accumulators (GE-2). The system of the second-stage hydro-accumulators consists of four groups of hydro-accumulating tanks with a total coolant volume of 960 m 3 . The system is intended for the core flooding with coolant during 24 hours. In each group of the hydro-accumulators, the graded coolant flowrate is provided, which depends on residual heat in the reactor. The special check valves are tuned to open at the pressure drop in the circuit below 1.5 MPa. The paper presents the thermalhydraulic substantiation of the serviceability of the second-stage hydro-accumulators system for passive heat removal from the VVER reactor core and the basic design solutions on the GE-2 system. (authors)

  5. ESBWR related passive decay heat removal tests in PANDA

    International Nuclear Information System (INIS)

    Huggenberger, M.; Aubert, C.; Bandurski, T.; Dreier, J.; Fischer, O.; Strassberger, H.J.; Yadigaroglu, G.

    1999-01-01

    A number of test series to investigate passive safety systems for the next generation of Light Water Reactors have been performed in the PANDA multi-purpose facility at the Paul Scherrer Institut (PSI). The large scale thermal-hydraulic test facility allows to investigate LWR containment phenomena and system behaviour. PANDA was first used to examine the Passive Containment Cooling System (PCCS) for the Simplified Boiling Water Reactor (SBWR). In 1996 new test series were initiated; all related to projects of the EC Fourth Framework Programme on Nuclear Fission Safety. One of these projects (TEPSS) is focused on the European Simplified Boiling Water Reactor (ESBWR). The ESBWR containment features and PCCS long-term post LOCA response were investigated in PANDA. The PCCS start-up was demonstrated, the effect of nitrogen hidden somewhere in the drywell and released later in the transient was simulated and the effect of light gases (helium) on the PCCS performance was investigated. Finally, the influence of low PCC pool levels on PCCS and containment performance was examined. The main findings were that the PCCS works as intended and shows generally a favorable and robust long-term post LOCA behaviour. The system starts working even under extreme conditions and trapped air released from the drywell later in the transient does only temporarily reduce the PCCS performance. The new PANDA test series provided an extensive data base which will contribute to further improve containment design of passive plants and allow for system code assessment in a wide parameter range. (author)

  6. Passive cooling of control rod drive mechanisms

    International Nuclear Information System (INIS)

    Hankinson, M.F.; Schwirian, R.E.

    1992-01-01

    A method and apparatus are provided for passively cooling the control rod drive mechanisms (CRDMs) in the reactor vessel of a nuclear power plant. Passive cooling is achieved by dispersing a plurality of chimneys within the CRDM array in positions where a control rod is not required. The chimneys induce convective air currents which cause ambient air from within the containment to flow over the CRDM coils. The air heated by the coils is guided into inlets in the chimneys by baffles. The chimney is insulated and extends through the seismic support platform and missile shield disposed above the closure head. A collar of adjustable height mates with plate elements formed at the distal end of the CRDM pressure housings by an interlocking arrangement so that the seismic support platform provides lateral restraint for the chimneys. (Author)

  7. Feasibility of passive heat removal systems

    Energy Technology Data Exchange (ETDEWEB)

    Ashurko, Yu M [Institute of Physics and Power Engineering, Obninsk (Russian Federation)

    1996-12-01

    This paper presents a review of decay heat removal systems (DHRSs) used in liquid metal-cooled fast reactors (LMFRs). Advantages and the disadvantages of these DHRSs, extent of their passivity and prospects for their use in advanced fast reactor projects are analyzed. Methods of extending the limitations on the employment of individual systems, allowing enhancement in their effectiveness as safety systems and assuring their total passivity are described. (author). 10 refs, 10 figs.

  8. Advanced light water reactor program at ABB-Combustion Engineering Nuclear Power

    International Nuclear Information System (INIS)

    Cahn, H.

    1990-01-01

    To meet the needs of Electric Utilities ordering nuclear power plants in the 1990s, ABB-Combustion Engineering is developing two designs which will meet EPRI consensus requirements and new licensing issues. The System 80 Plus design is an evolutionary pressurized water reactor plant modelled after the successful System 80 design in operation in Palo Verde and under construction in Korea. System Plus is currently under review by the US Nuclear Regulatory Commission with final design approval expected in 1991 and design certification in 1992. The Safe Integral Reactor (SIR) plant is a smaller facility with passive safety features and modular construction intended for design certification in the late 1990s. (author)

  9. MARS, 600 MWth NUCLEAR POWER PLANT

    International Nuclear Information System (INIS)

    Cumo, M.; Naviglio, A.; Sorabella, L.

    2004-01-01

    MARS (Multipurpose Advanced Reactor, inherently Safe) is a 600 MWth, single loop, pressurized light water reactor (PWR), developed at the Dept. of Nuclear Engineering and Energy Conversion of the University of Rome ''La Sapienza''. The design was focused to a multipurpose reactor to be used in high population density areas also for industrial heat production and, in particular, for water desalting. Using the well-proven technology and the operation experience of PWRs, the project introduces a lot of innovative features hugely improving the safety performance while keeping the cost of KWh competitive with traditional large power plants. Extensive use of passive safety, in depth plant simplification and decommissioning oriented design were the guidelines along the design development. The latest development in the plant design, in the decommissioning aspects and in the experimental activities supporting the project are shown in this paper

  10. Proposal for a advanced PWR core with adequate characteristics for passive safety concept

    International Nuclear Information System (INIS)

    Perrotta, Jose Augusto

    1999-01-01

    This work presents a discussion upon the suitable from an advanced PWR core, classified by the EPRI as 'Passive PWR' (advanced reactor with passive safety concept to power plants with less than 600 MW electrical power). The discussion upon the type of core is based on nuclear fuel engineering concepts. Discussion is made on type of fuel materials, structural materials, geometric shapes and manufacturing process that are suitable to produce fuel assemblies which give good performance for this type of reactors. The analysis is guided by the EPRI requirements for Advanced Light Water Reactor (ALWR). By means of comparison, the analysis were done to Angra 1 (old type of 600 MWe PWR class), and the design of the Westinghouse Advanced PWR-AP600. It was verified as a conclusion of this work that the modern PWR fuels are suitable for advanced PWR's Nevertheless, this work presents a technical alternative to this kind of fuel, still using UO 2 as fuel, but changing its cylindrical form of pellets and pin type fuel element to plane shape pallets and plate type fuel element. This is not a novelty fuel, since it was used in the 50's at Shippingport Reactor and as an advanced version by CEA of France in the 70's. In this work it is proposed a new mechanical assembly design for this fuel, which can give adequate safety and operational performance to the core of a 'Passive PWR'. (author)

  11. Safety problems of nuclear power plants with reactors of new generation; Voprosy bezopasnosti v proehktakh AEhS novogo pokoleniya s WWER

    Energy Technology Data Exchange (ETDEWEB)

    Fedorov, V; Rogov, M; Biryukov, G [Opytno-Konstruktorskoe Byuro Gidropress, Podol` sk (Russian Federation)

    1996-12-31

    Modernization schemes for safety enhancement of WWER-1000 reactors are proposed. In the case of WWER-1000/V-392 it is based on introduction of additional safety systems and overall design improvement. For WWER-1100 (1000-1100 MW) the safety is enhanced by passive systems built from two-stage heat exchangers. For WWER-500/600 the use of passive safety system is extended to emergency cooling of the active zone and removal of the residual heat emissions from the reactor. The technical characteristics of the three reactors are compared. 3 figs., 1 tab.

  12. The United States fluoride-salt-cooled high-temperature reactor program

    International Nuclear Information System (INIS)

    Holcomb, David E.

    2013-01-01

    The United States is pursuing the development of fluoride-salt-cooled high-temperature reactors (FHRs) through the Department of Energy's Office of Nuclear Energy (DOE-NE). FHRs, by definition, feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. FHRs, in principle, have the potential to economically generate large amounts of electricity while maintaining full passive safety. FHRs, however, remain a longer-term power production option. A principal development focus is, thus, on shortening, to the extent possible, the overall development time by focusing initial efforts on the longest lead-time issues. While FHRs represent a distinct reactor class, they inherit desirable attributes from other thermal power plants whose characteristics can be studied to provide general guidance on plant configuration, anticipated performance, and costs. Molten salt reactors provide experience on the materials, procedures, and components necessary to use liquid fluoride salts. Liquid-metal reactors provide design experience on using low-pressure liquid coolants, passive decay heat removal, and hot refueling. High-temperature gas-cooled reactors provide experience with coated-particle fuel and graphite components. Light-water reactors show the potential of transparent, high-heat-capacity coolants with low chemical reactivity. The FHR development efforts include both reactor concept and technology developments and are being broadly pursued. Oak Ridge National Laboratory (ORNL) provides technical leadership to the effort and is performing concept development on both a large base-load-type FHR as well as a small modular reactor (SMR) in addition to performing a broad scope of technology developments. Idaho National Laboratory (INL) is providing coated-particle fuel irradiation testing as well as developing high-temperature steam generator technology. The Massachusetts Institute of Technology (MIT

  13. Nuclear Power Station Kalkar, 300 MWe Prototype Nuclear Power Plant with Fast Sodium Cooled Reactor (SNR-300), Plant description

    International Nuclear Information System (INIS)

    1984-06-01

    The nuclear power station Kalkar (SNR-300) is a prototype with a sodium cooled fast reactor and a thermal power of 762 MW. The present plant description has been made available in parallel to the licensing procedure for the reactor plant and its core Mark-Ia as supplementary information for the public. The report gives a detailed description of the whole plant including the prevention measures against the impact of external and plant internal events. The radioactive materials within the reactor cooling system and the irradiation protection and surveillance measures are outlined. Finally, the operation of the plant is described with the start-up procedures, power operation, shutdown phases with decay heat removal and handling procedures

  14. MODULAR AND FULL SIZE SIMPLIFIED BOILING WATER REACTOR DESIGN WITH FULLY PASSIVE SAFETY SYSTEMS

    International Nuclear Information System (INIS)

    Ishii, M.; Revankar, S. T.; Downar, T.; Xu, Y.; Yoon, H. J.; Tinkler, D.; Rohatgi, U. S.

    2003-01-01

    OAK B204 The overall goal of this three-year research project was to develop a new scientific design of a compact modular 200 MWe and a full size 1200 MWe simplified boiling water reactors (SBWR). Specific objectives of this research were: (1) to perform scientific designs of the core neutronics and core thermal-hydraulics for a small capacity and full size simplified boiling water reactor, (2) to develop a passive safety system design, (3) improve and validate safety analysis code, (4) demonstrate experimentally and analytically all design functions of the safety systems for the design basis accidents (DBA) and (5) to develop the final scientific design of both SBWR systems, 200 MWe (SBWR-200) and 1200 MWe (SBWR-1200). The SBWR combines the advantages of design simplicity and completely passive safety systems. These advantages fit well within the objectives of NERI and the Department of Energy's focus on the development of Generation III and IV nuclear power. The 3-year research program was structured around seven tasks. Task 1 was to perform the preliminary thermal-hydraulic design. Task 2 was to perform the core neutronic design analysis. Task 3 was to perform a detailed scaling study and obtain corresponding PUMA conditions from an integral test. Task 4 was to perform integral tests and code evaluation for the DBA. Task 5 was to perform a safety analysis for the DBA. Task 6 was to perform a BWR stability analysis. Task 7 was to perform a final scientific design of the compact modular SBWR-200 and the full size SBWR-1200. A no cost extension for the third year was requested and the request was granted and all the project tasks were completed by April 2003. The design activities in tasks 1, 2, and 3 were completed as planned. The existing thermal-hydraulic information, core physics, and fuel lattice information was collected on the existing design of the simplified boiling water reactor. The thermal-hydraulic design were developed. Based on a detailed integral

  15. The US Advanced Liquid Metal Reactor and the Fast Flux Test Facility Phase IIA passive safety tests

    International Nuclear Information System (INIS)

    Shen, P.K.; Harris, R.A.; Campbell, L.R.; Dautel, W.A.; Dubberley, A.E.; Gluekler, E.L.

    1992-07-01

    This report discusses the safety approach of the Advanced Liquid Metal reactor program, sponsored by the US Department of Energy, which relies upon passive reactor responses to off-normal condition to limit power and temperature excursions to levels that allow safety margins. Gas expansion modules (GEM) have included in the design to provide negative reactivity to enhance these margins in the extremely unlikely event that pumping power is lost and the highly reliable scram system fails to operate. The feasibility and beneficial features of these devices were first demonstrated in the core of the Fast Flux Test Facility (FFTF) in 1986. Preapplication safety evaluations by the US Nuclear Regulatory Commission have identified areas that must be addressed if these devices are to be relied on. One of these areas is the response of the reactor when it is critical and the pumps are turned on, resulting in positive reactivity being added to the core. Tests to examine such transients have been performed as part of the continuing FFTF program to confirm the passive safety characteristics of liquid metal reactors (LMR). The primary tests consisted of starting the main coolant pumps, which forced sodium coolant into the GEMS, decreasing neutron leakage and adding positive reactivity. The resulting transients were shown to be benign and easily mitigated by the reactivity feedbacks inherent in the FFTF and all LMRs. Steady-state auxiliary tests of the GEM and feedback reactivity worths accurately predicted the transient results. The auxiliary GEM worth tests also demonstrated that the worth can be determined at a subcritical state, which allows for a verification of the GEM's availability prior to ascending to power

  16. Safety analysis of an expert reactor protection system in nuclear power plants

    International Nuclear Information System (INIS)

    El-Kafas, A.A.

    1997-01-01

    The purpose of the dissertation is to develop real time expert reactor protection system (ERPS) for operational safety of pressurized water reactor nuclear power plant. The system is developed to diagnose plant failures and for identification plant transients (with and without scram). For this erps, probabilistic safety analysis techniques are used to check the availability and priority of the recommended safety system in case of plant accidents. The real - time information during transients and accidents can be obtained to assess the operator in his decision - making. Also, the ERPS is able to give advice for the reactor operator to take the appropriate corrective action during abnormal situations. 5-15 figs., 42 refs

  17. Dynamic simulation of a sodium-cooled fast reactor power plant

    Energy Technology Data Exchange (ETDEWEB)

    Shinaishin, M.A.M.

    1976-08-01

    Simulation of the dynamic behavior of the Clinch River Breeder Reactor Plant (CRBRP) is the subject of this dissertation. The range of transients under consideration extends from a moderate transient, of the type referred to as Anticipated Transient Without Scram (ATWS), to a transient initiated by an unexpected accident followed by reactor scram. The moderate range of transients can be simulated by a digital simulator referred to as the CRBRP ATWS simulator. Two versions of this simulator were prepared; in one, the plant controllers were not included, whereas, in the other, the controllers were incorporated. A simulator referred to as the CRBRP-DCHT simulator was constructed for studying transients due to unexpected accidents followed by reactor scram. In this simulator emphasis was placed on simulating the auxiliary heat removal system, in order to determine its capability to remove the after-shut down fission and decay heat. The transients studied using the two versions of the ATWS simulator include step and ramp reactivity perturbations, and an electrical load perturbation in the controlled plant. An uncontrolled control rod withdrawal followed by reactor scram was studied using the DCHT simulator, although the duration of this transient was restricted to 20 sec. because of computer limitations. The results agree very well with the expected physical behavior of the plant.

  18. Dynamic simulation of a sodium-cooled fast reactor power plant

    International Nuclear Information System (INIS)

    Shinaishin, M.A.M.

    1976-08-01

    Simulation of the dynamic behavior of the Clinch River Breeder Reactor Plant (CRBRP) is the subject of this dissertation. The range of transients under consideration extends from a moderate transient, of the type referred to as Anticipated Transient Without Scram (ATWS), to a transient initiated by an unexpected accident followed by reactor scram. The moderate range of transients can be simulated by a digital simulator referred to as the CRBRP ATWS simulator. Two versions of this simulator were prepared; in one, the plant controllers were not included, whereas, in the other, the controllers were incorporated. A simulator referred to as the CRBRP-DCHT simulator was constructed for studying transients due to unexpected accidents followed by reactor scram. In this simulator emphasis was placed on simulating the auxiliary heat removal system, in order to determine its capability to remove the after-shut down fission and decay heat. The transients studied using the two versions of the ATWS simulator include step and ramp reactivity perturbations, and an electrical load perturbation in the controlled plant. An uncontrolled control rod withdrawal followed by reactor scram was studied using the DCHT simulator, although the duration of this transient was restricted to 20 sec. because of computer limitations. The results agree very well with the expected physical behavior of the plant

  19. Analysis of reactor trips originating in balance of plant systems

    International Nuclear Information System (INIS)

    Stetson, F.T.; Gallagher, D.W.; Le, P.T.; Ebert, M.W.

    1990-09-01

    This report documents the results of an analysis of balance-of-plant (BOP) related reactor trips at commercial US nuclear power plants of a 5-year period, from January 1, 1984, through December 31, 1988. The study was performed for the Plant Systems Branch, Office of Nuclear Reactor Regulation, US Nuclear Regulatory Commission. The objectives of the study were: to improve the level of understanding of BOP-related challenges to safety systems by identifying and categorizing such events; to prepare a computerized data base of BOP-related reactor trip events and use the data base to identify trends and patterns in the population of these events; to investigate the risk implications of BOP events that challenge safety systems; and to provide recommendations on how to address BOP-related concerns in regulatory context. 18 refs., 2 figs., 27 tabs

  20. Safety analysis of increase in heat removal from reactor coolant system with inadvertent operation of passive residual heat removal at no load conditions

    Energy Technology Data Exchange (ETDEWEB)

    Shao, Ge; Cao, Xuewu [School of Mechanical and Engineering, Shanghai Jiao Tong University, Shanghai (China)

    2015-06-15

    The advanced passive pressurized water reactor (PWR) is being constructed in China and the passive residual heat removal (PRHR) system was designed to remove the decay heat. During accident scenarios with increase of heat removal from the primary coolant system, the actuation of the PRHR will enhance the cooldown of the primary coolant system. There is a risk of power excursion during the cooldown of the primary coolant system. Therefore, it is necessary to analyze the thermal hydraulic behavior of the reactor coolant system (RCS) at this condition. The advanced passive PWR model, including major components in the RCS, is built by SCDAP/RELAP5 code. The thermal hydraulic behavior of the core is studied for two typical accident sequences with PRHR actuation to investigate the core cooling capability with conservative assumptions, a main steam line break (MSLB) event and inadvertent opening of a steam generator (SG) safety valve event. The results show that the core is ultimately shut down by the boric acid solution delivered by Core Makeup Tank (CMT) injections. The effects of CMT boric acid concentration and the activation delay time on accident consequences are analyzed for MSLB, which shows that there is no consequential damage to the fuel or reactor coolant system in the selected conditions.

  1. Coupling of RO-MSF hybrid desalination plants with nuclear reactors

    International Nuclear Information System (INIS)

    Al-Sulaiman, Khalil; Al-Mutaz, Ibrahim S.

    1999-01-01

    Full text.Reverse osmosis (RO) and multistage flash (MSF) desalination are the most widely commercial available processes. MSF utilizes stream in the brine heater as a primary source of energy. RO is derived mainly by electricity that pumps the feed water against the mambranes. Steam and electricity and be produced easily by nuclear reactors. Nuclear reactors may be coupled with deslination plants (MSF, RO or combined (hybrid) RO/MSF configuration). This integrated plant will be capable of producing power and water at reasonable cost. The capital and operating cost will be reduced and the excess power can be efficiently utilized. Maintenance and operating cost will drop significantly. In this paper, a techno-economic study of hybrid reverses osmosis /multistage flash desalination will be carried. The proposed configuration (hybrid RO/MSF) coupled with nuclear reactor is considered the most appropriate candidate system for the application of dual-purpose nuclear desalination plants. the design parameters for such a desalination hybrid system will be the applied pressure and recovery for reverse osmosis plant and the number of stages and the heat transfer areas for multistage flash plant

  2. New nuclear plant design and licensing process

    International Nuclear Information System (INIS)

    Luangdilok, W.

    1996-01-01

    This paper describes latest developments in the nuclear power reactor technology with emphasis on three areas: (1) the US technology of advanced passive light water reactors (AP600 and S BWR), (2) regulatory processes that certify their safety, and (3) current engineering concerns. The goal is to provide and insight of how the government's regulatory agency guarantees public safety by looking into how new passive safety features were designed and tested by vendors and how they were re-evaluated and retested by the US NRC. The paper then discusses the US 1989 nuclear licensing reform (10 CFR Part 52) whose objectives are to promote the standardization of nuclear power plants and provide for the early and definitive resolution of site and design issues before plants are built. The new licensing process avoids the unpredictability nd escalated construction cost under the old licensing process. Finally, the paper summarizes engineering concerns found in current light water reactors that may not go away in the new design. The concerns are related the material and water chemistry technology in dealing with corrosion problems in water-cooled nuclear reactor systems (PWRs and BWRs). These engineering concerns include core shroud cracking (BWRs), jet pump hold-down beam cracking (BWRs), steam generator tube stress corrosion cracking (PWR)

  3. Power unit with GT-MHR reactor plant for electricity production and district heating

    International Nuclear Information System (INIS)

    Kiryushin, A.L.; Kodochigov, N.G.; Kuzavkov, N.G.; Golovko, V.F.

    2000-01-01

    Modular helium reactor with the gas turbine (GT-MHR) is a perspective power reactor plant for the next century. The project reactor is based on experience of operation more than 50 gas-cooled reactors on carbon dioxide and helium, and also on subsequent achievements in the field of realization direct gas turbine Brayton cycle. To the beginning of 90 years, achievements in technology of gas turbines, highly effective recuperators and magnetic bearings made it possible to start development of the reactor plant project combining a safe modular gas cooled reactor and a power conversion system, realizing the highly effective Brayton cycle. The conceptual project of the commercial GT-MHR reactor plant fulfilled in 1997 by joint efforts of international firms, combines a safe modular reactor with an annular active core of prismatic fuel blocks and a power conversion system with direct gas turbine cycle. The efficiency of GT-MHR gas turbine cycle at level of about 48% makes it competitive in the electricity production market in comparison with any fossil or nuclear power stations

  4. Direct-contact condensation regime map for core makeup tank of passive reactors

    International Nuclear Information System (INIS)

    Lee, Sang Il; No, Hee Cheon

    1998-01-01

    The condensation regime map in the core makeup tank of passive reactors is experimentally investigated. The condensation regimes identified through the experiments are divided into three distinct ones: sonic jet, subsonic jet, and steam cavity. The steam cavity regime is a unique regime of downward injection with the present geometry not previously observed in other experiments. The condensation regime map is constructed using Froude number and Jacob number. It turns out that the buoyancy force has a large influence on the regime transition because the regime map using the Froude number better fits data with different geometries than other dimensionless parameters. Simple correlations for the regime boundaries are proposed using the Froude number and the Jacob number

  5. Design of an Experimental Facility for Passive Heat Removal in Advanced Nuclear Reactors

    Science.gov (United States)

    Bersano, Andrea

    With reference to innovative heat exchangers to be used in passive safety system of Gen- eration IV nuclear reactors and Small Modular Reactors it is necessary to study the natural circulation and the efficiency of heat removal systems. Especially in safety systems, as the decay heat removal system of many reactors, it is increasing the use of passive components in order to improve their availability and reliability during possible accidental scenarios, reducing the need of human intervention. Many of these systems are based on natural circulation, so they require an intense analysis due to the possible instability of the related phenomena. The aim of this thesis work is to build a scaled facility which can reproduce, in a simplified way, the decay heat removal system (DHR2) of the lead-cooled fast reactor ALFRED and, in particular, the bayonet heat exchanger, which transfers heat from lead to water. Given the thermal power to be removed, the natural circulation flow rate and the pressure drops will be studied both experimentally and numerically using the code RELAP5 3D. The first phase of preliminary analysis and project includes: the calculations to design the heat source and heat sink, the choice of materials and components and CAD drawings of the facility. After that, the numerical study is performed using the thermal-hydraulic code RELAP5 3D in order to simulate the behavior of the system. The purpose is to run pretest simulations of the facility to optimize the dimensioning setting the operative parameters (temperature, pressure, etc.) and to chose the most adequate measurement devices. The model of the system is continually developed to better simulate the system studied. High attention is dedicated to the control logic of the system to obtain acceptable results. The initial experimental tests phase consists in cold zero power tests of the facility in order to characterize and to calibrate the pressure drops. In future works the experimental results will be

  6. Model-Based Analysis and Efficient Operation of a Glucose Isomerization Reactor Plant

    DEFF Research Database (Denmark)

    Papadakis, Emmanouil; Madsen, Ulrich; Pedersen, Sven

    2015-01-01

    efficiency. The objective of this study is the application of the developed framework on an industrial case study of a glucose isomerization (GI) reactor plant that is part of a corn refinery, with the objective to improve the productivity of the process. Therefore, a multi-scale reactor model...... is developedfor use as a building block for the GI reactor plant simulation. An optimal operation strategy is proposed on the basis of the simulation results...

  7. Studies on the behaviour of a passive containment cooling system for the Indian advanced heavy water reactor

    International Nuclear Information System (INIS)

    Maheshwari, N.K.; Saha, D.; Chandraker, D.K.; Kakodkar, A.; Venkat Raj, V.

    2001-01-01

    A passive containment cooling system has been proposed for the advanced heavy water reactor being designed in India. This is to provide long term cooling for the reactor containment following a loss of coolant accident. The system removes energy released into the containment through immersed condensers kept in a pool of water. An important aspect of immersed condenser's working is the potential degradation of immersed condenser's performance due to the presence of noncondensable gases. An experimental programme to investigate the passive containment cooling system behaviour and performance has been undertaken in a phased manner. In the first phase, system response tests were conducted on a small scale model to understand the phenomena involved. Tests were conducted with constant energy input rate and with varying energy input rate simulating decay heat. With constant energy input rate, pressures in volume V 1 and V 2 reached almost steady value. With varying energy input rate V 1 pressure dropped below the pressure in V 2 . The system could efficiently purge air from V 1 to V 2 . The paper deals with the details of the tests conducted and the results obtained. (orig.) [de

  8. Plant accident dynamics of high-temperature reactors with direct gas turbine cycle

    International Nuclear Information System (INIS)

    Waloch, M.L.

    1977-01-01

    In the paper submitted, a one-dimensional accident simulation model for high-temperature reactors with direct-cycle gas turbine (single-cycle facilities) is described. The paper assesses the sudden failure of a gas duct caused by the double-ended break of one out of several parallel pipes before and behind the reactor for a non-integrated plant, leading to major loads in the reactor region, as well as the complete loss of vanes of the compressor for an integrated plant. The results of the calculations show especially high loads for the break of a hot-gas pipe immediately behind the flow restrictors of the reactor outlet, because of prolonged effects of pressure gradients in the reactor region and the maximum core differential pressure. A plant accident dynamics calculation therefore allows to find a compromise between the requirements of stable compressor operation, on the one hand, and small loads in the reactor in the course of an accident, on the other, by establishing in a co-ordinated manner the narrowing ratio of the flow restrictors. (GL) [de

  9. The concept of the innovative power reactor

    Directory of Open Access Journals (Sweden)

    Sang Won Lee

    2017-10-01

    Full Text Available The Fukushima accident reveals the vulnerability of existing active nuclear power plant (NPP design against prolonged loss of external electricity events. The passive safety system is considered an attractive alternative to cope with this kind of disaster. Also, the passive safety system enhances both the safety and the economics of NPPs. The adoption of a passive safety system reduces the number of active components and can minimize the construction cost of NPPs. In this paper, reflecting on the experience during the development of the APR+ design in Korea, we propose the concept of an innovative Power Reactor (iPower, which is a kind of passive NPP, to enhance safety in a revolutionary manner. The ultimate goal of iPower is to confirm the feasibility of practically eliminating radioactive material release to the environment in all accident conditions. The representative safety grade passive system includes a passive emergency core cooling system, a passive containment cooling system, and a passive auxiliary feedwater system. Preliminary analysis results show that these concepts are feasible with respect to preventing and/or mitigating the consequences of design base accidents and severe accidents.

  10. Development of next-generation light water reactor

    International Nuclear Information System (INIS)

    Ishibashi, Fumihiko; Yasuoka, Makoto

    2010-01-01

    The Next-Generation Light Water Reactor Development Program, a national project in Japan, was inaugurated in April 2008. The primary objective of this program is to meet the need for the replacement of existing nuclear power plants in Japan after 2030. With the aim of setting a global standard design, the reactor to be developed offers greatly improved safety, reliability, and economic efficiency through several innovative technologies, including a reactor core system with uranium enrichment of 5 to 10%, a seismic isolation system, long-life materials, advanced water chemistry, innovative construction techniques, optimized passive and active safety systems, innovative digital technologies, and so on. In the first three years, a plant design concept with these innovative features is to be established and the effectiveness of the program will be reevaluated. The major part of the program will be completed in 2015. Toshiba is actively engaged in both design studies and technology development as a founding member of this program. (author)

  11. Scaling for Mixed Convection Heat Transfer in Passive Containments and Experiment Design

    International Nuclear Information System (INIS)

    Wang, Shengfei; Yu, Yu; Lv, Xuefeng; Niu, Fenglei; Yan, Xiuping

    2012-01-01

    Most of the advanced nuclear reactor design utilizes passive systems to remove heat from the core by natural circulation. The passive systems will be widely used in generation III pressurized water reactor. One of the typical passive systems is passive containment cooling system (PCCS), which is a passive condenser system designed to remove heat from the containment for long term cooling after a postulated reactor accident. In order to establish empirical correlations and develop simulation models, a scaling analysis is performed in designing an experiment for the prototype PCCS. This paper presents a scaling method and the design of the experimental facility. The key dimensionless parameters governing the dominant processes are given at last

  12. Boiling water reactor with innovative safety concept: The Generation III+ SWR-1000

    Energy Technology Data Exchange (ETDEWEB)

    Stosic, Zoran V. [AREVA NP GmbH, Koldestr. 16, 91052 Erlangen (Germany)], E-mail: Zoran.Stosic@areva.com; Brettschuh, Werner; Stoll, Uwe [AREVA NP GmbH, Koldestr. 16, 91052 Erlangen (Germany)

    2008-08-15

    AREVA NP has developed an innovative boiling water reactor (BWR) SWR-1000 in close cooperation with German nuclear utilities and with support from various European partners. This Generation III+ reactor design marks a new era in the successful tradition of BWR and, with a net electrical output of approximately 1250 MWe, is aimed at ensuring competitive power generating costs compared to gas and coal fired stations. It is particularly suitable for countries whose power networks cannot facilitate large power plants. At the same time, the SWR-1000 meets the highest safety standards, including control of core melt accidents. These objectives are met by supplementing active safety systems with passive safety equipment of various designs for accident detection and control and by simplifying systems needed for normal plant operation on the basis of past operating experience. The plant is also protected against airplane crash loads. The functional capabilities and capacities of all new systems and components were successfully tested under realistic and conservative boundary conditions in large-scale test facilities in Finland, Switzerland and Germany. In general, the SWR-1000 design is based on well-proven analytical codes and design tools validated for BWR applications through recalculation of relevant experiments and independent licensing activities performed by authorities or their experts. The overview of used analytical codes and design tools as well as performed experimental validation programs is presented. Effective implementation of passive safety systems is demonstrated through the numerical simulation of transients and loss of coolant accidents (LOCAs) as well as through analytical simulation of a severe accident associated with the core melt. In the LOCA simulation presented the existing active core flooding systems were not used for emergency control: only passive systems were relevant for the analyses. Despite this - no core heat-up occurred. In the case of

  13. Passive cooling in modern nuclear reactors

    International Nuclear Information System (INIS)

    Rouai, N. M.

    1998-01-01

    This paper presents some recent experimental results performed with the aim of understanding the mechanism of passive cooling. The AP 600 passive containment cooling system is simulated by an electrically heated vertical pipe, which is cooled by a naturally induced air flow and by a water film descending under gravity. The results demonstrate that although the presence of the water film improved the heat transfer significantly, the mode of heat transfer was very dependent on the experimental parameters. Preheating the water improved both film stability and overall cooling performance

  14. Evaluation on driving force of natural circulation in downcomer for passive residual heat removal system in JAERI passive safety reactor JPSR

    International Nuclear Information System (INIS)

    Kunii, Katsuhiko; Iwamura, Takamichi; Murao, Yoshio

    1997-01-01

    The driving-force of the natural circulation in the residual heat removal (RHR) system for the JPSR (JAERI Passive Safety Reactor) is given as a gravity force of the density difference between hotter coolant in core and upper plenum and cooler coolant in downcomer. The amount of density difference and time to achieve the enough density difference for the RHR system change directly dependent on the thermal fluid flow pattern in downcomer of annulus flow pass. The purposes of the present study are to investigate the possibilities of the followings by evaluating the three-dimensional thermal fluid flow in downcomer by numerical analysis using the STREAM code; 1) promotion of making the flow pattern uniform in downcomer by installing a baffle, 2) achievement of an enough driving-force of the natural circulation, 3) validity of one-point assumption, that is, complete mixing down-flow assumption for the three-dimensional thermal fluid flow in downcomer to evaluate the function of the passive RHR system. The following conclusions were obtained: (1) The effect of baffle on the thermal fluid flow and driving-force is little, (2) The driving-force required for natural circulation cooling can be obtained in wide range of inlet velocity even if the flow is multi-dimensional, (3) Both in initial transient stage and in steady-state, the one-point assumption can be applied to evaluate the driving-force of natural circulation in the passive RHR system. (author)

  15. National nuclear power planning of China and advanced reactor

    International Nuclear Information System (INIS)

    Qian Jihui

    1990-01-01

    The necessity of investigation on the trends of advanced reactor technology all over the world is elabrated while China is going to set up its long-term national nuclear power programme. In author's opinion, thermal reactor power plants will have a quite long period development in the next century and a new trend of second generation NPPs might emerge in the beginning of next century. These new generation advanced reactors are characterized with new design concepts based on the inherent or passive safety features. Among them, most promising ones are those of AP-600 and MHTGR. Chinese experts are paying special attention to and closely following these two directions

  16. Gen-III/III+ reactors. Solving the future energy supply shortfall. The SWR-1000 option

    International Nuclear Information System (INIS)

    Stosic, Z.V.

    2006-01-01

    Deficiency of non-renewable energy sources, growing demand for electricity and primary energy, increase in population, raised concentration of greenhouse gases in the atmosphere and global warming are the facts which make nuclear energy currently the most realistic option to replace fossil fuels and satisfy global demand. The nuclear power industry has been developing and improving reactor technology for almost five decades and is now ready for the next generation of reactors which should solve the future energy supply shortfall. The advanced Gen-III/III+ (Generation III and/or III+) reactor designs incorporate passive or inherent safety features which require no active controls or operational intervention to manage accidents in the event of system malfunction. The passive safety equipment functions according to basic laws of physics such as gravity and natural convection and is automatically initiated. By combining these passive systems with proven active safety systems, the advanced reactors can be considered to be amongst the safest equipment ever made. Since the beginning of the 90's AREVA NP has been intensively engaged in the design of two advanced Gen-III+ reactors: (i) PWR (Pressurized Water Reactor) EPR (Evolutionary Power Reactor) and (ii) BWR (Boiling Water Reactor) SWR-1000. The SWR-1000 reactor design marks a new era in the successful tradition of BWR technology. It meets the highest safety standards, including control of a core melt accident. This is achieved by supplementing active safety systems with passive safety equipment of diverse design for accident detection and control and by simplifying systems needed for normal plant operation. A short construction period, flexible fuel cycle lengths and a high fuel discharge burn-up contribute towards meeting economic goals. The SWR-1000 completely fulfils international nuclear regulatory requirements. (author)

  17. Nuclear power plant V-2

    International Nuclear Information System (INIS)

    1999-01-01

    In this leaflet the short history of commissioning of Bohunice V-2 NPP is reviewed (beginning of construction December 1976; First controlled reactor power, Reactor Unit 1 (RU1): 7 August 1984, Reactor Unit 2 (RU2): 2 August 1985; Connection to the grid: RU1 20 August 1984, RU2 9 August 1985; Commercial operation: RU1 14 February 1985, RU2 18 December 1985. The scheme of the nuclear reactor WWER 440/V213 is depicted. The major technological equipment are described. Principles of nuclear power plant operation safety (safety barriers, active and passive safety systems, centralized heat supply system, as well as technical data of the Bohunice V-2 NPP are presented

  18. Tritium instrumentation for a fusion reactor power plant

    International Nuclear Information System (INIS)

    Shank, K.E.; Easterly, C.E.

    1976-09-01

    A review of tritium instrumentation is presented. This includes a discussion of currently available in-plant instrumentation and methods required for sampling stacks, monitoring process streams and reactor coolants, analyzing occupational work areas for air and surface contamination, and personnel monitoring. Off-site instrumentation and collection techniques are also presented. Conclusions are made concerning the adequacy of existing instrumentation in relation to the monitoring needs for fusion reactors

  19. Interactive nuclear plant analyzer for the VVER-440 reactor

    International Nuclear Information System (INIS)

    Shier, W.; Kennett, R.

    1993-01-01

    An interactive nuclear plant analyzer (NPA) has been developed for a VVER-440 model 213 reactor for use in the training of plant personnel, the development and verification of plant operating procedures, and in the analysis of various anticipated operational occurrences and accident scenarios. This NPA is operational on an IBM RISC-6000 workstation and utilizes the RELAP5/MOD2 computer code for the calculation of the VVER-440 reactor response to the interactive commands initiated by the NPA operator. Results of the interactive calculation can be through the user-defined, digital display of various plant parameters and through color changes that reflect changes in primary system fluid temperatures, fuel and clad temperatures, and the temperatures of other metal structures. In addition, changes in the status of various components and system can be initiated and/or displayed both numerically and graphically on the mask

  20. Capital cost: gas cooled fast reactor plant

    International Nuclear Information System (INIS)

    1977-09-01

    The results of an investment cost study for a 900 MW(e) GCFR central station power plant are presented. The capital cost estimate arrived at is based on 1976 prices and a conceptual design only, not a mature reactor design

  1. Passive cooling containment study

    International Nuclear Information System (INIS)

    Shin, J.J.; Iotti, R.C.; Wright, R.F.

    1993-01-01

    Pressure and temperature transients of nuclear reactor containment following postulated loss of coolant accident with a coincident station blackout due to total loss of all alternating current power are studied analytically and experimentally for the full scale NPR (New Production Reactor). All the reactor and containment cooling under this condition would rely on the passive cooling system which removes reactor decay heat and provides emergency core and containment cooling. Containment passive cooling for this study takes place in the annulus between containment steel shell and concrete shield building by natural convection air flow and thermal radiation. Various heat transfer coefficients inside annular air space were investigated by running the modified CONTEMPT code CONTEMPT-NPR. In order to verify proper heat transfer coefficient, temperature, heat flux, and velocity profiles were measured inside annular air space of the test facility which is a 24 foot (7.3m) high, steam heated inner cylinder of three foot (.91m) diameter and five and half foot (1.7m) diameter outer cylinder. Comparison of CONTEMPT-NPR and WGOTHIC was done for reduced scale NPR

  2. Chinese nuclear heating test reactor and demonstration plant

    International Nuclear Information System (INIS)

    Wang Dazhong; Ma Changwen; Dong Duo; Lin Jiagui

    1992-01-01

    In this report the importance of nuclear district heating is discussed. From the viewpoint of environmental protection, uses of energy resources and transport, the development of nuclear heating in China is necessary. The development program of district nuclear heating in China is given in the report. At the time being, commissioning of the 5 MW Test Heating Reactor is going on. A 200 MWt Demonstration Plant will be built. In this report, the main characteristics of these reactors are given. It shows this type of reactor has a high inherent safety. Further the report points out that for this type of reactor the stability is very important. Some experimental results of the driving facility are included in the report. (orig.)

  3. A capital cost reduction study on the fast breeder reactor plant

    International Nuclear Information System (INIS)

    Taniyama, H.; Kamei, M.; Moriyama, M.

    1991-01-01

    A capital cost reduction study has been performed for large fast breeder reactor designs. The primary objective of this study is to show a trend of capital cost reduction between FBR plants at the prototype stage, the demonstration stage, and the future commercialization stage. For the FBR plant at the demonstration stage a construction cost comparison with a light water reactor has also been performed, and the target cost of FBR of below 1.5 times that of the light water reactor cost was achieved. To extend the capital cost reduction study, a feasibility study was made to achieve a capital cost of an FBR less than that of a light water reactor. The recommended design is shown as a future commercialization FBR design concept. (author)

  4. Reactor limitation system improves the safety and availability of the Angra 2 nuclear power plant

    International Nuclear Information System (INIS)

    Souza Mendes, J.E. de

    1987-01-01

    Beyond the classic Reactor Protection System and Reactor Control System, nuclear plant Angra 2 has a third system called Reactor Limitation System which combines the intelligence features of the control systems with the high reliability of the protection systems. In determined events, which are not controlled by the control system (e.g.: load rejection, failure of one main reactor coolant pump), the Reactor Limitation System actuates automatically in order to lead the plant to a safe operating condition and so it avoids the actuation of the Reactor Protection System and consequently the reactor trip. This increases safety and availability of the plant and reduces component stresses. After the safe operating condition is reached, the process guidance automatically returns to the control systems. (Author) [pt

  5. Preliminary analysis of the inclusion of security passive systems to reduce the impact of accidents in nuclear power plants

    International Nuclear Information System (INIS)

    Sanchez J, J.; Morales S, J. B.; Espinosa P, G.

    2011-11-01

    In this work is presented a conceptual analysis of possible benefits and limitations that potentially represents the introduction of security passive systems to reduce the events impact of very low probability and high risk to the systems of radioactive material confinement of a light water reactor. These events are related with the possibility that a hydrogen explosion is presented as consequence of the accumulation of the same hydrogen in the contention of the reactor, in a scenario of severe accident. This accumulated hydrogen can be liberated in the reactor building or primary contention, where the conditions of their atmosphere make but prone the combustion. The catalytic recombination represents a viable option for the hydrogen concentrations decrease and because this recombination is highly exothermic, is important to analyze and to know if the recombined vapor to high temperature can be used in combination with vapor injectors that in turn, they are passive systems. In the following sections an explanation is presented about the use of the mentioned systems, as well as some results on the behavior of the vapor injectors. (Author)

  6. Functional and performance evaluation of 28 bar hot shutdown passive valve (HSPV) at integral test loop (ITL) for advanced heavy water reactor (AHWR)

    International Nuclear Information System (INIS)

    Sapra, M.K.; Kundu, S.; Pal, A.K.; Sharma, B.S.V.G.

    2007-02-01

    During reactor shutdown in advanced heavy water reactor (AHWR), core decay heat is removed by eight isolation condensers (IC) submerged in gravity driven water pool. Passive valves are provided on the down stream of each isolation condenser. On increase in steam drum pressure beyond a set value, these passive valves start opening and establish steam flow by natural circulation between the four steam drums and corresponding isolation condensers under hot shutdown and therefore they are termed as Hot Shut Down Passive Valves (HSPVs). The HSPV is a self acting type valve requiring no external energy, i.e. neither air nor electric supply for actuation. This feature makes the valve functioning independent of external systems such as compressed air supply or electric power supply, thereby providing inherent safety feature in line with reactor design philosophy. The high pressure and high temperature HSPV s for nuclear reactor use, are non-standard valves and therefore not manufactured by the valve industry worldwide. In the process of design and development of a prototype valve for AHWR, a 28 bar HSPV was configured and successfully tested at Integral Test Loop (ITL) at Engineering Hall No.7. During ten continuous experiments spread over 14 days, the HSPV has proved its functional capabilities and its intended use in decay heat removal system. The in-situ pressure setting and calibration aspect of HSPV has also been successfully established during these experiments. This report gives an insight into the HSPV's functional behavior and role in reactor decay heat removal system. The report not only provides the quantitative measure of performance for 28 bar HSPV in terms of valve characteristics, pressure controllability, linearity and hysteresis but also sets qualitative indicators for prototype 80 bar HSPV, being developed for AHWR. (author)

  7. Plant with nuclear reactor, in particular a thermal reactor

    International Nuclear Information System (INIS)

    Straub, H.

    1988-01-01

    The reactor core of the plant has tubular and vertically movable control rods moved by a flow of coolant under pressure. Each control rod surrounds a similarly tubular guide rod, stationary relative to the reactor core, leaving an annular slot-like space therebetween. The inside of each guide rod forms a first pressure chamber supplied with the coolant under pressure. The upper end of each control rod is closed and has a vertical shaft that extends into the inside of the guide rod and forms therewith a second annular slot-like space. At least one first restriction is provided in the first annular slot-like space and at least one second restriction is provided in the second annular slot-like space. A second pressure chamber is formed between both restrictions. The coolant supplied to the guide rod thus returns to the pressure vessel surrounding the reactor core through the second annular slot-like space, the second pressure chamber and the first annular slot-like space. Controlling means are provided, with which pressure thrusts can be generated if necessary in the coolant within the first pressure chamber. (author) 5 refs., 10 figs

  8. Significance assessment of small-medium sized reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kanno, Minoru [Japan Atomic Power Co., Research and Development Dept., Tokyo (Japan)

    2002-12-01

    Preliminary assessment for deployment of small-medium sized reactor (S and M reactor) as a future option has been conducted at the JAPCO (Japan Atomic Power Company) under the cooperation with the CRIERI (Central Research Institute of Electric Power Industry). Significance of the S and M reactor introduction is listed as follows; lower investment cost, possible siting near demand side, enlarged freedom of siting, shorter transmission line, good compatibility with slow increase of demand and plain explanation of safety using simpler system such as integral type vessel without piping, natural convection core cooling and passive safety system. The deployment of simpler plant system, modular shop fabrication, ship-shell structured building and longer operation period can assure economics comparable with that of a large sized reactor, coping with scale-demerit. Also the S and M reactor is preferable in size for the nuclear heat utilization such as hydrogen production. (T. Tanaka)

  9. Nuclear reactor plants and control systems therefor

    International Nuclear Information System (INIS)

    de Boer, G.A.; de Hex, M.

    1976-01-01

    A nuclear reactor plant is described comprising at least two hydraulically separated but thermally interconnected heat conveying circuits, of which one is the reactor circuit filled with a non-water medium and the other one is the water-steam-circuit equipped with a steam generator, a feed water conduit controlled by a valve and a steam turbine, and a control system mainly influenced by the pressure drop caused in said feed water conduit and its control valve and having a value of at least 10 bars at full load

  10. Safety significance of ATR passive safety response attributes

    International Nuclear Information System (INIS)

    Atkinson, S.A.

    1990-01-01

    The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory was designed with some passive safety response attributes which contribute to the safety of the facility. The three passive safety attributes being evaluated in the paper are: 1) In-core and in-vessel natural convection cooling, 2) a passive heat sink capability of the ATR primary coolant system (PCS) for the transfer of decay power from the uninsulated piping to the confinement, and 3) gravity feed of emergency coolant makeup. The safety significance of the ATR passive safety response attributes is that the reactor can passively respond to most transients, given a reactor scram, to provide adequate decay power removal and a significant time for operator action should the normal active heat removal systems and their backup systems both fail. The ATR Interim Level 1 Probabilistic Risk Assessment (PRA) models and results were used to evaluate the significance to ATR fuel damage frequency (or probability) of the above three passive response attributes. The results of the evaluation indicate that the first attribute is a major safety characteristic of the ATR. The second attribute has a noticeable but only minor safety significance. The third attribute has no significant influence on the ATR firewater injection system (emergency coolant system)

  11. Study on diverse passive decay heat removal approach and principle

    International Nuclear Information System (INIS)

    Lin Qian; Si Shengyi

    2012-01-01

    Decay heat removal in post-accident is one of the most important aspects concerned in the reactor safety analysis. Passive decay heat removal approach is used to enhance nuclear safety. In advanced reactors, decay heat is removed by multiple passive heat removal paths through core to ultimate heat sink by passive residual heat removal system, passive injection system, passive containment cooling system and so on. Various passive decay heat removal approaches are summarized in this paper, the common features and differences of their heat removal paths are analyzed, and the design principle of passive systems for decay heat removal is discussed. It is found that. these decay heat removal paths is combined by some basic heat transfer processes, by the combination of these basic processes, diverse passive decay heat removal approach or system design scheme can be drawn. (authors)

  12. Technological aspects of intrinsically safe and economical reactor (ISER)

    International Nuclear Information System (INIS)

    Yamada, Nobuyuki; Oda, Junro; Yamanaka, Kazuo; Sugawara, Ichiro.

    1987-01-01

    ISER is a modified version of process inherent ultimate safe reactor (PIUS) developed by ASEA-ATOM, Sweden, and follows the same inherent safety principle, that is, passive reactor shutdown through the introduction of borated pool water into a core via an interface, and passive decay heat removal by natural circulation. The most significant deviation from the PIUS is that the ISER employs a steel reactor pressure vessel enclosed in the reactor pit, instead of a prestressed concrete reactor pressure vessel of the PIUS. The merits of using steel pressure vessels are siting versatility including barge-mounted plants, low cost, the standardization and serial production of total NSSSs through the weight reduction and compaction of primary system, as well as the possibility of utilizing current LWR technology, which minimizes R and D effort. In this paper, the design features of the latest version of ISERs are shown, and the specific problems of the key components are discussed. The primary system consists of a primary coolant loop and a borated water pool, which are connected with upper and lower interfaces. The nuclear design and thermohydraulic design, the operation and maintenance, and the design features of a steam generator, a pressurizer, interfaces and so on are described. (Kako, I.)

  13. Engineering reliability in design phase: An application to AP-600 reactor passive safety system

    International Nuclear Information System (INIS)

    Majumdr, D.; Siahpush, A.S.; Hills, S.W.

    1992-01-01

    A computerized reliability enhancement methodology is described that can be used at the engineering design phase to help the designer achieve a desired reliability of the system. It can take into account the limitation imposed by a constraint such as budget, space, or weight. If the desired reliability of the system is known, it can determine the minimum reliabilities of the components, or how many redundant components are needed to achieve the desired reliability. This methodology is applied to examine the Automatic Depressurization System (ADS) of the new passively safe AP-600 reactor. The safety goal of a nuclear reactor dictates a certain reliability level of its components. It is found that a series parallel valve configuration instead of the parallel-series configuration of the four valves in one stage would improve the reliability of the ADS. Other valve characteristics and arrangements are explored to examine different reliability options for the system

  14. On the status of the EFR Euro-Breeder and its passive safety features

    International Nuclear Information System (INIS)

    Marth, W.

    1992-01-01

    The Project of the EFR, the European Fast Reactor, is characterized by close European cooperation among power utilities, plant vendors, and research centers. In the present phase up until 1993 a consistent design of the nuclear part of the plant is being elaborated with the inclusion of a site-independent safety report. The most important design features, especially those in the field of passive safety, must be backed up by reliable R and D findings. These findings will enable the ad hoc Safety Club, a body of European safety experts, to pass its vote on the general licensability of the plant concept. (orig.) [de

  15. Use of a temperature-initiated passive cooling system (TIPACS) for the modular high-temperature gas-cooled reactor cavity cooling system (RCCS)

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Conklin, J.; Reich, W.J.

    1994-04-01

    A new type of passive cooling system has been invented (Forsberg 1993): the Temperature-Initiated Passive Cooling System (TIPACS). The characteristics of the TIPACS potentially match requirements for an improved reactor-cavity-cooling system (RCCS) for the modular high-temperature gas-cooled reactor (MHTGR). This report is an initial evaluation of the TIPACS for the MHTGR with a Rankines (steam) power conversion cycle. Limited evaluations were made of applying the TIPACS to MHTGRs with reactor pressure vessel temperatures up to 450 C. These temperatures may occur in designs of Brayton cycle (gas turbine) and process heat MHTGRs. The report is structured as follows. Section 2 describes the containment cooling issues associated with the MHTGR and the requirements for such a cooling system. Section 3 describes TIPACS in nonmathematical terms. Section 4 describes TIPACS's heat-removal capabilities. Section 5 analyzes the operation of the temperature-control mechanism that determines under what conditions the TIPACS rejects heat to the environment. Section 6 addresses other design and operational issues. Section 7 identifies uncertainties, and Section 8 provides conclusions. The appendixes provide the detailed data and models used in the analysis

  16. Use of a temperature-initiated passive cooling system (TIPACS) for the modular high-temperature gas-cooled reactor cavity cooling system (RCCS)

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W.; Conklin, J.; Reich, W.J.

    1994-04-01

    A new type of passive cooling system has been invented (Forsberg 1993): the Temperature-Initiated Passive Cooling System (TIPACS). The characteristics of the TIPACS potentially match requirements for an improved reactor-cavity-cooling system (RCCS) for the modular high-temperature gas-cooled reactor (MHTGR). This report is an initial evaluation of the TIPACS for the MHTGR with a Rankines (steam) power conversion cycle. Limited evaluations were made of applying the TIPACS to MHTGRs with reactor pressure vessel temperatures up to 450 C. These temperatures may occur in designs of Brayton cycle (gas turbine) and process heat MHTGRs. The report is structured as follows. Section 2 describes the containment cooling issues associated with the MHTGR and the requirements for such a cooling system. Section 3 describes TIPACS in nonmathematical terms. Section 4 describes TIPACS`s heat-removal capabilities. Section 5 analyzes the operation of the temperature-control mechanism that determines under what conditions the TIPACS rejects heat to the environment. Section 6 addresses other design and operational issues. Section 7 identifies uncertainties, and Section 8 provides conclusions. The appendixes provide the detailed data and models used in the analysis.

  17. The Concept of the Use of the Marine Reactor Plant in Small Electric Grids

    International Nuclear Information System (INIS)

    Khlopkin, N.; Makarov, V.; Pologikh, B.

    2002-01-01

    In report some aspects of the using marine nuclear reactor are considered for provision of need small non-interconnected power systems, as well as separate settlements and the mining enterprises disposed in regions with a undeveloped infrastructure. Recently for these purposes it is offered to use the nuclear small modular power plants. The required plant power for small electric grids lies within from 1 to several tens of MWe. Module can be collected and tested on machine-building plant, and then delivered in ready type to the working place on some transport, for instance, a barge. Through determined time it's possible to transport a module to the repair shop and also to the point of storage after the end of operation. Marine nuclear reactors on their powers, compactness, mass and size are ideal prototypes for creation of such modules. For instance, building at present floating power unit, intended for functioning in region of the Russian North, based on using reactor plants of nuclear icebreakers. Reliability and safety of the ship reactor are confirmed by their trouble-free operation during approximately 180 reactors-years. Unlike big stationary nuclear plant, working in base mode, power unit with marine reactor wholly capable to work in mode of the loading following. In contrast with reactor of nuclear icebreaker, advisable to increase the core lifetime and to reduce the enrichment of the uranium. This requires more uranium capacity fuel compositions and design of the core. In particular, possible transition from traditional for ship reactor of the channel core to cassette design. Other directions of evolution of the ship reactors, not touching the basic constructive decisions verified by practice, but promoting development of properties of self-security of plant are possible. Among such directions is reduction volumetric power density of a core. (author)

  18. Clinch River Breeder Reactor Plant Project: construction schedule

    International Nuclear Information System (INIS)

    Purcell, W.J.; Martin, E.M.; Shivley, J.M.

    1982-01-01

    The construction schedule for the Clinch River Breeder Reactor Plant and its evolution are described. The initial schedule basis, changes necessitated by the evaluation of the overall plant design, and constructability improvements that have been effected to assure adherence to the schedule are presented. The schedule structure and hierarchy are discussed, as are tools used to define, develop, and evaluate the schedule

  19. Reliability of reactor plant water cleanup pumps

    International Nuclear Information System (INIS)

    Pearson, J.L.

    1979-01-01

    Carolina Power and Light Company's Brunswick 2 nuclear plant experienced a high reactor water cleanup pump-failure rate until inlet temperature and flow were reduced and mechanical modifications were implemented. Failures have been zero for about one year, and water cleanup efficiency has increased

  20. Increase of efficiency of plant materials heat treatment in tubular reactors

    Directory of Open Access Journals (Sweden)

    A. V. Golubkovich

    2016-01-01

    Full Text Available In agriculture products of pyrolysis of plant materials in the form of waste of the main production can be applied as a source of heat and electric power. Besides, their use prevents ecological pollution of the soil and the atmosphere. Pyrolysis plants can be used for work with tubular reactors anywhere. Due to them farmers can dry grain, using waste heat of diesel generators, heatgenerators, boiler plants and receiving thus gaseous products, liquid and firm fractions. A technology based on cyclic and continuous plant mass movement by a piston in a pipe from a loading site to a place of unloading of a firm phase consistently through cameras of drying, pyrolysis, condensation of gaseous products. Exhaust furnace gases with a temperature up to 600 degrees Celsius are given countercurrent material movement from a power equipment. The gaseous, liquid and firm products from the pyrolysis camera are used for heat and electric power generation. Calculation of parameters of subdrying and pyrolysis cameras is necessary for effective and steady operation of the tubular reactor. The authors determined the speed of raw materials movement, and also duration of drying and pyrolysis in working chambers. An analysis of a simplified mathematical model of process was confirmed with results of experiments. Models of heat treatment of wet plant materials in tubular reactors are worked out on a basis of equality of speeds of material movement in the reactor and distribution of a temperature front in material on radius. The authors defined estimated characteristic for determination of tubular reactor productivity and size of heat, required for drying and pyrolysis.

  1. At-reactor storage of spent fuel for life-of-plant

    International Nuclear Information System (INIS)

    Fuierer, A.A.

    1990-01-01

    The management of commercial spent fuel is a fairly broad topic beginning with the discharge from a reactor, its storage on-site, its transport from the reactor site to a U.S. Department of Energy (DOE) facility, and its ultimate disposal in a geologic repository. This paper discusses spent-fuel management in the at-reactor phase. There are two basic methods for at-reactor storage of spent fuel. The first is wet storage in a pool, and the second is dry storage external to the plant in some form of cask or vault. Spent-fuel consolidation will impact the utility and the DOE waste system. Some of these impacts have a positive effect and some have a negative effect, and each will vary somewhat for each utility. Spent-fuel consolidation and life-of-plant storage will be an increased burden to utilities but will likely result in significant cost savings to the overall waste management system and by proper integration can result in significant institutional benefits

  2. The LOFA analysis of fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Yu, Z.-C.; Xie, H.

    2014-01-01

    The fusion-fission hybrid energy reactor can produce energy, breed nuclear fuel, and handle the nuclear waste, etc, with the fusion neutron source striking the subcritical blanket. The passive safety system, consisting of passive residual heat removal system, passive safety injection system and automatic depressurization system, was adopted into the fusion-fission hybrid energy reactor in this paper. Modeling and nodalization of primary loop, passive core cooling system and partial secondary loop of the fusion-fission hybrid energy reactor using RELAP5 were conducted and LOFA (Loss of Flow Accident) was analyzed. The results of key transient parameters indicated that the PRHRs could mitigate the accidental consequence of LOFA effectively. It is also concluded that it is feasible to apply the passive safety system concept to fusion-fission hybrid energy reactor. (author)

  3. The gas turbine-modular helium reactor (GT-MHR), high efficiency, cost competitive, nuclear energy for the next century

    International Nuclear Information System (INIS)

    Zgliczynski, J.B.; Silady, F.A.; Neylan, A.J.

    1994-04-01

    The Gas Turbine-Modular Helium Reactor (GT-MHR) is the result of coupling the evolution of a small passively safe reactor with key technology developments in the US during the last decade: large industrial gas turbines, large active magnetic bearings, and compact, highly effective plate-fin heat exchangers. The GT-MHR is the only reactor concept which provides a step increase in economic performance combined with increased safety. This is accomplished through its unique utilization of the Brayton cycle to produce electricity directly with the high temperature helium primary coolant from the reactor directly driving the gas turbine electrical generator. This cannot be accomplished with another reactor concept. It retains the high levels of passive safety and the standardized modular design of the steam cycle MHTGR, while showing promise for a significant reduction in power generating costs by increasing plant net efficiency to a remarkable 47%

  4. Inherent/passive safety in fusion power plants

    International Nuclear Information System (INIS)

    Piet, S.J.; Crocker, J.G.

    1986-01-01

    The concept of inherent or passive safety for fusion energy is explored, defined, and partially quantified. Four levels of safety assurance are defined, which range from true inherent safety to passive safety to protection via active engineered safeguard systems. Fusion has the clear potential for achieving inherent or passive safety, which should be an objective of fusion research and design. Proper material choice might lead to both inherent/passive safety and high mass power density, improving both safety and economics. When inherent or passive safety is accomplished, fusion will be well on the way to achieving its ultimate potential and to be a truly superior energy source for the future

  5. Steam turbine chemistry in light water reactor plants

    International Nuclear Information System (INIS)

    Svoboda, Robert; Haertel, Klaus

    2008-01-01

    Steam turbines in boiling water reactor (BWR) and pressurized water reactor (PWR) power plants of various manufacturers have been affected by corrosion fatigue and stress corrosion cracking. Steam chemistry has not been a prime focus for related research because the water in nuclear steam generating systems is considered to be of high purity. Steam turbine chemistry however addresses more the problems encountered in fossil fired power plants on all volatile treatment, where corrosive environments can be formed in zones where wet steam is re-evaporated and dries out, or in the phase transition zone, where superheated steam starts to condense in the low-pressure (LP) turbine. In BWR plants the situation is aggravated by the fact that no alkalizing agents are used in the cycle, thus making any anionic impurity immediately acidic. This is illustrated by case studies of pitting corrosion of a 12 % Cr steel gland seal and of flow-oriented corrosion attack on LP turbine blades in the phase transition zone. In PWR plants, volatile alkalizing agents are used that provide some buffering of acidic impurities, but they also produce anionic decomposition products. (orig.)

  6. Component-Level Prognostics Health Management Framework for Passive Components - Advanced Reactor Technology Milestone: M2AT-15PN2301043

    Energy Technology Data Exchange (ETDEWEB)

    Ramuhalli, Pradeep; Roy, Surajit; Hirt, Evelyn H.; Prowant, Matthew S.; Pitman, Stan G.; Tucker, Joseph C.; Dib, Gerges; Pardini, Allan F.

    2015-06-19

    This report describes research results to date in support of the integration and demonstration of diagnostics technologies for prototypical advanced reactor passive components (to establish condition indices for monitoring) with model-based prognostics methods. Achieving this objective will necessitate addressing several of the research gaps and technical needs described in previous technical reports in this series.

  7. The PANDA tests for the SWR 1000 passive containment cooling system

    International Nuclear Information System (INIS)

    Dreier, J.; Aubert, C.; Huggenberger, M.; Strassberger, H.J.; Yadigaroglu, G.

    1999-01-01

    Since 1992, Siemens has been developing the SWR 1000, a new boiling water reactor with passive safety features. This development has been performed in close co-operation with the German nuclear utilities and with support from various European partners. Within the European Union sponsored project 'BWR R+D Cluster for Innovative Passive Safety Systems' and a bilateral contract between Siemens and the Paul Scherrer Institute, the passive containment cooling system of the SWR 1000 design has been investigated in the large-scale PANDA test facility at the Paul Scherrer Institute. A series of six tests were performed to simulate transients selected to cover a range of failure assumptions and accident severity, including core heat up and hydrogen generation. The results graphically demonstrate the self regulating character of the passive heat removal systems and their effectiveness, even under severe load, in limiting the containment pressurisation. Some tentative conclusions for the SWR 1000 are drawn, to be established by detailed analyses of the data, to support models and codes for application to plant transients. (author)

  8. The PANDA tests for the SWR 1000 passive containment cooling system

    International Nuclear Information System (INIS)

    Dreier, J.; Aubert, C.; Huggenberger, M.; Strassberger, H.J.; Meseth, J.; Yadigaroglu, G.

    1999-01-01

    Since 1992, Siemens has been developing the SWR 1000, a new boiling water reactor with passive safety features. This development has been performed in close co-operation with the German nuclear utilities and with support from various European partners. Within the European Union sponsored project 'BWR R and D Cluster for Innovative Passive Safety Systems' and a bilateral contract between Siemens and the Paul Scherrer Institute, the passive containment cooling system of the SWR 1000 design has been investigated in the large-scale PANDA test facility at the Paul Scherrer Institute. A series of six tests were performed to simulate transients selected to cover a range of failure assumptions and accident severity, including core heat up and hydrogen generation. The results graphically demonstrate the self regulating character of the passive heat removal systems and their effectiveness, even under severe load, in limiting the containment pressurisation. Some tentative conclusions for the SWR1000 are drawn, to be established by detailed analyses of the data, to support models and codes for application to plant transients. (author)

  9. Use of color-change indicators to quantify passive films on the stainless steel valves of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, Cong Qian [School of Materials Science, Engineering, Dalian University of Technology, Dalian 116085 (China); Yang, Shu Kai [School of Materials Science, Engineering, Dalian University of Technology, Dalian 116085 (China); Suzhou Nuclear Power Research Institute, Suzhou 215004 (China); Zhao, Jie, E-mail: jiezhao@dlut.edu.cn [School of Materials Science, Engineering, Dalian University of Technology, Dalian 116085 (China)

    2016-02-15

    Highlights: • A facile method to evaluate passivation quality by color change indicator. • Two indicators were compared in lab and applied on vales in nuclear power plants. • It shows that the higher value of color change the worse quality of passivation. • Traditional ferroxyl solution is unstable and might impair the vale surface. • The new indicator is more practicable than the ferroxyl test for on-site inspection. - Abstract: The passive film on nuclear-grade stainless steels was evaluated by quantifying its color changes. Coloration reactions were compared by using ferroin and blue dot solutions as indicators on the basis of the measured results in a laboratory. The reactions were then applied on stainless steel valves in a nuclear power plant. The degree of color change indicates the degree of growth of a passive film. The ferroin solution exhibits higher accuracy and more stable than blue dot solution in determining passive film quality. The potentiodynamic polarization curves show that blue dot solution might cause surface damage compared with ferroin solution. The inspection result on stainless steel valves supports our laboratory result. However, stainless steel exhibited a dramatic decrease in sensitivity to blue dot because of the intrinsic instability and high acidity of this solution. Ferroin solution is superior to blue dot solution for stainless steel facilities in a nuclear power plant.

  10. Reactor core cooling device for nuclear power plant

    International Nuclear Information System (INIS)

    Tsuda, Masahiko.

    1992-01-01

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

  11. High temperature reactor module power plant. Plant and safety concept June 1986 - 38.07126.2

    International Nuclear Information System (INIS)

    1986-06-01

    The modular HTR power plant is a universally applicable energy source for the co-generation of electricity, process steam or district heating. The modular HTR concept is characterized by the fact that standardized reactor units with power ratings of 200 MJ/s (so-called modules) can be combined to form power plants with a higher power rating. Consequently the special safety features of small high-temperature reactors (HTR) are also available at higher power plant ratings. The safety features, the technical design and the mode of operation are briefly described in the following, taking a power plant with two HTR-Modules for the co-generation of electricity and process steam as an example. Due to its universal applicability and excellent safety features, the modular HTR power plant is suitable for erection on any site, but particularly on sites near other industrial plants or in densely populated areas. The co-generation of electricity and process steam or district heating with a modular HTR power plant as described here is primarily tailored to the requirements of industrial and communal consumers. The site for such a plant is a typical industrial one. The anticipated features of such sites were taken into consideration in the design of the modular HTR power plant

  12. High temperature reactor module power plant. Plant and safety concept June 1986 - 38.07126.2

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1986-06-15

    The modular HTR power plant is a universally applicable energy source for the co-generation of electricity, process steam or district heating. The modular HTR concept is characterized by the fact that standardized reactor units with power ratings of 200 MJ/s (so-called modules) can be combined to form power plants with a higher power rating. Consequently the special safety features of small high-temperature reactors (HTR) are also available at higher power plant ratings. The safety features, the technical design and the mode of operation are briefly described in the following, taking a power plant with two HTR-Modules for the co-generation of electricity and process steam as an example. Due to its universal applicability and excellent safety features, the modular HTR power plant is suitable for erection on any site, but particularly on sites near other industrial plants or in densely populated areas. The co-generation of electricity and process steam or district heating with a modular HTR power plant as described here is primarily tailored to the requirements of industrial and communal consumers. The site for such a plant is a typical industrial one. The anticipated features of such sites were taken into consideration in the design of the modular HTR power plant.

  13. The roles of EBR-II and TREAT [Transient Reactor Test] in establishing liquid metal reactor safety

    International Nuclear Information System (INIS)

    Sackett, J.I.; Lehto, W.K.; Solbrig, C.W.

    1990-01-01

    This paper examines the role of the Experimental Breeder Reactor II (EBR-II) and Transient Reactor Test (TREAT) facilities in contributing to the understanding and resolution of key safety issues in liquid metal reactor safety during the decade of the 80's. Fuels and materials testing has been carried out to address questions on fuels behavior during steady-state and upset conditions. In addition, EBR-II has conducted plant tests to demonstrate passive response to ATWS events and to develop control and diagnostic strategies for safe operation of advanced LMRs. TREAT and EBR-II complement each other and between them provide a transient testing capability that covers the whole range of concerns during overpower conditions. EBR-II, with use of the special Automatic Control Rod Drive System, can generate power change rates that overlap the lower end of the TREAT capability. 21 refs

  14. Prestressed reactor vessel for nuclear power plants

    International Nuclear Information System (INIS)

    Schoening, J.; Schwiers, H.G.

    1982-01-01

    With usual pressure vessels for nuclear reactor plants, especially for gas-cooled nuclear reactors, the load occurring due to the inner overpressure, especially the tensile load affecting the vessel top and/or bottom, their axis of inertia being horizontal, shall be compensated without a supplementary modification in design of the top and/or the bottom. This is attained by choosing an appropriate prestressing system of the vessel wall in the field the top and/or the bottom, so that the top and/or the bottom form a tension vault directed towards the interior of the vessel. (orig.) [de

  15. Main gas circulator for VG-400 reactor plant

    International Nuclear Information System (INIS)

    Mitenkov, F.M.; Kostin, V.I.; Novinskij, E.G.; Kuropatov, A.I.; Protsenko, A.N.; Smirnov, V.P.; Stolyarevskij, A.Ya.

    1988-01-01

    Principle parameters and operating conditions of the main gas circulator (MGC) in VG-400 reactor plant are presented. Brief MGC design description and experimental work scope are given. (author). 4 refs, 4 figs, 1 tab

  16. Inherently safe reactors

    International Nuclear Information System (INIS)

    Maartensson, Anders

    1992-01-01

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

  17. Survey of the passive safety systems of the BWR 1000 concept from SIEMENS

    Energy Technology Data Exchange (ETDEWEB)

    Mattern, J; Brettschuh, W; Palavecino, C [SIEMENS, Energieerzeugung, Offenbach (Germany)

    1996-12-01

    Through the use of passive safety systems and components for accident control in addition to the active systems required for plant operation, a higher degree of safety against core-endangering conditions is achieved which is no longer ruled by complex system engineering dependent on power supply and activation by I and C systems. A low core power density and large water inventories stored inside the reactor pressure vessel as well as inside and outside the containment ensure good plant behaviour in the event of transients or accidents. These passive safety systems - which required neither electric power to function nor I and C systems for actuation, being activated solely on the basis of changes in process variables such as water level, pressure and temperature - provide a grace period of more than 5 days after the onset of accident conditions before manual intervention becomes necessary. 8 figs.

  18. L-Reactor operation, Savannah River Plant: environmental assessment

    International Nuclear Information System (INIS)

    1982-08-01

    The purpose of this document is to assess the significance of the effects on the human environment of the proposed resumption of L-reactor operation at the Savannah River Plant, scheduled for October 1983. The discussion is presented under the following section headings: need for resumption of L-Reactor operations and purpose of this environmental assessment; proposed action and alternative; affected environment (including, site location and description, land use, historic and archeological resources, socioeconomic and community characteristics, geology and seismology, hydrology, meteorology and climatology, ecology, and radiation environment); environmental consequences; summary of projected L-Reactor releases and impacts; and Federal and State permits and approval. The three appendices are entitled: radiation dose calculation methods and assumptions; floodplain/wetlands assessment - L-Reactor operations; and, conversion table. A list of references is included at the end of each chapter

  19. Application of a steam injector for passive emergency core cooling during a station blackout

    International Nuclear Information System (INIS)

    Heinze, D.; Behnke, L.; Schulenberg, T.

    2012-01-01

    One of the basic protection targets of reactor safety is the safe heat removal during normal operation but also following shut-down. Since the reactor accident in Fukushima an optimization of the plant robustness in case of beyond-design accident is performed. Special attention is given to the increase of time available for starting appropriate measures for emergency core cooling in case of a station blackout. The state-of the art in engineering and research is presented. Investigations on the applicability of a steam injector for passive emergency core cooling during a station blackout in BWR-type reactors have progressed, experiments on dynamic behavior of the injector are described. A precise design with respect to the thermal hydraulic boundary conditions has been performed.

  20. Design and development of self-powered sensors on wireless sensor network for standalone plant critical data management during SBO and beyond design basis events

    International Nuclear Information System (INIS)

    Aparna, J.; Dulera, I.V.; Rama Rao, A.; Vijayan, P.K.

    2015-01-01

    Advanced reactors are designed with an aim of maximum safety, optimized fuel utilization and effective system design. Safety aspects in reactor designs are being viewed for all possible vulnerabilities, and as a result, robust self-regulating passive safety features have been favored in Gen IV and advanced reactor designs. In addition to passive systems, the accidents scenarios at Fukushima indicate the dire need of reliable and stand-alone self-powered sensors, for monitoring plant critical parameters for effective damage control actions. There is a strong need for plant critical data management and situation awareness during the unavailability of all conventional power sources in a nuclear power plant, during extended station blackout (SBO) conditions. These self-powered sensors would assist the operators in managing events like SBO and help in containing any Beyond Design Basis Events (BDBE) conditions, well away from the public domain

  1. DESAIN KONSEP TANGKI PENAMPUNG BAHAN BAKAR PASSIVE COMPACT MOLTEN SALT REACTOR

    Directory of Open Access Journals (Sweden)

    A. Hadiwinata

    2015-04-01

    Full Text Available Passive Compact Molten Salt Reactor (PCMSR merupakan pengembangan dari reaktor MSR. Desain reaktor PCMSR membutuhkan tempat khusus penampung sementara bahan bakar pada saat terjadi insiden, misalnya kecelakaan yang menyebabkan peningkatan suhu bahan bakar. Tangki penampung bahan bakar tersusun dari 3 bagian yang saling terhubung yaitu bagian penampung cairan bahan bakar, cerobong (chimney, dan penukar kalor. Dalam penelitian ini, tangki dimodelkan secara lump dan dilakukan variasi daya awal reaktor dan ketinggian cerobong. Syarat batas model ditetapkan suhu bahan bakar maksimum 1400 °C, yang didasarkan pada titik didih larutan garam LiF-BeF2-ThF4-UF4. Analisis dilakukan dengan cara menghitung rugi tekanan total dan transfer kalor untuk variasi daya awal antara 1800-3000 MWth dan ketinggian cerobong antara 1-10 m. Hasil penelitian menunjukan semakin besar daya reaktor, maka tinggi tangki penampung bahan bakar dan tinggi alat penukar kalor yang dibutuhkan akan semakin besar, tejadi kenaikan suhu fluida pendingin dan suhu udara pendingin, dan menyebabkan kenaikan laju aliran masa fluida pendingin, sedangkan laju aliran masa udara menurun. Peningkatan ketinggian cerobong menyebabkan ketinggian tangki penampung bahan bakar dan ketinggian alat penukar kalor semakin menurun, penurunan suhu fluida pendingin, tetapi suhu udara meningkat, dan menyebabkan peningkatan laju aliran masa fluida pendingin, tetapi laju aliran masa udara akan semakin menurun. Kata kunci: PCMSR, cerobong, alat penukar kalor, variasi daya.   The Passsive Compact Molten Salat Reactor (PCMSR reactor is developed from MSR reactor. The PCMSR reactor design requires special place to temporarily storage for reactor fuel when incident occurs, such as when there is an accident which caused the temperature of the fuel increases. The tank consist of three interconnected parts, the reservoir liquid fuel, chimney, and the heat exchanger. In this research, the tank system is modeled based on

  2. Overall plant concept for a tank-type fast reactor

    International Nuclear Information System (INIS)

    Yamaki, Hideo; Davies, S.M.; Goodman, L.

    1984-01-01

    Japanese nuclear industries are expressing interest in the merits of the tank-type FBR as a large plant (demonstration) after JOYO (experimental, in operation) and MONJU (prototype, under construction). In response to this growing interest in a tank-type FBR demonstration plant, Hitachi has initiated a conceptual study of a 1000 MWe tank plant concept in collaboration with GE and Bechtel. Key objectives of this study have been: to select reliable and competitive tank plant concepts, with emphases on a seismic-resistant and compact tank reactor system;to select reliable shutdown heat removal system;and to identify R and D items needed for early 1990s construction. Design goals were defined as follows: capital costs must be less than twice, and as close as practical to 1.5 those of equivalent LWR plants;earthquake resistant structures to meet stringent Japanese seismic conditions must be as simple and reliable as practical;safety must be maintained at LWR-equivalent risks;and R and D needs must be limited to minimum cost for the limited time allowed. This paper summarizes the overall plant concepts with some selected topics, whereas detailed descriptions of the reactor assembly and the layout design are found in separate papers

  3. Design Configurations and Coupling High Temperature Gas-Cooled Reactor and Hydrogen Plant

    International Nuclear Information System (INIS)

    Chang H. Oh; Eung Soo Kim; Steven Sherman

    2008-01-01

    The US Department of Energy is investigating the use of high-temperature nuclear reactors to produce hydrogen using either thermochemical cycles or high-temperature electrolysis. Although the hydrogen production processes are in an early stage of development, coupling either of these processes to the high-temperature reactor requires both efficient heat transfer and adequate separation of the facilities to assure that off-normal events in the production facility do not impact the nuclear power plant. An intermediate heat transport loop will be required to separate the operations and safety functions of the nuclear and hydrogen plants. A next generation high-temperature reactor could be envisioned as a single-purpose facility that produces hydrogen or a dual-purpose facility that produces hydrogen and electricity. Early plants, such as the proposed Next Generation Nuclear Plant (NGNP), may be dual-purpose facilities that demonstrate both hydrogen and efficient electrical generation. Later plants could be single-purpose facilities. At this stage of development, both single- and dual-purpose facilities need to be understood

  4. Plant Control Concept for the Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Kim, Eui Kwang; Kim, S. O.

    2010-12-01

    A power plant is designed for incorporation into a utility's grid system and follows the load demand through the steam generator, intermediate heat exchanger(IHX), from the nuclear core. During the load-following transients, various plant parameters must be controlled to protect the reactor core and other components in the plant. The purpose of this report is to review design considerations to establish SFR plant control and to design plant control concepts. The governing equations and solution procedure of the computer code to calculate plant temperature conditions during the part-load operation was reviewed and 4 types of plant operation concepts were designed, and the results of the calculations were compared

  5. Knowledge and abilities catalog for nuclear power plant operators: Pressurized water reactors. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-08-01

    This document provides the basis for the development of content-valid licensing examinations for reactor operators and senior reactor operators. The examinations developed using the PWR catalog will cover those topics listed under Title 10, (ode of Federal Regulations Part 55. The PWR catalog contains approximately 5100 knowledge and ability (K/A) statements for reactor operators and senior reactor operators. The catalog is organized into six major sections: Catalog Organization; Generic Knowledge and Abilities; Plant Systems; Emergency and Abnormal Plant Evolutions; Components and Theory.

  6. Knowledge and abilities catalog for nuclear power plant operators: Pressurized water reactors. Revision 1

    International Nuclear Information System (INIS)

    1995-08-01

    This document provides the basis for the development of content-valid licensing examinations for reactor operators and senior reactor operators. The examinations developed using the PWR catalog will cover those topics listed under Title 10, (ode of Federal Regulations Part 55. The PWR catalog contains approximately 5100 knowledge and ability (K/A) statements for reactor operators and senior reactor operators. The catalog is organized into six major sections: Catalog Organization; Generic Knowledge and Abilities; Plant Systems; Emergency and Abnormal Plant Evolutions; Components and Theory

  7. A new design concept for offshore nuclear power plants with enhanced safety features

    International Nuclear Information System (INIS)

    Lee, Kihwan; Lee, Kang-Heon; Lee, Jeong Ik; Jeong, Yong Hoon; Lee, Phill-Seung

    2013-01-01

    Highlights: ► A new design concept for offshore nuclear power plants is proposed. ► The total general arrangement for the concept is suggested. ► A new emergency passive containment cooling system (EPCCS) is proposed. ► A new emergency passive reactor-vessel cooling system (EPRVCS) is proposed. ► Safety features against earthquakes, tsunamis, and storms are discussed. - Abstract: In this paper, we present a new concept for offshore nuclear power plants (ONPP) with enhanced safety features. The design concept of a nuclear power plant (NPP) mounted on gravity-based structures (GBSs), which are widely used offshore structures, is proposed first. To demonstrate the feasibility of the concept, a large-scale land-based nuclear power plant model APR1400, which is the most recent NPP model in the Republic of Korea, is mounted on a GBS while minimizing modification to the original features of APR1400. A new total general arrangement (GA) and basic design principles are proposed and can be directly applied to any existing land based large scale NPPs. The proposed concept will enhance the safety of a NPP due to several aspects. A new emergency passive containment cooling system (EPCCS) and emergency passive reactor-vessel cooling system (EPRVCS) are proposed; their features of using seawater as coolant and safety features against earthquakes, Tsunamis, storms, and marine collisions are also described. We believe that the proposed offshore nuclear power plant is more robust than conventional land-based nuclear power plants and it has strong potential to provide great opportunities in nuclear power industries by decoupling the site of construction and that of installation.

  8. Nuclear reactor vessel fuel thermal insulating barrier

    Science.gov (United States)

    Keegan, C. Patrick; Scobel, James H.; Wright, Richard F.

    2013-03-19

    The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel that has a hemispherical lower section that increases in volume from the center line of the reactor to the outer extent of the diameter of the thermal insulating barrier and smoothly transitions up the side walls of the vessel. The space between the thermal insulating harrier and the reactor vessel forms a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive inlet valve for the cooling water includes a buoyant door that is normally maintained sealed under its own weight and floats open when the cavity is Hooded. Passively opening steam vents are also provided.

  9. A new advanced safe nuclear reactor concept

    International Nuclear Information System (INIS)

    Sefidvash, Farhang

    1999-01-01

    The reactor design is based on fluidized bed concept and utilizes pressurized water reactor technology. The fuel is automatically removed from the reactor by gravity under any accident condition. The reactor demonstrates the characteristics of inherent safety and passive cooling. Here two options for modification to the original design are proposed in order to increase the stability and thermal efficiency of the reactor. A modified version of the reactor involves the choice of supercritical steam as the coolant to produce a plant thermal efficiency of about 40%. Another is to modify the shape of the reactor core to produce a non-fluctuating bed and consequently guarantee the dynamic stability of the reactor. The mixing of Tantalum in the fuel is also proposed as an additional inhibition to power excursion. The spent fuel pellets may not be considered nuclear waste since they are in the shape and size that can easily be used as a a radioactive source for food irradiation and industrial applications. The reactor can easily operate with any desired spectrum by varying the porosity in order to be a plutonium burner or utilize a thorium fuel cycle. (author)

  10. Decommissioning strategy and schedule for a multiple reactor nuclear power plant site

    Energy Technology Data Exchange (ETDEWEB)

    Monteiro, Deiglys Borges; Moreira, Joao M.L.; Maiorino, Jose Rubens, E-mail: deiglys.monteiro@ufabc.edu.br, E-mail: joao.moreira@ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br [Universidade Federal do ABC (CECS/UFABC), Santo Andre, SP (Brazil). Centro de Engenharia, Modelagem e Ciencias Aplicadas

    2015-07-01

    The decommissioning is an important part of every Nuclear Power Plant life cycle gaining importance when there are more than one plant at the same site due to interactions that can arise from the operational ones and a decommissioning plant. In order to prevent undesirable problems, a suitable strategy and a very rigorous schedule should implemented and carried. In this way, decommissioning tasks such as fully decontamination and dismantling of activated and contaminated systems, rooms and structures could be delayed, posing as an interesting option to multiple reactor sites. The present work aims to purpose a strategy and a schedule for the decommissioning of a multiple reactor site highlighting the benefits of delay operational tasks and constructs some auxiliary services in the site during the stand by period of the shutdown plants. As a case study, will be presented a three-reactor site which the decommissioning process actually is in planning stage and that should start in the next decade. (author)

  11. Decommissioning strategy and schedule for a multiple reactor nuclear power plant site

    International Nuclear Information System (INIS)

    Monteiro, Deiglys Borges; Moreira, Joao M.L.; Maiorino, Jose Rubens

    2015-01-01

    The decommissioning is an important part of every Nuclear Power Plant life cycle gaining importance when there are more than one plant at the same site due to interactions that can arise from the operational ones and a decommissioning plant. In order to prevent undesirable problems, a suitable strategy and a very rigorous schedule should implemented and carried. In this way, decommissioning tasks such as fully decontamination and dismantling of activated and contaminated systems, rooms and structures could be delayed, posing as an interesting option to multiple reactor sites. The present work aims to purpose a strategy and a schedule for the decommissioning of a multiple reactor site highlighting the benefits of delay operational tasks and constructs some auxiliary services in the site during the stand by period of the shutdown plants. As a case study, will be presented a three-reactor site which the decommissioning process actually is in planning stage and that should start in the next decade. (author)

  12. Medium-Power Lead-Alloy Reactors: Missions for This Reactor Technology

    International Nuclear Information System (INIS)

    Todreas, Neil E.; MacDonald, Philip E.; Hejzlar, Pavel; Buongiorno, Jacopo; Loewen, Eric P.

    2004-01-01

    A multiyear project at the Idaho National Engineering and Environmental Laboratory and the Massachusetts Institute of Technology investigated the potential of medium-power lead-alloy-cooled technology to perform two missions: (1) the production of low-cost electricity and (2) the burning of actinides from light water reactor (LWR) spent fuel. The goal of achieving a high power level to enhance economic performance simultaneously with adoption of passive decay heat removal and modularity capabilities resulted in designs in the range of 600-800 MW(thermal), which we classify as a medium power level compared to the lower [∼100 MW(thermal)] and higher [2800 MW(thermal)] power ratings of other lead-alloy-cooled designs. The plant design that was developed shows promise of achieving all the Generation-IV goals for future nuclear energy systems: sustainable energy generation, low overnight capital cost, a very low likelihood and degree of core damage during any conceivable accident, and a proliferation-resistant fuel cycle. The reactor and fuel cycle designs that evolved to achieve these missions and goals resulted from study of the following key trade-offs: waste reduction versus reactor safety, waste reduction versus cost, and cost versus proliferation resistance. Secondary trade-offs that were also considered were monolithic versus modular design, active versus passive safety systems, forced versus natural circulation, alternative power conversion cycles, and lead versus lead-bismuth coolant.These studies led to a selection of a common modular design with forced convection cooling, passive decay heat removal, and a supercritical CO 2 power cycle for all our reactor concepts. However, the concepts adopt different core designs to optimize the achievement of the two missions. For the low-cost electricity production mission, a design approach based on fueling with low enriched uranium operating without costly reprocessing in a once-through cycle was pursued to achieve a

  13. Training nuclear power plant personnel on SR-O reactor

    International Nuclear Information System (INIS)

    Cerny, K.; Boucek, F.; Kveton, M.; Prokopec, Z.; Fleischhans, J.

    1983-01-01

    The SR-O reactor is an experimental pool-type reactor with a maximum output of 1 MW and maximum thermal neutron flux density of 5.3x10 13 m -2 s -1 . The reactor is described in detail and its specifications are given. The protection and control systems of the reactor permit both manual and automatic operation. The reactor is used for training courses for nuclear power plant operators and for post-graduate study courses for other specialists. Intensive courses for 4 to 6 persons take 15 to 20 days. The course is adjusted to the results of introductory theoretical tests. An optimal teaching method has been developed based on the flowchart algorithmic method, dividing activities into operations (manipulations with controls, issuing commands, making records, etc.) and decision making (information reception and processing). (M.D.)

  14. Plant-scale anodic dissolution of unirradiated N-Reactor fuel

    International Nuclear Information System (INIS)

    Gay, E.C.; Miller, W.E.; Laidler, J.J.

    1995-01-01

    Anodic dissolution tests were made with unirradiated N-Reactor fuel to determine the fuel segment length, diameter, and shape required for high throughput electrorefiner treatment for ultimate disposal in a geologic repository. Based on these tests, a conceptual design was produced of an electrorefiner for a full-scale plant to process N-Reactor spent fuel. In this design, the diameter of an electrode assembly is about 0.6 m (25 in.). Eight of these assemblies in an electrorefiner would accommodate a 1.333-metric-ton batch of N-Reactor fuel. Electrorefining would proceed at a rate of 40 kg uranium per hour

  15. Instrumentation and control strategies for an integral pressurized water reactor

    Directory of Open Access Journals (Sweden)

    Belle R. Upadhyaya

    2015-03-01

    Full Text Available Several vendors have recently been actively pursuing the development of integral pressurized water reactors (iPWRs that range in power levels from small to large reactors. Integral reactors have the features of minimum vessel penetrations, passive heat removal after reactor shutdown, and modular construction that allow fast plant integration and a secure fuel cycle. The features of an integral reactor limit the options for placing control and safety system instruments. The development of instrumentation and control (I&C strategies for a large 1,000 MWe iPWR is described. Reactor system modeling—which includes reactor core dynamics, primary heat exchanger, and the steam flashing drum—is an important part of I&C development and validation, and thereby consolidates the overall implementation for a large iPWR. The results of simulation models, control development, and instrumentation features illustrate the systematic approach that is applicable to integral light water reactors.

  16. Advanced light-water reactors

    International Nuclear Information System (INIS)

    Golay, M.W.; Todreas, N.E.

    1990-01-01

    Environmental concerns, economics and the earth's finite store of fossil fuels argue for a resuscitation of nuclear power. The authors think improved light-water reactors incorporating passive safety features can be both safe and profitable, but only if attention is paid to economics, effective management and rigorous training methods. The experience of nearly four decades has winnowed out designs for four basic types of reactor: the heavy-water reactor (HWR), the gas-cooled rector (GCR), the liquid-metal-cooled reactor (LMR) and the light-water reactor (LWR). Each design is briefly described before the paper discusses the passive safety features of the AP-600 rector, so-called because it employs an advanced pressurized water design and generates 600 MW of power

  17. Development of design technology for advanced pressurized water reactor

    International Nuclear Information System (INIS)

    Kim, Si Hwan; Chang, Moon Hee; Lee, Jong Chul

    1991-08-01

    In order to investigate the feasibility of the domestic passive reactor development, the analysis and evaluation on the development status, technical characteristics, and the safety and economy for the overseas passive reactors were carried out based on the vendor's information. Also the domestic nuclear technology basis was surveyed. The analysis and evaluation of the development status and technical characteristics were performed mainly for the AP-600 developed by Westing house and the SIR of UKAEA. The new design concepts and system characteristics have been evaluated by utilizing EPRI Utility Requirement Documents and Lahmeyer evaluation criteria. Based on this evaluation the recommendable design concepts in each major system were selected. The feasibility for the domestic passive reactor development has focused on the safety, technology and economy aspects, and on the applicability of the existing domestic technology to the design of the passive reactor. And the development plan for the domestic passive reactor was recommended in a step by step way. (Author)

  18. Behavior study on Na heat pipe in passive heat removal system of new concept molten salt reactor

    International Nuclear Information System (INIS)

    Wang Chenglong; Tian Wenxi; Su Guanghui; Zhang Dalin; Wu Yingwei; Qiu Suizheng

    2013-01-01

    The high temperature Na heat pipe is an effective device for transporting heat, which is characterized by remarkable advantages in conductivity, isothermally and passively working. The application of Na heat pipe on passive heat removal system of new concept molten salt reactor (MSR) is significant. The transient performance of high temperature Na heat pipe was simulated by numerical method under the MSR accident. The model of the Na heat pipe was composed of three conjugate heat transfer zones, i.e. the vapor, wick and wall. Based on finite element method, the governing equations were solved by making use of FORTRAN to acquire the profiles of the temperature, velocity and pressure for the heat pipe transient operation. The results show that the high temperature Na heat pipe has a good performance on operating characteristics and high heat transfer efficiency from the frozen state. (authors)

  19. 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

  20. Development of advanced nuclear reactors in Russia

    International Nuclear Information System (INIS)

    Sotoudeh, M.; Silakhori, K.; Sepanloo, K.; Jahanfarnia, G.; Moattar, F.

    2008-01-01

    Several advanced reactor designs have been so far developed in Russia. The AES-91 and AES-92 plants with the VVER-1000 reactors have been developed at the beginning of 1990. However, the former design has been built in China and the latest which is certified meeting European Utility Requirements is being built in India. Moreover, the model VVER-1500 reactor with 50-60 MWd/t burn-up and an enhanced safety was being developed by Gidropress about 2005, excepting to be completed in 2007. But, this schedule has slipped in favor of development of the AES-2006 power plant incorporating a third-generation standardized VVER-1200 reactor of 1170 MWe. This is an evolutionary development of the well-proven VVER-1000 reactor in the AES-92 plant, with longer life, greater power and efficiency and its lead units are being built at Novovoronezh II, to start operation in 2012-13. Based on Atomenergoproekt declaration, the AES-2006 conforms to both Russian standards and European Utility Requirements. The most important features of the AES-2006 design are mentioned as: a design based on the passive safety systems, double containment, longer plant service life of 50 years with a capacity factor of 92%, longer irreplaceable components service life of 60 years, a 28.6% lower amount of concrete and metal, shorter construction time of 54 months, a Core Damage Frequency of 1x10 -7 / year and lower liquid and solid wastes by 70% and 80% respectively. The presented paper includes a comparative analysis of technological and safety features, economic parameters and environmental impact of the AES-2006 design versus the other western advanced reactors. Since the Bushehr phase II NPP and several other NPPs are planning in Iran, such analysis would be of a great importance

  1. Assessment of ALWR passive safety system reliability. Phase 1: Methodology development and component failure quantification

    International Nuclear Information System (INIS)

    Hake, T.M.; Heger, A.S.

    1995-04-01

    Many advanced light water reactor (ALWR) concepts proposed for the next generation of nuclear power plants rely on passive systems to perform safety functions, rather than active systems as in current reactor designs. These passive systems depend to a great extent on physical processes such as natural circulation for their driving force, and not on active components, such as pumps. An NRC-sponsored study was begun at Sandia National Laboratories to develop and implement a methodology for evaluating ALWR passive system reliability in the context of probabilistic risk assessment (PRA). This report documents the first of three phases of this study, including methodology development, system-level qualitative analysis, and sequence-level component failure quantification. The methodology developed addresses both the component (e.g. valve) failure aspect of passive system failure, and uncertainties in system success criteria arising from uncertainties in the system's underlying physical processes. Traditional PRA methods, such as fault and event tree modeling, are applied to the component failure aspect. Thermal-hydraulic calculations are incorporated into a formal expert judgment process to address uncertainties in selected natural processes and success criteria. The first phase of the program has emphasized the component failure element of passive system reliability, rather than the natural process uncertainties. Although cursory evaluation of the natural processes has been performed as part of Phase 1, detailed assessment of these processes will take place during Phases 2 and 3 of the program

  2. Reactor Pressure Vessel (RPV) Acquisition Strategy

    Energy Technology Data Exchange (ETDEWEB)

    Mizia, Ronald Eugene [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2008-04-01

    The Department of Energy has selected the High Temperature Gas-cooled Reactor design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic or pebble-bed reactor and use low-enriched uranium, TRISO-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development (R&D) Program is responsible for performing R&D on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while at the same time setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. The purpose of this report is to address the acquisition strategy for the NGNP Reactor Pressure Vessel (RPV). This component will be larger than any nuclear reactor pressure vessel presently in service in the United States. The RPV will be taller, larger in diameter, thicker walled, heavier and most likely fabricated at the Idaho National Laboratory (INL) site of multiple subcomponent pieces. The pressure vessel steel can either be a conventional materials already used in the nuclear industry such as listed within ASME A508/A533 specifications or it will be fabricated from newer pressure vessel materials never before used for a nuclear reactor in the US. Each of these characteristics will present a

  3. Passive heat removal characteristics of SMART

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Jae Kwang; Kang, Hyung Seok; Yoon, Joo Hyun; Kim, Hwan Yeol; Cho, Bong Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    A new advanced integral reactor of 330 MWt thermal capacity named SMART (System-Integrated Modular Advanced Reactor) is currently under development in Korea Atomic Energy Research Institute (KAERI) for multi-purpose applications. Modular once-through steam generator (SG) and self-pressurizing pressurizer equipped with wet thermal insulator and cooler are essential components of the SMART. The SMART provides safety systems such as Passive Residual Heat Removal System (PRHRS). In this study, a computer code for performance analysis of the PRHRS is developed by modeling relevant components and systems of the SMART. Using this computer code, a performance analysis of the PRHRS is performed in order to check whether the passive cooling concept using the PRHRS is feasible. The results of the analysis show that PRHRS of the SMART has excellent passive heat removal characteristics. 2 refs., 4 figs., 1 tab. (Author)

  4. Design of passive decay heat removal system using thermosyphon for low temperature and low pressure pool type LWR

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Jangsik; You, Byung Hyun; Jung, Yong Hun; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2013-10-15

    In seawater desalination process which doesn't need high temperature steam, the reactor has profitability. KAIST has be developing the new reactor design, AHR400, for only desalination. For maximizing safety, the reactor requires passive decay heat removal system. In many nuclear reactors, DHR system is loop form. The DHR system can be designed simple by applying conventional thermosyphon, which is fully passive device, shows high heat transfer performance and simple structure. DHR system utilizes conventional thermosyphon and its heat transfer characteristics are analyzed for AHR400. For maximizing safety of the reactor, passive decay heat removal system are prepared. Thermosyphon is useful device for DHR system of low pressure and low temperature pool type reactor. Thermosyphon is operated fully passive and has simple structure. Bundle of thermosyphon get the goal to prohibit boiling in reactor and high pressure in reactor vessel.

  5. Nuclear reactor cavity floor passive heat removal system

    Science.gov (United States)

    Edwards, Tyler A.; Neeley, Gary W.; Inman, James B.

    2018-03-06

    A nuclear reactor includes a reactor core disposed in a reactor pressure vessel. A radiological containment contains the nuclear reactor and includes a concrete floor located underneath the nuclear reactor. An ex vessel corium retention system includes flow channels embedded in the concrete floor located underneath the nuclear reactor, an inlet in fluid communication with first ends of the flow channels, and an outlet in fluid communication with second ends of the flow channels. In some embodiments the inlet is in fluid communication with the interior of the radiological containment at a first elevation and the outlet is in fluid communication with the interior of the radiological containment at a second elevation higher than the first elevation. The radiological containment may include a reactor cavity containing a lower portion of the pressure vessel, wherein the concrete floor located underneath the nuclear reactor is the reactor cavity floor.

  6. Development of safety analysis technology for integral reactor; evaluation on safety concerns of integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hee Chul; Kim, Woong Sik; Lee, J. H. [Korea Institute of Nuclear Safety, Taejeon (Korea)

    2002-03-01

    The Nuclear Desalination Plant (NDP) is being developed to produce electricity and fresh water, and is expected to locate near population zone. In the aspect of safety, it is required to protect the public and environment from the possible releases of fission products and to prevent the fresh water from the contamination of radioactivity. Thus, in this study, the safety characteristics of the integral reactor adopting passive and inherent safety features significantly different from existing nuclear power plants were investigated. Also, safety requirements applicable to the NDP were analyzed based on the regulatory requirements for current light water reactor and advanced reactor designs, and user requirements for small-medium size reactors. Based on these analyses, some safety concerns to be considered in the design stage have been identified and discussed. They include the use of proven technology for new safety features, systematic event classification and selection, strengthening containment function, and the safety impacts on desalination-related systems. The study presents the general safety requirements applicable to licensing of an integral reactor and suggests additional regulatory requirements, which need to be developed, based on the direction to resolution of the safety concerns. The efforts to identify and technically resolve the safety concerns in the design stage will provide the early confidence of SMART safety and the technical basis to evaluate the safety to designers and reviewers in the future. Suggestion on the development of additional regulatory requirements will contribute for the regulator to taking actions for licensing of an integral reactor. 66 refs., 5 figs., 24 tabs. (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. Passive Flora? Reconsidering Nature’s Agency through Human-Plant Studies (HPS

    Directory of Open Access Journals (Sweden)

    John Charles Ryan

    2012-08-01

    Full Text Available Plants have been—and, for reasons of human sustenance and creative inspiration, will continue to be—centrally important to societies globally. Yet, plants—including herbs, shrubs, and trees—are commonly characterized in Western thought as passive, sessile, and silent automatons lacking a brain, as accessories or backdrops to human affairs. Paradoxically, the qualities considered absent in plants are those employed by biologists to argue for intelligence in animals. Yet an emerging body of research in the sciences and humanities challenges animal-centred biases in determining consciousness, intelligence, volition, and complex communication capacities amongst living beings. In light of recent theoretical developments in our understandings of plants, this article proposes an interdisciplinary framework for researching flora: human-plant studies (HPS. Building upon the conceptual formations of the humanities, social sciences, and plant sciences as advanced by Val Plumwood, Deborah Bird Rose, Libby Robin, and most importantly Matthew Hall and Anthony Trewavas, as well as precedents in the emerging areas of human-animal studies (HAS, I will sketch the conceptual basis for the further consideration and exploration of this interdisciplinary framework.

  9. Survey of Regulations Applicable to the Finned Containment in Korean Nuclear Power Plant for Light Water Reactor

    International Nuclear Information System (INIS)

    Noh, Hyung Gyun; Kang, Hie Chan

    2016-01-01

    In severe accident, the molten corium would discharge into the reactor cavity and interact with water and concrete of cavity. Molten corium includes non-oxidation metals such as Zr, Fe and Cr. These metal species reacted with water emit hydrogen gas. In addition to this, a mount of steam can be emitted to the containment such as steam line break accident. As a result, steam and hydrogen gas can pressurize containment over the design pressure and threaten its integrity. For this reasons, a concept equipped with finned on the containment building was proposed for coping with prolonged accident. Finned containment can enhance heat transfer to the ambient, and the building itself is working as a heat sink. Multiple metal fins and metal rod are penetrated into containment wall, and the rods are working as an additional path of heat removal. To be accepted in the nuclear power plants, this configuration should satisfy the requirement of heat removal and follow all regulations related with containment also. For applying to Korean nuclear power plants, the finned containment should follow all regulations specialized in Korea such as Nuclear regulatory criteria for light water reactor and Guidelines of nuclear safety examination for light water reactor. A concept of containment as a passive cooling system has been proposed. Furthermore, the new containment concept can be applied on the real containment which satisfies the various regulations. Finned containment would be expected positive effects on heat removal from the containment. If the fins are properly welded to the liner, finned containment could satisfy the leak tightness and prevention of external influences. Finned containment could be favorable to protect external impact like aircraft crash because of the additional structural integrity by the fins

  10. Survey of Regulations Applicable to the Finned Containment in Korean Nuclear Power Plant for Light Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Noh, Hyung Gyun [Pohang University, Pohang (Korea, Republic of); Kang, Hie Chan [Kunsan University, Gunsan (Korea, Republic of)

    2016-05-15

    In severe accident, the molten corium would discharge into the reactor cavity and interact with water and concrete of cavity. Molten corium includes non-oxidation metals such as Zr, Fe and Cr. These metal species reacted with water emit hydrogen gas. In addition to this, a mount of steam can be emitted to the containment such as steam line break accident. As a result, steam and hydrogen gas can pressurize containment over the design pressure and threaten its integrity. For this reasons, a concept equipped with finned on the containment building was proposed for coping with prolonged accident. Finned containment can enhance heat transfer to the ambient, and the building itself is working as a heat sink. Multiple metal fins and metal rod are penetrated into containment wall, and the rods are working as an additional path of heat removal. To be accepted in the nuclear power plants, this configuration should satisfy the requirement of heat removal and follow all regulations related with containment also. For applying to Korean nuclear power plants, the finned containment should follow all regulations specialized in Korea such as Nuclear regulatory criteria for light water reactor and Guidelines of nuclear safety examination for light water reactor. A concept of containment as a passive cooling system has been proposed. Furthermore, the new containment concept can be applied on the real containment which satisfies the various regulations. Finned containment would be expected positive effects on heat removal from the containment. If the fins are properly welded to the liner, finned containment could satisfy the leak tightness and prevention of external influences. Finned containment could be favorable to protect external impact like aircraft crash because of the additional structural integrity by the fins.

  11. Experimental investigation of a two-phase closed thermosyphon assembly for passive containment cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Kyung Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Sang Nyung [Kyunghee Univ., Gyeonggi-do (Korea, Republic of)

    2017-06-15

    After the Fukushima accident, increasing interest has been raised in passive safety systems that maintain the integrity of the containment building. To improve the reliability and safety of nuclear power plants, long-term passive cooling concepts have been developed for advanced reactors. In a previous study, the proposed design was based on an ordinary cylindrical Two-Phase Closed Thermosyphon (TPCT). The exact assembly size and number of TPCTs should be elaborated upon through accurate calculations based on experiments. While the ultimate goal is to propose an effective MPHP design for the PCCS and experimentally verify its performance, a TPCT assembly that was manufactured based on the conceptual design in this paper was tested.

  12. Clinch River Breeder Reactor Plant: a building block in nuclear technology

    International Nuclear Information System (INIS)

    McCormack, M.

    1979-01-01

    Interest in breeder reactors dates from the Manhatten Project to the present effort to build the Clinch River Liquid Metal Fast Breeder Reactor (LMFBR) demonstration plant. Seven breeder-type reactors which were built during this time are described and their technological progress assessed. The Clinch River Breeder Reactor Project (CRBRP) has been designed to demonstrate that it can be licensed, can operate on a large power grid, and can provide industry with important experience. As the next logical step in LMFBR development, the project has suffered repeated cancellation efforts with only minor modifications to its schedule. Controversies have developed over the timing of a large-scale demonstration plant, the risks of proliferation, economics, and other problems. Among the innovative developments adopted for the CRBRP is a higher thermal efficiency potential, the type of development which Senator McCormack feels justifies continuing the project. He argues that the nuclear power program can and should be revitalized by continuing the CRBRP

  13. Clinch River Breeder Reactor Plant: a building block in nuclear technology

    Energy Technology Data Exchange (ETDEWEB)

    McCormack, M.

    1979-01-01

    Interest in breeder reactors dates from the Manhatten Project to the present effort to build the Clinch River Liquid Metal Fast Breeder Reactor (LMFBR) demonstration plant. Seven breeder-type reactors which were built during this time are described and their technological progress assessed. The Clinch River Breeder Reactor Project (CRBRP) has been designed to demonstrate that it can be licensed, can operate on a large power grid, and can provide industry with important experience. As the next logical step in LMFBR development, the project has suffered repeated cancellation efforts with only minor modifications to its schedule. Controversies have developed over the timing of a large-scale demonstration plant, the risks of proliferation, economics, and other problems. Among the innovative developments adopted for the CRBRP is a higher thermal efficiency potential, the type of development which Senator McCormack feels justifies continuing the project. He argues that the nuclear power program can and should be revitalized by continuing the CRBRP.

  14. Learning and nuclear safety: New reactors and US regulation

    International Nuclear Information System (INIS)

    Nichols, E.; Wildavsky, A.

    1992-01-01

    Gathering and analyzing data from operating reactors has become part of government and industry programs to improve performance in plants already on line and to inform development of future reactors. In the United States, however, early development and certain other factors combined to encourage a bias in learning. Regulation and learning from operational data intersect in ways that limit participation, data collection, and positive response to findings. Past learning has shown the advantage of simpler more standard designs with passive or inherent safety features. However, even designs incorporating these past lessons are apt to face tough regulatory tests and much criticism as operating experience is gathered. Only the operational success of new standardized reactors is apt to help rationalize regulation. (orig.)

  15. The working lifetime of nuclear power plants and new types of power reactors

    International Nuclear Information System (INIS)

    Bataille, Ch.; Birraux, C.

    2003-01-01

    The report on the working lifetime of nuclear power plants and new reactor types, by Mr Christian Bataille, deputy for the Nord, and Mr Claude Birraux, deputy for Haute-Savoie as well as President of the Office, supplements the studies carried out by the Parliamentary Office on the Safety of Nuclear Installations and Radioactive Wastes: it examines the remaining working life of the EDF nuclear power plants and the current status of projects that might, if circumstances were right, replace the reactors at present in service. The report investigates the different physical and other factors that influence the ageing of nuclear power plants and tackles the question of whether the design life of 40 years could be exceeded in practice. The whole issue of French nuclear power plant is put in perspective and compared with the situation of nuclear plants in Finland, Sweden, Germany and the United States, from the technical and regulatory standpoints. Believing that any attempt to optimise the working lifetime of the power plants currently in service must be accompanied by simultaneous moves aimed at their replacement, Messrs. Christian Bataille and Claude Birraux go on to review in detail the various light water reactor projects being proposed around the world for completion by 2015, as developments of existing models, in particular the EPR reactor of Framatome ANP, characterised by its competitiveness. They suggest that a first such reactor should be built as quickly as possible. Describing the other nuclear systems being investigated by research organisations not only in France but also in the United States and Sweden, Mrs. Christian Bataille and Claude Birraux review the objectives of these and the circumstances in which they might be developed, which would be unlikely to be before 2035 in view of the technological problems to be overcome and the industrial demonstration plants that would be needed

  16. Fuel handling system of nuclear reactor plants

    International Nuclear Information System (INIS)

    Faulstich, D.L.

    1991-01-01

    This patent describes a fuel handing system for nuclear reactor plants comprising a reactor vessel having an openable top and removable cover for refueling and containing therein, submerged in coolant water substantially filling the reactor vessel, a fuel core including a multiplicity of fuel bundles formed of groups of sealed tube elements enclosing fissionable fuel assembled into units. It comprises a fuel bundle handing platform moveable over the open top of the reactor vessel; a fuel bundle handing mast extendable downward from the platform with a lower end projecting into the open top reactor vessel to the fuel core submerged in water; a grapple head mounted on the lower end of the mast provided with grappling hook means for attaching to and transporting fuel bundles into and out from the fuel core; and a camera with a prismatic viewing head surrounded by a radioactive resisting quartz cylinder and enclosed within the grapple head which is provided with at least three windows with at least two windows provided with an angled surface for aiming the camera prismatic viewing head in different directions and thereby viewing the fuel bundles of the fuel core from different perspectives, and having a cable connecting the camera with a viewing monitor located above the reactor vessel for observing the fuel bundles of the fuel core and for enabling aiming of the camera prismatic viewing head through the windows by an operator

  17. NPR and ANSI Containment Study Using Passive Cooling Techniques

    International Nuclear Information System (INIS)

    Shin, J. J.; Iotti, R. C.; Wright, R. F.

    1993-01-01

    Passive containment cooling study of NPR (New Production Reactor) and ANSI (Advanced Neutron Source) following postulated loss of coolant accident with a coincident station blackout due to total loss of all alternating current power are studied analytically and experimentally. All the reactor and containment cooling under this condition would rely on the passive cooling system which removes reactor decay heat and provides emergency core and containment cooling. Containment passive emergency core and containment cooling. Containment passive cooling for this study takes place in the annulus between containment steel shell and concrete shield building by natural convection air flow and concrete shield building by natural convection air flow and thermal radiation. Various heat transfer coefficients inside annular air space were investigated by running the modified Contempt code Contempt-Npr. In order to verify proper heat transfer coefficient, temperature, heat flux and velocity profiles were measured inside annular air space of the test facility which is a 24 foot (7.3m) high, steam heated inner cylinder of three foot (.91m) diameter and five and halt foot (1.7m) diameter outer cylinder. Comparison of Contempt-Npr and WGOTHIC was done for reduced scale Npr. It is concluded that Npr and ANSI containments can be passively cooled with air alone without extended cooling surfaces or passive water spray

  18. Proposed chemical plant initiated accident scenarios in a sulphur-iodine cycle plant coupled to a pebble bed modular reactor

    International Nuclear Information System (INIS)

    Brown, N.R.; Revankar, S.T.; Seker, V.; Downar, Th.J.

    2010-01-01

    In the sulphur-iodine (S-I) cycle nuclear hydrogen generation scheme the chemical plant acts as the heat sink for the very high temperature nuclear reactor (VHTR). Thus, any accident which occurs in the chemical plant must feedback to the nuclear reactor. There are many different types of accidents which can occur in a chemical plant. These accidents include intra-reactor piping failure, inter-reactor piping failure, reaction chamber failure and heat exchanger failure. Since the chemical plant acts as the heat sink for the nuclear reactor, any of these accidents induce a loss-of-heat-sink accident in the nuclear reactor. In this paper, several chemical plant initiated accident scenarios are presented. The following accident scenarios are proposed: i) failure of the Bunsen chemical reactor; ii) product flow failure from either the H 2 SO 4 decomposition section or HI decomposition section; iii) reactant flow failure from either the H 2 SO 4 decomposition section or HI decomposition section; iv) rupture of a reaction chamber. Qualitative analysis of these accident scenarios indicates that each result in either partial or total loss of heat sink accidents for the nuclear reactor. These scenarios are reduced to two types: i) discharge rate limited accidents; ii) discontinuous reaction chamber accidents. A discharge rate limited rupture of the SO 3 decomposition section of the SI cycle is proposed and modelled. Since SO 3 decomposition occurs in the gaseous phase, critical flow out of the rupture is calculated assuming ideal gas behaviour. The accident scenario is modelled using a fully transient control volume model of the S-I cycle coupled to a THERMIX model of a 268 MW pebble bed modular reactor (PBMR-268) and a point kinetics model. The Bird, Stewart and Lightfoot source model for choked gas flows from a pressurised chamber was utilised as a discharge rate model. A discharge coefficient of 0.62 was assumed. Feedback due to the rupture is observed in the nuclear

  19. Evolutionary water cooled reactors: Strategic issues, technologies and economic viability. Proceedings of a symposium

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-12-01

    Symposium on evolutionary water cooled reactors: Strategic issues, technologies and economic viability was intended for managers in utilities, reactor design organizations and hardware manufacturing companies and for government decision makers who need to understand technological advances and the potential of evolutionary water cooled reactors to contribute to near and medium term energy demands. The topics addressed include: strategic issues (global energy outlook, the role of nuclear power in sustainable energy strategies, power generation costs, financing of nuclear plant projects, socio-political factors and nuclear safety requirements); technological advances (instrumentation and control, means od improving prevention and mitigation of severe accidents, development of passive safety systems); keys to economic viability (simplification, standardization, advances in construction and project management, feedback of experience from utilities into new designs, and effective management of plant operation)

  20. Plant-scale anodic dissolution of unirradiated N-Reactor fuel

    International Nuclear Information System (INIS)

    Gay, E.C.; Miller, W.E.; Laidler, J.J.

    1995-01-01

    Anodic dissolution tests were made with unirradiated N-Reactor fuel to determine the fuel segment length, diameter, and shape required for high throughput electro-refiner treatment for ultimate disposal in a geologic repository. Based on these tests, a conceptual design was produced of an electro-refiner for a full-scale plant to process N-Reactor spent fuel. In this design, the diameter of an electrode assembly is about 0.6 m (25 in.). Eight of these assemblies in an electro-refiner would accommodate a 1.333-metric-ton batch of N-Reactor fuel. Electrorefining would proceed at a rate of 40 kg uranium per hour. (author)

  1. Advanced Concepts for Pressure-Channel Reactors: Modularity, Performance and Safety

    Science.gov (United States)

    Duffey, Romney B.; Pioro, Igor L.; Kuran, Sermet

    Based on an analysis of the development of advanced concepts for pressure-tube reactor technology, we adapt and adopt the pressure-tube reactor advantage of modularity, so that the subdivided core has the potential for optimization of the core, safety, fuel cycle and thermal performance independently, while retaining passive safety features. In addition, by adopting supercritical water-cooling, the logical developments from existing supercritical turbine technology and “steam” systems can be utilized. Supercritical and ultra-supercritical boilers and turbines have been operating for some time in coal-fired power plants. Using coolant outlet temperatures of about 625°C achieves operating plant thermal efficiencies in the order of 45-48%, using a direct turbine cycle. In addition, by using reheat channels, the plant has the potential to produce low-cost process heat, in amounts that are customer and market dependent. The use of reheat systems further increases the overall thermal efficiency to 55% and beyond. With the flexibility of a range of plant sizes suitable for both small (400 MWe) and large (1400 MWe) electric grids, and the ability for co-generation of electric power, process heat, and hydrogen, the concept is competitive. The choice of core power, reheat channel number and exit temperature are all set by customer and materials requirements. The pressure channel is a key technology that is needed to make use of supercritical water (SCW) in CANDU®1 reactors feasible. By optimizing the fuel bundle and fuel channel, convection and conduction assure heat removal using passive-moderator cooling. Potential for severe core damage can be almost eliminated, even without the necessity of activating the emergency-cooling systems. The small size of containment structure lends itself to a small footprint, impacts economics and building techniques. Design features related to Canadian concepts are discussed in this paper. The main conclusion is that development of

  2. Development status of PIUS/ISER - a inherently safe reactor for the international use

    International Nuclear Information System (INIS)

    Wakabayashi, Hiroaki

    1987-01-01

    It is just in early 1980s that LWR-based nuclear power has become a substantial power source. Though the safety level of nuclear power is always claimed to be sufficiently high by the industry, it rests on the idea of defense in depth, the calculation by probabilistic risk assessment (PRA) or probabilistic safety assessment (PSA). The TMI-2 and Chernobyl-4 accidents occurred in the industrially most advanced countries. In this paper, an alternative way to safe nuclear power is sought in so-called inherently safe reactors (ISR) including the LWR type PIUS/ISER. With proper consideration into the design of nuclear reactor plants, those can be made basically safe through the use of passive safe mechanism for their design. In short, an ISR is a nuclear power reactor which has passive and intrinsic core cooling capability and automatic shutdown capability. As the nuclear power reactors which are currently claimed to be inherently safe, there are the process inherent and ultimately safe reactor (PIUS) of ASEA-ATOM Sweden and the inherently safe and economical reactor (ISER) of the University of Tokyo, Japan, of LWR type. The current status of the development, the reliability, and some technical problems of ISER/PIUS and the attitude of various countries toward ISER/PIUS are described. (Kako, I.)

  3. Reactor power control device in BWR power plant

    International Nuclear Information System (INIS)

    Kurosawa, Tsuneo.

    1997-01-01

    The present invention provides a device for controlling reactor power based on a start-up/shut down program in a BWR type reactor, as well as for detecting deviation, if occurs, of the power from the start-up/shut down program, to control a recycling flow rate control system or control rod drive mechanisms. Namely, a power instruction section successively executes the start-up/shut down program and controls the coolant recycling system and the control rod driving mechanisms to control the power. A current state monitoring and calculation section receives a process amount, calculates parameters showing the plant state, compares/monitors them with predetermined values, detecting the deviation, if occurs, of the plant state from the start-up/shut down program, and prevents output of a power increase control signal which leads to power increase. A forecasting and monitoring/calculation section forecasts and calculates the plant state when not yet executed steps of the start-up/shut down program are performed, stops the execution of the start-up/shut down program in the next step in a case of forecasting that the results of the calculation will deviate from the start-up/shut down program. (I.S.)

  4. MINIMARS tandem mirror reactor study

    International Nuclear Information System (INIS)

    Perkins, L.J.; Logan, B.G.; Doggett, J.N.

    1986-01-01

    During 1985-1986, Lawrence Livermore National Lab., in partnership with the Fusion Engineering Design Center of Oak Ridge National Lab., the Univ. of Wisconsin, TRW, Grumman Aerospace Corporation, General Dynamics/Convair, Argonne National Lab., and the Canadian Fusion Fuels Technology Project, has conducted the conceptual design of MINIMARS, a small commercial tandem mirror reactor with novel octopole end plugs. With a net electric output of 600 MW(e), MINIMARS is expressly designed for short (∼4- to 5-yr) construction time, factory-built modules, and a passively safe blanket and thermal cycle. In this way, we intend to achieve a small reactor based on the tandem mirror principle that will minimize utility financial risk, thereby providing an attractive alternative to the more conventional large fusion plant designs encountered to date

  5. Design and development of small and medium integral reactor core

    International Nuclear Information System (INIS)

    Zee, Sung Quun; Chang, M. H.; Lee, C. C.; Song, J. S.; Cho, B. O.; Kim, K. Y.; Kim, S. J.; Park, S. Y.; Lee, K. B.; Lee, C. H.; Chun, T. H.; Oh, D. S.; In, W. K.; Kim, H. K.; Lee, C. B.; Kang, H. S.; Song, K. N.

    1997-07-01

    Recently, the role of small and medium size integral reactors is remarkable in the heat applications rather than the electrical generations. Such a range of possible applications requires extensive used of inherent safety features and passive safety systems. It also requires ultra-longer cycle operations for better plant economy. Innovative and evolutionary designs such as boron-free operations and related reactor control methods that are necessary for simple reactor system design are demanded for the small and medium reactor (SMR) design, which are harder for engineers to implement in the current large size nuclear power plants. The goals of this study are to establish preliminary design criteria, to perform the preliminary conceptual design and to develop core specific technology for the core design and analysis for System-integrated Modular Advanced ReacTor (SMART) of 330 MWt power. Based on the design criteria of the commercial PWR's, preliminary design criteria will be set up. Preliminary core design concept is going to be developed for the ultra-longer cycle and boron-free operation and core analysis code system is constructed for SMART. (author). 100 refs., 40 tabs., 92 figs

  6. Preliminary Study of Applying Phase Change Materials (PCM) for Containment Passive Cooling

    International Nuclear Information System (INIS)

    Ko, A Reum; Lee, Jeong Ik; Yoon, Ho Joon

    2016-01-01

    Most of Pressurized Water Reactor (PWR) containments use fan cooler systems and containment spray systems. However, the importance of passive safety system has increased after the Fukushima accident. As the main passive safety system, Passive Containment Cooling System (PCCS), which utilizes natural phenomena to remove the heat released from the reactor, is suggested in the advanced pressurized water reactor (APWR). To increase the efficiency of passive cooling, additional passive containment cooling method using Phase Change Material (PCM) is suggested in this paper. For containment using PCMs, there are many advantages. Phase Change Material (PCM) is proposed as an additional passive containment cooling method to increase the efficiency of passive cooling in this paper. To apply proper PCMs to containment, commercially available PCMs were screened while reviewing thermophysical properties data and suggested selection criteria. A sensitivity study was also carried out to identify the effect of potential installation location of PCM using the CAP code. The pressure of containment in most cases showed slightly higher than that of the initial case. For the temperature of steam and water and humidity, similar results with the initial case were showed in most cases

  7. Preliminary Study of Applying Phase Change Materials (PCM) for Containment Passive Cooling

    Energy Technology Data Exchange (ETDEWEB)

    Ko, A Reum; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of); Yoon, Ho Joon [KUSTAR, Abu Dhabi (United Arab Emirates)

    2016-05-15

    Most of Pressurized Water Reactor (PWR) containments use fan cooler systems and containment spray systems. However, the importance of passive safety system has increased after the Fukushima accident. As the main passive safety system, Passive Containment Cooling System (PCCS), which utilizes natural phenomena to remove the heat released from the reactor, is suggested in the advanced pressurized water reactor (APWR). To increase the efficiency of passive cooling, additional passive containment cooling method using Phase Change Material (PCM) is suggested in this paper. For containment using PCMs, there are many advantages. Phase Change Material (PCM) is proposed as an additional passive containment cooling method to increase the efficiency of passive cooling in this paper. To apply proper PCMs to containment, commercially available PCMs were screened while reviewing thermophysical properties data and suggested selection criteria. A sensitivity study was also carried out to identify the effect of potential installation location of PCM using the CAP code. The pressure of containment in most cases showed slightly higher than that of the initial case. For the temperature of steam and water and humidity, similar results with the initial case were showed in most cases.

  8. Predicting plant attractiveness to pollinators with passive crowdsourcing.

    Science.gov (United States)

    Bahlai, Christie A; Landis, Douglas A

    2016-06-01

    observations were not associated with Internet images, but were slightly associated with BP. Our results suggest that passively crowdsourced image data can potentially be a useful screening tool to identify candidate plants for pollinator habitat restoration efforts directed at wild bee conservation. Increasing our understanding of the attractiveness of a greater diversity of plants increases the potential for more rapid and efficient research in creating pollinator-supportive landscapes.

  9. The challenge of introducing high-temperature reactor plants onto the international power plant market

    International Nuclear Information System (INIS)

    Bogen, J.; Stoelzl, D.

    1987-01-01

    Growth of world population increases energy demand until the year 2000 and afterwards. Electricity growth rates in industrialized nations are lower after the oil price escalation in 1973 and 1979, and in developing countries grid sizes are often too small for the operation of large LWR plants. This indicates a potential for small and medium-sized power reactors such as the HTR-100 and the HTR-500. These plants can compete with coal fired plants of comparable size. An HTR-500 is even competitive, considering the electricity generating cost of large LWR plants. The special advantages of HTR plants in the small and medium-capacity range are discussed. (orig.)

  10. The challenge of introducing high-temperature-reactor plants onto the international power plant market

    International Nuclear Information System (INIS)

    Bogen, J.; Stoelzl, D.

    1988-01-01

    Growth of world population increases energy demand until the year 2000 and afterwards. Electricity growth rates in industrialized nations are lower after the oil price escalation in 1973 and 1979, and in developing countries grid sizes are often too small for the operation of large LWR plants. This indicates a potential for small and medium-sized power reactors such as the HTR-100 and the HTR-500. These plants can compete with coal fired plants of comparable size. An HTR-500 is even competitive, considering the electricity generating cost of large LWR plants. The special advantages of HTR plants in the small and medium-capacity range are discussed. (orig.)

  11. High Efficiency Nuclear Power Plants Using Liquid Fluoride Thorium Reactor Technology

    Science.gov (United States)

    Juhasz, Albert J.; Rarick, Richard A.; Rangarajan, Rajmohan

    2009-01-01

    An overall system analysis approach is used to propose potential conceptual designs of advanced terrestrial nuclear power plants based on Oak Ridge National Laboratory (ORNL) Molten Salt Reactor (MSR) experience and utilizing Closed Cycle Gas Turbine (CCGT) thermal-to-electric energy conversion technology. In particular conceptual designs for an advanced 1 GWe power plant with turbine reheat and compressor intercooling at a 950 K turbine inlet temperature (TIT), as well as near term 100 MWe demonstration plants with TITs of 950 and 1200 K are presented. Power plant performance data were obtained for TITs ranging from 650 to 1300 K by use of a Closed Brayton Cycle (CBC) systems code which considered the interaction between major sub-systems, including the Liquid Fluoride Thorium Reactor (LFTR), heat source and heat sink heat exchangers, turbo-generator machinery, and an electric power generation and transmission system. Optional off-shore submarine installation of the power plant is a major consideration.

  12. Main Steam Line Break Analysis for the Fully Passive Safety System of SMART

    International Nuclear Information System (INIS)

    Kim, Seong Wook; Chun, Ji Han; Bae, Kyoo Hwan; Kim, Keung Koo

    2013-01-01

    The standard design approval of SMART (System-integrated Modular Advanced ReacTor) developed by KAERI and KEPCO consortium was issued on July 4, 2012. Although SMART has enhanced safety compared to the conventional reactor, there is a demand to meet the 'passive safety performance requirements' after the Fukushima accident. The passive safety performance requirements are the capabilities to maintain the plant at a safe shutdown condition for a minimum of 72 hours without AC power supply or operator action in case of design basis accident (DBA). To satisfy the requirements, KAERI is developing a safety enhanced SMART by adopting a passive safety injection system. The passive safety injection system developed for SMART is a gravity-driven injection system, which consists of four trains, each of which includes a pressure balance line, core makeup tank (CMT), safety injection tank (SIT) and injection line. The CMT plays an important role to inject borated water into the RCS to prevent or dissolve the return to power (re-criticality) condition during the event of increase in heat removal by the secondary system. The main steam line break accident (MSLB) is the most limiting accident for an increase in heat removal by the secondary system. In this study, the safety analysis results of MSLBs at hot full power condition and at hot zero power condition in view of re-criticality are given. The MSLB accident has been analyzed for the SMART adopting fully passive safety system in the aspect of re-criticality. The results show that the core remains subcritical condition throughout the transient due to the borated water injected by the CMT. As further works, many kinds of analyses and sensitivity studies should be performed for the design establishment and improvement of the fully passive system of SMART

  13. Capital cost: pressurized water reactor plant. Commerical electric power cost studies

    International Nuclear Information System (INIS)

    1977-06-01

    The investment cost study for the 1139-MW(e) pressurized water reactor (PWR) central station power plant consists of two volumes. This volume includes in addition to the foreword and summary, the plant description and the detailed cost estimate

  14. A fast track approach to commercializing the sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Boardman, C.E.; Hui, Marvin; Carroll, Douglas

    1999-01-01

    As a result of more than 50 years of Liquid Metal Reactor design and development work the basic technology is well understood. However, commercialization of the Fast Breeder Reactor (FBR) has been delayed while various approaches to achieving competitive plant and fuel cycle costs are explored, developed, and demonstrated in prototype systems. Most designers have elected to take advantage of the economy of scale but are burdened by the cost and risk associated with the need for incremental scale up through the design, construction, and operation of multiple demonstration plants. An alternative commercialization path developed by GE would utilize a modular plant design to reduce the plant construction, R and D, and economic risk associated with the need to build multiple demonstration plants to reach a competitive size'. The key question is can a modular FBR compete with alternative electrical generation systems? Recently completed studies indicate that the answer to this question is yes if the modular plant designers keep the design simple by incorporating passive safety features and optimizing the manner in which supporting service systems are shared. (author)

  15. Simplified nuclear plant design for tomorrow's energy needs

    International Nuclear Information System (INIS)

    Slember, R.

    1989-09-01

    Commercial nuclear powered plants play an important role in the strategic energy plans of many countries throughout the world. Many energy planners agree that nuclear plants will have to supply an increasing amount of electrical energy in the 1990s and beyond. Just as other major industries are continually taking steps to update and improve existing products, the United States' nuclear industry has embarked on a program to simplify plant systems, shorten construction time and improve economics for new plant models. One of the models being developed by Westinghouse Electric Corporation and Burns and Roe Company is the Advanced Passive 600 MWe design which incorporates safety features that passively protect the reactor during assumed abnormal operating events. These passive safety systems utilize natural circulation/cooling for mitigating abnormal events and simplify plant design and operation. This type of system eliminates the need for costly active safety grade components, results in a reduction of ancillary equipment and assists in shortening construction time. The use of passive safety systems also permits design simplification of the auxiliary systems effectively reducing operating and maintenance requirements. Collectively, the AP600 design features result in a safe plant that addresses and alleviates the critical industry issues that developed in the 1980s. Further, the design addresses utility and regulatory requirements for safety, reliability, maintainability, operations and economics. Program results to date give confidence that the objectives of the Advanced Passive 600 design are achievable through overall plant simplification. The report will include timely results from the work being performed on the salient technical features of the design, plant construction and operation. Other required institutional changes, such as the prerequisite for a design which is complete and licensed prior to start of construction, will also be presented

  16. Systematic simulation of a tubular recycle reactor on the basis of pilot plant experiments

    Energy Technology Data Exchange (ETDEWEB)

    Paar, H; Narodoslawsky, M; Moser, A [Technische Univ., Graz (Austria). Inst. fuer Biotechnologie, Mikrobiologie und Abfalltechnologie

    1990-10-10

    Systematic simulatiom may decisively help in development and optimization of bioprocesses. By applying simulation techniques, optimal use can be made of experimental data, decreasing development costs and increasing the accuracy in predicting the behavior of an industrial scale plant. The procedure of the dialogue between simulation and experimental efforts will be exemplified in a case study. Alcoholic fermentation of glucose by zymomonas mobilis bacteria in a gasified turbular recycle reactor was studied first by systematic simulation, using a computer model based solely on literature data. On the base of the results of this simulation, a 0.013 m{sup 3} pilot plant reactor was constructed. The pilot plant experiments, too, were based on the results of the systematic simulation. Simulated and experimental data were well in agreement. The pilot plant experiments reiterated the trends and limits of the process as shown by the simulation results. Data from the pilot plant runs were then used to improve the simulation model. This improved model was subsequently used to simulate the performances of an industrial scale plant. The results of this simulation are presented. They show that the alcohol fermentation in a tubular recycle reactor is potentially advantageous to other reactor configurations, especially to continuous stirred tanks. (orig.).

  17. Uranium-236 in light water reactor spent fuel recycled to an enriching plant

    International Nuclear Information System (INIS)

    de la Garza, A.

    1977-01-01

    The introduction of 236 U to an enriching plant by recycling spent fuel uranium results in enriched products containing 236 U, a parasitic neutron absorber in reactor fuel. Convenient approximate methodology determines 235 236 U, and total uranium flowsheets with associated separative work requirements in enriching plant operations for use by investigators of the light water reactor fuel cycle not having recourse to specialized multicomponent cascade technology. Application of the methodology has been made to compensation of an enriching plant product for 236 U content and to the value at an enriching plant of spent fuel uranium. The approximate methodology was also confirmed with more exact calculations and with some experience with 236 U in an enriching plant

  18. ALWR safety approaches and trends. Implementation of passive safety features in the design

    Energy Technology Data Exchange (ETDEWEB)

    Ignatiev, V

    1995-11-01

    Reactor vendors world-wide are examining various advanced light water reactors (ALWR) options to reach utility goals. The amount of information available about each design varies essentially depending on its maturity. Some advanced reactor designs are the evolutionary results of combining old structures, systems and components in new ways, others use innovative solutions. A summary review is given for better understanding of new ALWR design trends and approaches in different countries and subsequent R and D activities. An attempt was made to describe and assess specific innovative and passive features implemented in the leading ALWR designs for further plant design safety improvements. The advantages and disadvantages of these innovations in obtaining reliable systems have been considered. Also, this report indicates the importance of uncertainties remaining and identifies the additional work needed. 51 refs, 27 figs, 7 tabs.

  19. ALWR safety approaches and trends. Implementation of passive safety features in the design

    International Nuclear Information System (INIS)

    Ignatiev, V.

    1995-11-01

    Reactor vendors world-wide are examining various advanced light water reactors (ALWR) options to reach utility goals. The amount of information available about each design varies essentially depending on its maturity. Some advanced reactor designs are the evolutionary results of combining old structures, systems and components in new ways, others use innovative solutions. A summary review is given for better understanding of new ALWR design trends and approaches in different countries and subsequent R and D activities. An attempt was made to describe and assess specific innovative and passive features implemented in the leading ALWR designs for further plant design safety improvements. The advantages and disadvantages of these innovations in obtaining reliable systems have been considered. Also, this report indicates the importance of uncertainties remaining and identifies the additional work needed. 51 refs, 27 figs, 7 tabs

  20. Basic safety principles of KLT-40C reactor plants

    International Nuclear Information System (INIS)

    Beliaev, V.; Polunichev, V.

    2000-01-01

    The KLT-40 NSSS has been developed for a floating power block of a nuclear heat and power station on the basis of ice-breaker-type NSSS (Nuclear Steam Supply System) with application of shipbuilding technologies. Basic reactor plant components are pressurised water reactor, once-through coil-type steam generator, primary coolant pump, emergency protection rod drive mechanisms of compensate group-electromechanical type. Basic RP components are incorporated in a compact steam generating block which is arranged within metal-water shielding tank's caissons. Domestic regulatory documents on safety were used for the NSSS design. IAEA recommendations were also taken into account. Implementation of basic safety principles adopted presently for nuclear power allowed application of the KLT-40C plant for a floating power unit of a nuclear co-generation station. (author)

  1. Study on reactor vessel replacement (RVR) for 1100 MW class BWR plants in Japan

    International Nuclear Information System (INIS)

    Mizutani, J.; Kawamura, S.; Aoki, M.; Mori, T.

    2001-01-01

    Plant Life Management (PLM) is being studied in Japan, and reactor vessel replacement (RVR) is being considered as one option. Since reactor internals, except for reusable parts, and the reactor pressure vessel (RPV) are replaced, the RVR provides an effective technology for extending the service life of nuclear power plants substantially. At ICONE 7, we reported on the technical viability of the RVR for BWR4-type 800 MWe class plants. This time, we rationalized the RVR method through a study for BWR5-type 1100 MWe class plants to reduce the RVR duration and evaluated the technical viability and the economic efficiency of the method. In addition, we discuss how to dispose of the RPV to complete a scenario of the process from the RVR to its final disposal. (author)

  2. ELMO Bumpy Torus Reactor and power plant: conceptual design study

    International Nuclear Information System (INIS)

    Bathke, C.G.; Dudziak, D.J.; Krakowski, R.A.

    1981-08-01

    A complete power plant design of a 1200-MWe ELMO Bumpy Torus Reactor (EBTR) is presented. An emphasis is placed on those features that are unique to the EBT confinement concept, with subsystems and balance-of-plant items that are more generic to magnetic fusion being adapted from past, more extensive tokamak reactor designs. Similar to the latter tokamak studies, this conceptual EBTR design also emphasizes the use of conventional or near state-of-the-art engineering technology and materials. An emphasis is also placed on system accessibility, reliability, and maintainability, as these crucial and desirable characteristics relate to the unique high-aspect-ratio configuration of EBTs. Equal and strong emphasis is given to physics, engineering/technology, and costing/economics components of this design effort. Parametric optimizations and sensitivity studies, using cost-of-electricity as an object function, are reported. Based on these results, the direction for future improvement on an already attractive reactor design is identified

  3. Aging management strategy for reactor internals of Korean nuclear power plants

    International Nuclear Information System (INIS)

    Hwang, Seong Sik; Kim, Soung Woo; Lee, Sam Lai; Hong, Seung Mo; Kim, Hong Pyo; Kim, Dong Jin; Lim, Yun Soo; Kim, Joung Soo; Jung, Man Kyo; Park, Jang Yul

    2010-01-01

    This report describes various factors on the IASCC of reactor internals in terms of fluence, stress, water chemistries and materials. Materials of each components of Korean nuclear power plants have been surveyed. A technical report for a management of reactor internals issued by EPRI was reviewed for a selection of most susceptible area among many components. Baffle former bolts are considered to be the most susceptible parts due to high irradiation level(fluence) and high tensile stress. Neutron fluence of Kori-1 and Kori-2 was calculated based on fuel exchange history, fuel performance and plant operation history. This report will be used for more advanced inspection and maintenance guidelines development by supplying inspection intervals and components (most susceptible regions) for the long term operation plants

  4. Trends in advanced reactor development and the role of the IAEA

    International Nuclear Information System (INIS)

    Semenov, B.; Dastidar, P.; Kupitz, J.; Cleveland, J.; Goodjohn, A.

    1992-01-01

    This report discusses advanced reactors are being developed for all principal reactor types, i.e. the light and heavy water-cooled reactors, the liquid-metal-cooled reactors and the gas-cooled reactors. Some of these developments are primarily of an evolutionary nature, i.e. they represent improvements in component and system technology, and in construction and operating practices as a result of experience gained with presently operating plants. Other developments are also evolutionary but with some incorporation of innovative features such as providing passive systems for assuring continuous cooling for removal of decay heat from the reactor core. If there is a revival of nuclear power, which may be dictated by ecological and economical factors, advanced reactors now being developed could help to meet the large demand for new plants in developed and developing countries, not only for electricity generation, but also for district heating, desalination and for process heat. The IAEA, as the only global international governmental organization dealing with nuclear power, has promoted international information exchange and international cooperation between all countries with their own advanced nuclear power programmes and has offered assistance to countries with an interest in exploratory or research programmes. In the future the IAEA could play an even more-important role

  5. Multifrequency tests in the EBR-II reactor plant

    International Nuclear Information System (INIS)

    Feldman, E.E.; Mohr, D.; Gross, K.C.

    1989-01-01

    A series of eight multifrequency tests was conducted on the Experimental Breeder Reactor II. In half of the tests a control rod was oscillated and in the other half the controller input voltage to the intermediate-loop-sodium pump was perturbed. In each test the input disturbance consisted of several superimposed single-frequency sinusoidal harmonics of the same fundamental. The tests are described along with the theoretical and practical aspects of their development and design. Samples of measured frequency responses are also provided for both the reactor and the power plant. 22 refs., 5 figs., 2 tabs

  6. Concept of safety related I and C and power supply systems in the passive safety concept of the HTR-module

    International Nuclear Information System (INIS)

    Juengst, U.

    1990-01-01

    The main motivation for the passive safety concepts is to gain a better quality of safety or at least to achieve higher public acceptance for nuclear power plants. This strategy has been introduced into the European Fast Reactor (EER), a common project of France, UK and Germany is applied stringently to the German high-temperature gas-cooled reactor ''HTR - Module''. The following fields are briefly described in the paper: Safety design features of the HTR - Module, overview of I and C concept, reactor protection system, emergency control room, power supply concept, system arrangement and protection against external hazards, accidents sequence of station black-out. (author). 3 figs

  7. PRISM reactor system design and analysis of postulated unscrammed events

    International Nuclear Information System (INIS)

    Van Tuyle, G.J.; Slovik, G.C.

    1991-01-01

    Key safety characteristics of the PRISM reactor system include the passive reactor shutdown characteristic and the passive shutdown heat removal system, RVACS. While these characteristics are simple in principle, the physical processes are fairly complex, particularly for the passive reactor shutdown. It has been possible to adapt independent safety analysis codes originally developed for the Clinch River Breeder Reactor review, although some limitations remain. In this paper, the analyses of postulated unscrammed events are discussed, along with limitations in the predictive capabilities and plans to correct the limitations in the near future. (author)

  8. Penn State advanced light water reactor concept

    International Nuclear Information System (INIS)

    Borkowski, J.A.; Smith, K.A.; Edwards, R.M.; Robinson, G.E.; Schultz, M.A.; Klevans, E.H.

    1987-01-01

    The accident at Three Mile Island heightened concerns over the safety of nuclear power. In response to these concerns, a research group at the Pennsylvania State University (Penn State) undertook the conceptual design of an advanced light water reactor (ALWR) under sponsorship of the US Dept. of Energy (DOE). The design builds on the literally hundreds of years worth of experience with light water reactor technology. The concept is a reconfigured pressurized water reactor (PWR) with the capability of being shut down to a safe condition simply by removing all ac power, both off-site and on-site. Using additional passively activated heat sinks and replacing the pressurizer with a pressurizing pump system, the concept essentially eliminates the concerns of core damage associated with a total station blackout. Evaluation of the Penn State ALWR concept has been conducted using the EPRI Modular Modeling System (MMS). Results show that a superior response to normal operating transients can be achieved in comparison to the response with a conventional PWR pressurizer. The DOE-sponsored Penn State ALWR concept has evolved into a significant reconfiguration of a PWR leading to enhanced safety characteristics. The reconfiguration has touched a number of areas in overall plant design including a shutdown turbine in the secondary system, additional passively activated heat sinks, a unique primary side pressurizing concept, a low pressure cleanup system, reactor building layout, and a low power density core design

  9. Italy: Analysis of Solutions for Passively Actuated Safety Shutdown Devices

    International Nuclear Information System (INIS)

    Burgazzi, L.

    2015-01-01

    This article looks at different special shutdown systems specifically engineered for prevention of severe accidents, to be implemented on Fast Reactors, with main focus on the investigation of the performance of the self-actuated shutdown systems in Sodium Fast Reactors. The passive shut-down systems are designed to shut-down system only by inherent passive reactivity feedback mechanism, under unprotected accident conditions, implying failure of reactor protection system. They are conceived to be self-actuated without any signal elaboration, since the actuation of the system is triggered by the effects induced by the transient like material dilatation, in case of overheating of the coolant for instance, according to Fast Reactor design to meet the safety requirements

  10. Nuclear power plants without the risk of a reactor disaster

    International Nuclear Information System (INIS)

    Merz, L.

    1980-01-01

    Nowadays everybody believes that all nuclear power plants are an inevitable risk for the life of the people in its environment; this is a prejudice. This article points out that it is possible to plan nuclear power plants in a way that no deaths in the population have to be feared, even in the heaviest possible disturbances. In order to realise such nuclear power plants, it is necessary either to improve the cooling of the fuel rods by passive additional safety measures of which one unlimitedly efficient thus excluding a nuclear melting accident or to improve the safety containers so that it remains sufficiently tight. (orig./HP) [de

  11. The use of ferritic materials in light water reactor power plants

    International Nuclear Information System (INIS)

    Marston, T.V.

    1984-01-01

    This paper reviews the use of ferritic materials in LWR power plant components. The two principal types of LWR systems, the boiling water reactor (BWR) and the pressurized water reactor (PWR) are described. The evolution of the construction materials, including plates and forgings, is presented. The fabrication process for both reactors constructed with plates and forgings are described in detail. Typical mechanical properties of the reactor vessel materials are presented. Finally, one critical issue radiation embrittlement dealing with ferritic materials is discussed. This has been one of the major issues regarding the use of ferritic material in the construction of LWR pressure vessels

  12. Structural design aspects of innovative designs under development in the current US Liquid Metal-Cooled Reactor program

    International Nuclear Information System (INIS)

    Seidensticker, R.W.

    1986-01-01

    The US Liquid Metal-Cooled Reactor (LMR) program has been restructured and is now focussed on the development of innovative plant designs which emphasize shorter construction times, increased use of passive, inherently safe features, cost-competitiveness with LWR plants, and minimization of safety-related systems. These changes have a considerable effect on the structural design aspects of the LMR plant. These structural problems and their solutions now under study form the main focus of this paper. (orig.)

  13. Plant dynamics analyses of fast reactor concept: RAPID-A without any control rod

    International Nuclear Information System (INIS)

    Kambe, Mitsuru

    1996-01-01

    Plant dynamics analyses of a fast reactor concept RAPID-A without any control rod have been demonstrated in case of reactor startup and sudden change of the primary flow rate. RAIP-A concept involves Lithium Expansion Module (LEM) for inherent reactivity feedback, Lithium Injection Module (LIM) for inherent ultimate shutdown and Lithium Release Module (LRM) for automated reactor startup. LEM consists of Quick-LEM and Slow-LEM. Slow-LEM provides with moderate reactivity addition as decreasing temperature. Quick-LEM assures quick negative reactivity feedback as increasing temperature. Plant dynamics analyses revealed that reactor power is nearly proportional to the primary flow rate even if the flow rate increases suddenly. Fully automated reactor startup from the subcritical condition has been attempted by inserting reactivity at a constant rate by LRM. Allowable rate of reactivity addition has been obtained in respect to Quick-LEM reactivity worth. (author)

  14. Requirements for advanced simulation of nuclear reactor and chemicalseparation plants.

    Energy Technology Data Exchange (ETDEWEB)

    Palmiotti, G.; Cahalan, J.; Pfeiffer, P.; Sofu, T.; Taiwo, T.; Wei,T.; Yacout, A.; Yang, W.; Siegel, A.; Insepov, Z.; Anitescu, M.; Hovland,P.; Pereira, C.; Regalbuto, M.; Copple, J.; Willamson, M.

    2006-12-11

    This report presents requirements for advanced simulation of nuclear reactor and chemical processing plants that are of interest to the Global Nuclear Energy Partnership (GNEP) initiative. Justification for advanced simulation and some examples of grand challenges that will benefit from it are provided. An integrated software tool that has its main components, whenever possible based on first principles, is proposed as possible future approach for dealing with the complex problems linked to the simulation of nuclear reactor and chemical processing plants. The main benefits that are associated with a better integrated simulation have been identified as: a reduction of design margins, a decrease of the number of experiments in support of the design process, a shortening of the developmental design cycle, and a better understanding of the physical phenomena and the related underlying fundamental processes. For each component of the proposed integrated software tool, background information, functional requirements, current tools and approach, and proposed future approaches have been provided. Whenever possible, current uncertainties have been quoted and existing limitations have been presented. Desired target accuracies with associated benefits to the different aspects of the nuclear reactor and chemical processing plants were also given. In many cases the possible gains associated with a better simulation have been identified, quantified, and translated into economical benefits.

  15. Cost analysis of light water reactor power plants

    International Nuclear Information System (INIS)

    Mooz, W.E.

    1978-06-01

    A statistical analysis is presented of the capital costs of light water reactor (LWR) electrical power plants. The objective is twofold: to determine what factors are statistically related to capital costs and to produce a methodology for estimating these costs. The analysis in the study is based on the time and cost data that are available on U.S. nuclear power plants. Out of a total of about 60 operating plants, useful capital-cost data were available on only 39 plants. In addition, construction-time data were available on about 65 plants, and data on completed construction permit applications were available for about 132 plants. The cost data were first systematically adjusted to constant dollars. Then multivariate regression analyses were performed by using independent variables consisting of various physical and locational characteristics of the plants. The dependent variables analyzed were the time required to obtain a construction permit, the construction time, and the capital cost

  16. Unprotected Accident Analyses of the 1200MWe GEN-IV Sodium-Cooled Fast Reactor Using the SSC-K Code

    International Nuclear Information System (INIS)

    Kwon, Young Min; Lee, Kwi Lim; Ha, Kwi Seok; Jeong, Hae Yong; Chang, Won Pyo; Seok, Su Dong; Lee, Yong Bum

    2010-02-01

    A conceptual design of an advanced breakeven sodium-cooled fast reactor (G4SFR) has recently been developed by KAERI under the national nuclear R and D plan. The G4SFR is a 1,200MWe metal-fueled pool-type sodium-cooled fast reactor adopting advanced safety design features. The G4SFR development plan focuses on particular technology development efforts to effectively meet the goals of the Generation-IV (GEN-IV) nuclear system such as efficient utilization of resources, economic competitiveness, a high standard of safety, and enhanced proliferation resistance. To enhance the safety of G4SFR, advanced design features of metal-fueled core, simple and large sodium-inventory primary heat transport system, and passive safety decay heat removal system are included in the reactor design. To evaluate potential safety characteristics of such advanced design features, the plant responses and safety margins were investigated using the system transient code SSC-K for three unprotected accidents of UTOP, ULOF, and ULOHS. It was shown that the G4SFR design has inherent and passive safety characteristics and is accommodating the selected ATWS events. The inherent safety mechanism of the reactor design makes the core shutdown with sufficient margin and passive removal of decay heat with matching the core power to heat sink by passive self-regulation. The self-regulation of power without scram is mainly due to the inherent negative reactivity feedback in conjunction with the large thermal inertia of the primary heat transport system and the passive decay heat removal. Such favorable inherent and passive safety behaviors of G4SFR are expected to virtually exclude the probability of severe accidents with potential for core damage

  17. Materialistic Aspects of Raising Resource of Pressurized Water Reactors for Low-Power Nuclear Plants

    International Nuclear Information System (INIS)

    Parshin, A.M.; Muratov, O.E.

    2005-01-01

    The opportunity of using ships reactors for low-power nuclear plants is considered. Some aspects of working constructional materials on cases of water-water reactors of ships nuclear units are considered. Advantages of raising resource of ships reactors are shown

  18. CFD simulations of moderator flow inside Calandria of the Passive Moderator Cooling System of an advanced reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pal, Eshita [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400 094 (India); Kumar, Mukesh [Reactor Engineering Division, Bhabha Atomic Research Center, Trombay, Mumbai 400 085 (India); Joshi, Jyeshtharaj B., E-mail: jbjoshi@gmail.com [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400 094 (India); Department of Chemical Engineering, Institute of Chemical Technology, Matunga, Mumbai 400019 India (India); Nayak, Arun K. [Reactor Engineering Division, Bhabha Atomic Research Center, Trombay, Mumbai 400 085 (India); Vijayan, Pallippattu K., E-mail: vijayanp@barc.gov.in [Reactor Engineering Division, Bhabha Atomic Research Center, Trombay, Mumbai 400 085 (India)

    2015-10-15

    Highlights: • CFD simulations in the Calandria of an advanced reactor under natural circulation. • Under natural convection, majority of the flow recirculates within the Calandria. • Maximum temperature is located at the top and center of the fuel channel matrix. • During SBO, temperature inside Calandria is stratified. - Abstract: Passive systems are being examined for the future Advanced Nuclear Reactor designs. One of such concepts is the Passive Moderator Cooling System (PMCS), which is designed to remove heat from the moderator in the Calandria vessel passively in case of an extended Station Black Out condition. The heated heavy-water moderator (due to heat transferred from the Main Heat Transport System (MHTS) and thermalization of neutrons and gamma from radioactive decay of fuel) rises upward due to buoyancy, gets cooled down in a heat exchanger and returns back to Calandria, completing a natural circulation loop. The natural circulation should provide sufficient cooling to prevent the increase of moderator temperature and pressure beyond safe limits. In an earlier study, a full-scale 1D transient simulation was performed for the reactor including the MHTS and the PMCS, in the event of a station blackout scenario (Kumar et al., 2013). The results indicate that the systems remain within the safe limits for 7 days. However, the flow inside a geometry like Calandria is quite complex due to its large size and inner complexities of dense fuel channel matrix, which was simplified as a 1D pipe flow in the aforesaid analysis. In the current work, CFD simulations are performed to study the temperature distributions and flow distribution of moderator inside the Calandria vessel using a three-dimensional CFD code, OpenFoam 2.2.0. First, a set of steady state simulation was carried out for a band of inlet mass flow rates, which gives the minimum mass flow rate required for removing the maximum heat load, by virtue of prediction of hot spots inside the Calandria

  19. An overview of reactor vessel internals segmentation for nuclear plant decommissioning

    International Nuclear Information System (INIS)

    Litka, T.J.

    1994-01-01

    Several nuclear plants have undergone reactor vessel (RV) internals segmentation as part of or in preparation for decommissioning the plant. In addition, several other nuclear facilities are planning for similar work efforts. The primary technology used for segmentation of RV internals, whether in-air or underwater is Plasma Arc Cutting (PAC). Metal Disintegration Machining (MDM) is also used for difficult to make cuts. PAC and MDM are deployed by various means including Long Handled Tools (LHTs), fixtures, tracks, and multi-axis manipulators. These enable remote cutting due to the radiation and/or underwater environment. A Boiling Water Reactor (BWR), a Pressurized Water Reactor (PWR), and a High Temperature Gas Reactor (HTGR) have had their internals removed and segmented using PAC and MDM. The cutting technology used for each component, location of cut, cut geometry and environment had to be determined well before the actual cutting operations. This allowed for the design, fabrication, and testing of the delivery systems. The technologies, selection process, and methodology for RV internals segmentation will be discussed in this paper

  20. Physical aspects of the Canadian generation IV supercritical water-cooled pressure tube reactor plant design

    Energy Technology Data Exchange (ETDEWEB)

    Gaudet, M.; Yetisir, M.; Haque, Z. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    The form of the containment building is a function of the requirements imposed by various systems. In order to provide sufficient driving force for naturally-circulated emergency cooling systems, as well as providing a gravity-driven core flooding pool function, the Canadian SCWR reactor design relies on elevation differences between the reactor and the safety systems. These elevation differences, the required cooling pool volumes and the optimum layout of safety-related piping are major factors influencing the plant design. As a defence-in-depth, the containment building and safety systems also provide successive barriers to the unplanned release of radioactive materials, while providing a path for heat flow to the ultimate heat sink, the atmosphere. Access to the reactor for refuelling is from the top of the reactor, with water used as shielding during the refuelling operations. The accessibility to the reactor and protection of the environment are additional factors influencing the plant design. This paper describes the physical implementation of the major systems of the Canadian SCWR within the reactor building, and the position of major plant services relative to the reactor building. (author)

  1. Studies on environment safety and application of advanced reactor for inland nuclear power plants

    International Nuclear Information System (INIS)

    Wei, L.; Jie, L.

    2014-01-01

    To study environment safety assessment of inland nuclear power plants (NPPs), the impact of environment safety under the normal operation was researched and the environment risk of serious accidents was analyzed. Moreover, the requirements and relevant provisions of site selection between international nuclear power plant and China's are comparatively studied. The conclusion was that the environment safety assessment of inland and coastal nuclear power plant have no essential difference; the advanced reactor can meet with high criteria of environment safety of inland nuclear power plants. In this way, China is safe and feasible to develop inland nuclear power plant. China's inland nuclear power plants will be as big market for advanced reactor. (author)

  2. Analysis of unprotected overcooling events in the Integral Fast Reactor

    International Nuclear Information System (INIS)

    Vilim, R.B.

    1989-01-01

    Simple analytic models are developed for predicting the response of a metal fueled, liquid-metal cooled reactor to unprotected overcooling events in the balance of plant. All overcooling initiators are shown to fall into two categories. The first category contains these events for which there is no final equilibrium state of constant overcooling, as in the case for a large steam leak. These events are analyzed using a non-flow control mass approach. The second category contains those events which will eventually equilibrate, such as a loss of feedwater heaters. A steady flow control volume analysis shows that these latter events ultimately affect the plant through the feedwater inlet to the steam generator. The models developed for analyzing these two categories provide upper bounds for the reactor's passive response to overcooling accident initiators. Calculation of these bounds for a prototypic plant indicate that failure limits -- eutectic melting, sodium boiling, fuel pin failure -- are not exceeded in any overcooling event. 2 refs

  3. Decommissioning and decontamination of licensed reactor facilities and demonstration nuclear power plants

    International Nuclear Information System (INIS)

    Lear, G.; Erickson, P.B.

    1975-01-01

    Decommissioning of licensed reactors and demonstration nuclear power plants has been accomplished by mothballing (protective storage), entombment, and dismantling or a combination of these three. The alternative selected by a licensee seems to be primarily based on cost. A licensee must, however, show that the decommissioning process provides adequate protection of the health and safety of the public and no adverse impact on the environment. To date the NRC has approved each of the alternatives in the decommissioning of different facilities. The decommissioning of small research reactors has been accomplished primarily by dismantling. Licensed nuclear power plants, however, have been decommissioned primarily by being placed in a mothballed state in which they continue to retain a reactor license and the associated licensee responsibilities

  4. Market introduction of innovative reactors

    International Nuclear Information System (INIS)

    Heek, A.I.V.

    1996-01-01

    Besides the development of evolutionary and passive LWR, also that of innovative reactors is attractive, because other applications (new markets) besides base load electricity generation can be thought of, and interesting new features on the area of safety or waste incineration can be shown. For market introduction however, a (partial) new infrastructure and a demonstration plant are required. Taking the abundance of fossil fuels and the accompanying low fuel prices today and in the near future into account, the funds to finance this will only become available when 1)the projected energy generating costs will be substantially lower than those of today, and 2)the costs of market introduction (i.e. the demonstration plant and the required infrastructure) will be limited. Generally speaking, there are two ways to seek competitiveness of a reactor type: 1)application of economy of scale, and 2)simplification. In this paper, an example of the second possibility is pursued for an innovative reactor type. The HR1 is a 40 MWth high temperature gas cooled reactor for heat and power cogeneration, a simplified version of the German HTR Module. The power level is chosen so small that additional safety features become apparent. For example, after a total loss of coolant the fuel remains fully intact, even if the reactor shutdown system fails and the reactor goes critical again after a number of hours. These safety features are used to omit certain components, like the emergency core cooling system, or to select a cheaper version of components, e.g. replacing the containment building by a confinement. Moreover, degradation of the safety class of certain components comes within the realm of possibilities. The cost reduction offered by these two measures are used to more than offset the economy-of-scale disadvantage of this small reactor system. (author)

  5. Inventory of nuclear power plants and research reactors temporary or definitively stopped in industrialized countries

    International Nuclear Information System (INIS)

    Clauzon, J.; Vaubert, B.

    1984-12-01

    This paper presents data and information on the end of the life of nuclear reactors. One deals more particularly with installations of industrialized countries. This report gives the motivations which have involved the definitive shut down of nuclear power plants and of research reactors in the concerned countries. A schedule of definitive reactor shutdowns is presented. Then, one deals with nuclear power plants of which the construction has been stopped. The reasons of these situations are also given. The temporary difficulties met during the construction or the starting of nuclear power plants these last years are mentioned. Most times, there are economical or political considerations, or safety reasons. Finally, the nuclear power plants stopped for more than two years are mentioned [fr

  6. Analysis of the running-in phase of a Passively Safe Thorium Breeder Pebble Bed Reactor

    International Nuclear Information System (INIS)

    Wols, F.J.; Kloosterman, J.L.; Lathouwers, D.; Hagen, T.H.J.J. van der

    2015-01-01

    Highlights: • This work analyzes important trends of the running-in phase of a thorium breeder PBR. • Depletion equations are solved for important actinides and a fission product pair. • Breeding U-233 is achieved in 7 years by cleverly adjusting the feed fuel enrichment. • A safety analysis shows the thorium PBR is passively safe during the running-in phase. - Abstract: The present work investigates the running-in phase of a 100 MW th Passively Safe Thorium Breeder Pebble Bed Reactor (PBR), a conceptual design introduced in previous equilibrium core design studies by the authors. Since U-233 is not available in nature, an alternative fuel, e.g. U-235/U-238, is required to start such a reactor. This work investigates how long it takes to converge to the equilibrium core composition and to achieve a net production of U-233, and how this can be accelerated. For this purpose, a fast and flexible calculation scheme was developed to analyze these aspects of the running-in phase. Depletion equations with an axial fuel movement term are solved in MATLAB for the most relevant actinides (Th-232, Pa-233, U-233, U-234, U-235, U-236 and U-238) and the fission products are lumped into a fission product pair. A finite difference discretization is used for the axial coordinate in combination with an implicit Euler time discretization scheme. Results show that a time dependent adjustment scheme for the enrichment (in case of U-235/U-238 start-up fuel) or U-233 weight fraction of the feed driver fuel helps to restrict excess reactivity, to improve the fuel economy and to achieve a net production of U-233 faster. After using U-235/U-238 startup fuel for 1300 days, the system starts to work as a breeder, i.e. the U-233 (and Pa-233) extraction rate exceeds the U-233 feed rate, within 7 years after start of reactor operation. The final part of the work presents a basic safety analysis, which shows that the thorium PBR fulfills the same passive safety requirements as the

  7. Expansion of passive safety function

    International Nuclear Information System (INIS)

    Inai, Nobuhiko; Nei, Hiromichi; Kumada, Toshiaki.

    1995-01-01

    Expansion of the use of passive safety functions is proposed. Two notions are presented. One is that, in the design of passive safety nuclear reactors where aversion of active components is stressed, some active components are purposely introduced, by which a system is built in such a way that it behaves in an apparently passive manner. The second notion is that, instead of using a passive safety function alone, a passive safety function is combined with some active components, relating the passivity in the safety function with enhanced controllability in normal operation. The nondormant system which the authors propose is one example of the first notion. This is a system in which a standby safety system is a portion of the normal operation system. An interpretation of the nondormant system via synergetics is made. As an example of the second notion, a PIUS density lock aided with active components is proposed and is discussed

  8. Concept research on general passive system

    International Nuclear Information System (INIS)

    Han Xu; Yang Yanhua; Zheng Mingguang

    2009-01-01

    This paper summarized the current passive techniques used in nuclear power plants. Through classification and analysis, the functional characteristics and inherent identification of passive systems were elucidated. By improving and extending the concept of passive system, the general passive concept was proposed, and space and time relativity was discussed and assumption of general passive system were illustrated. The function of idealized general passive system is equivalent with the current passive system, but the design of idealized general passive system is more flexible. (authors)

  9. Reliability of thermal-hydraulic passive safety systems

    International Nuclear Information System (INIS)

    D'Auria, F.; Araneo, D.; Pierro, F.; Galassi, G.

    2014-01-01

    The scholar will be informed of reliability concepts applied to passive system adopted for nuclear reactors. Namely, for classical components and systems the failure concept is associated with malfunction of breaking of hardware. In the case of passive systems the failure is associated with phenomena. A method for studying the reliability of passive systems is discussed and is applied. The paper deals with the description of the REPAS (Reliability Evaluation of Passive Safety System) methodology developed by University of Pisa (UNIPI) and with results from its application. The general objective of the REPAS methodology is to characterize the performance of a passive system in order to increase the confidence toward its operation and to compare the performances of active and passive systems and the performances of different passive systems

  10. Medium-Power Lead-Alloy Fast Reactor Balance-of-Plant Options

    International Nuclear Information System (INIS)

    Dostal, Vaclav; Hejzlar, Pavel; Todreas, Neil E.; Buongiorno, Jacopo

    2004-01-01

    Proper selection of the power conversion cycle is a very important step in the design of a nuclear reactor. Due to the higher core outlet temperature (∼550 deg. C) compared to that of light water reactors (∼300 deg. C), a wide portfolio of power cycles is available for the lead alloy fast reactor (LFR). Comparison of the following cycles for the LFR was performed: superheated steam (direct and indirect), supercritical steam, helium Brayton, and supercritical CO 2 (S-CO 2 ) recompression. Heat transfer from primary to secondary coolant was first analyzed and then the steam generators or heat exchangers were designed. The direct generation of steam in the lead alloy coolant was also evaluated. The resulting temperatures of the secondary fluids are in the range of 530-545 deg. C, dictated by the fixed space available for the heat exchangers in the reactor vessel. For the direct steam generation situation, the temperature is 312 deg. C. Optimization of each power cycle was carried out, yielding net plant efficiency of around 40% for the superheated steam cycle while the supercritical steam and S-CO 2 cycles achieved net plant efficiency of 41%. The cycles were then compared based on their net plant efficiency and potential for low capital cost. The superheated steam cycle is a very good candidate cycle given its reasonably high net plant efficiency and ease of implementation based on the extensive knowledge and operating experience with this cycle. Although the supercritical steam cycle net plant efficiency is slightly better than that of the superheated steam cycle, its high complexity and high pressure result in higher capital cost, negatively affecting plant economics. The helium Brayton cycle achieves low net plant efficiency due to the low lead alloy core outlet temperature, and therefore, even though it is a simpler cycle than the steam cycles, its performance is mediocre in this application. The prime candidate, however, appears to be the S-CO 2

  11. Plant to reduce reactor scrams and ASSET related analysis

    Energy Technology Data Exchange (ETDEWEB)

    Piirto, A [Teollisuuden Voima Oy (Finland)

    1997-10-01

    The report of events at the plant follows established rules. Basically, three categories of reports exist: first, the reactor scram report, secondly, the operational disturbance report and thirdly, the special report. The last named category covers events defined by authorities. It concentrates on safety related events, for example on failures to follow the requirements stipulated in the plant`s Technical Specifications. In general, special events are nuclear safety on the plant, the safety of the plant personnel or overall, the radiation safety in the plant`s vicinity. In the TVO`s experience feedback activity the greatest emphasis is put on events at the TVO plant. The events on the same type of plants come second. Due to limited resources, somewhat less attention is paid to events on the other types of plants. However, the experience feedback should become wider in practice so that it would be a part of everyday life in nuclear power plant operation.

  12. Project of SVBR-75/100 reactor plant with improved safety for nuclear sources of small and medium power

    International Nuclear Information System (INIS)

    Dragunov, Yu. G.; Stepanov, V. S.; Klimov, N. N.; Dedul, A. V.; Bolvanchikov, S. N.; Zrodnikov, A. V.; Tolhinsky, G. I.; Komlev, O. G.

    2004-01-01

    As a result of the joint work performed recently by FSUE OKB Gidropress, SNC RF-IPPE and other organizations the technical feasibility is shown for creation and usage in nuclear power engineering of the unified reactor plant (RP) SVBR-75/100 with fast neutron reactor core and lead-bismuth coolant (LBC) in the primary circuit. Technical design of SVBR-75/100 reactor plant is based on the following: 50-year operation experience in development and operation of RP with LBC for nuclear submarines; experience in development and operation of fast reactor with sodium coolant; experience in optimization of LBC technology at nuclear submarines and ground-based test benches; conceptual design of SVBR-75 reactor plant (for renovation of Units 2, 3 and 4 of Novovoronezh NPP). Technical solutions laid down in the basis of SVBR-75/100 reactor plant design are oriented towards the industrial basis, structural materials existing in Russia, as well as the unique LBC technology with experimental and practical support. The concept of SVBR-75/100 reactor plant safety assurance is based on the following provisions: maximum usage of inherent safety supported by physical features of fast neutron reactor, chemically inert LBC in the primary circuit, integral layout and special design solutions; maximum possible combination of normal operation and safety functions in RP systems. Small power of SVBR-75/100 RP makes it possible to manufacture the complete set of RP main equipment at the factory and delivery it to NPP site as-finished practically using any transport including railway. Possible fields of application of SVBR-75/100 reactor plants: modular NPPs of different power; renovation of NPPs with light water reactors exhausted their service life; independent nuclear power sources for different applications (ground-based nuclear water-desalinating plants, etc. )(author)

  13. Quality assurance in technology development for The Clinch River Breeder Reactor Plant Project

    International Nuclear Information System (INIS)

    Anderson, J.W.

    1980-01-01

    The Clinch River Breeder Reactor Plant Project is the nation's first large-scale demonstration of the Liquid Metal Fast Breeder Reactor (LMFBR) concept. The Project has established an overall program of plans and actions to assure that the plant will perform as required. The program has been established and is being implemented in accordance with Department of Energy Standard RDT F 2-2. It is being applied to all parts of the plant, including the development of technology supporting its design and licensing activity. A discussion of the program as it is applied to development is presented

  14. Operational margin monitoring system for boiling water reactor power plants

    International Nuclear Information System (INIS)

    Fukutomi, S.; Takigawa, Y.

    1992-01-01

    This paper reports on an on-line operational margin monitoring system which has been developed for boiling water reactor power plants to improve safety, reliability, and quality of reactor operation. The system consists of a steady-state core status prediction module, a transient analysis module, a stability analysis module, and an evaluation and guidance module. This system quantitatively evaluates the thermal margin during abnormal transients as well as the stability margin, which cannot be evaluated by direct monitoring of the plant parameters, either for the current operational state or for a predicted operating state that may be brought about by the intended operation. This system also gives operator guidance as to appropriate or alternate operations when the operating state has or will become marginless

  15. Prism reactor system design and analysis of postulated unscrammed events

    International Nuclear Information System (INIS)

    Van Tuyle, G.J.; Slovik, G.C.

    1991-08-01

    Key safety characteristics of the PRISM reactor system include the passive reactor shutdown characteristic and the passive shutdown heat removal system, RVACS. While these characteristics are simple in principle, the physical processes are fairly complex, particularly for the passive reactor shutdown. It has been possible to adapt independent safety analysis codes originally developed for the Clinch River Breeder Reactor review, although some limitations remain. In this paper, the analyses of postulated unscrammed events are discussed, along with limitations in the predictive capabilities and plans to correct the limitations in the near future. 6 refs., 4 figs

  16. PRISM reactor system design and analysis of postulated unscrammed events

    International Nuclear Information System (INIS)

    Van Tuyle, G.J.; Slovik, G.C.; Rosztoczy, Z.; Lane, J.

    1991-01-01

    Key safety characteristics of the PRISM reactor system include the passive reactor shutdown characteristics and the passive shutdown heat removal system, RVACS. While these characteristics are simple in principle, the physical processes are fairly complex, particularly for the passive reactor shutdown. It has been possible to adapt independent safety analysis codes originally developed for the Clinch River Breeder Reactor review, although some limitations remain. In this paper, the analyses of postulated unscrammed events are discussed, along with limitations in the predictive capabilities and plans to correct the limitations in the near future. 6 refs., 4 figs

  17. East-Asia nuclear/fossil power plant competitiveness

    International Nuclear Information System (INIS)

    Braun, Ch.

    1996-01-01

    The competitiveness of a new nuclear plant vs. a new oil or gas fired combined cycle plant or a coal fired plant in East-Asia, is reviewed in the paper. Both the nuclear and the fossil fired plants are evaluated as either utility financed or independent power producer (IPP) financed. Two types of advanced light water reactors (ALWRs) are considered in this paper, namely evolutionary ALWRs (1200 MWe size) and passive ALWRs (600 MWe class). A range of capital and total generation costs for each plant type is reported here. The comparison centers on three elements of overall competitiveness: generation costs, hard currency requirements, and employment requirements. Each of these aspects is considered perspective. Year-by-Year generation cost history over the plant lifetime is shown in some cases. It is found here that a utility financed evolutionary and passive ALWRs are broadly competitive with an IPP financed gas fired combined cycle plant and are more economic than oil fired combined cycle or a coal fired plant. A single unit evolutionary ALWR may have a 12 - 15 % capital cost advantage over a single passive ALWR then adjusted on a per KWe basis. Front-end hard currency requirements of a passive ALWR are 2.5 times higher than for a combined plant and evolutionary ALWRs requires 3.6 times higher up-front cost. However, on a lifetime basis, passive ALWR net hard currency requirements are two times lower than for a combined cycle plant. Evolutionary ALWR net hard currency requirements are three times over than those of a combined cycle plant. The effects of domestic vs. world price of fossil fuels on relative nuclear competitiveness are reviewed in this nuclear competitiveness paper. Employment requirements in an ALWR during both the construction period and lifetime operation, exceed the requirements for oil or gas fired plants by a factor of five. While contributing to overall plant cost, employment requirements can also be viewed as opportunity to increase national

  18. Safety Philosophy in Process Heat Plants Coupled to High Temperature Reactors

    International Nuclear Information System (INIS)

    Brown, Nicholas R.; Revankar, Shripad T.

    2011-01-01

    With the future availability of fossil fuel resources in doubt, high temperature nuclear reactors have the potential to be an important technology in the near term. Due to a high coolant outlet temperature, high temperature reactors (HTR) can be used to drive chemical plants that directly utilize process heat. Additionally, the high temperature improves the thermodynamic efficiency of the energy utilization. Many applications of high temperature reactors exist as a thermal driving vector for endothermic chemical process plants. Hydrogen generation using the General Atomics (GA) sulfur iodine (SI) cycle is one promising application of high temperature nuclear heat. The main chemical reactions in the SI cycle are: 1. I 2 +SO 2 + 2H 2 O → 2HI + H 2 SO 4 (Bunsen reaction) 2. H 2 SO 4 → H 2 O + SO 2 + 1/2O 2 (Sulfuric acid decomposition) 3. 2HI → H 2 + I 2 (Hydrogen Iodide decomposition). With the exception of hydrogen and oxygen, all relevant reactants are recycled within the process. However, there are many unresolved safety and operational issues related to implementation of such a coupled plant

  19. Fuel handling grapple for nuclear reactor plants

    International Nuclear Information System (INIS)

    Rousar, D.L.

    1992-01-01

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

  20. Development of Safety Analysis Technology for Integral Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sim, S. K. [Korea Atomic Energy Research Institute, Taejeon (Korea); Seul, K. W.; Kim, W. S.; Kim, W. K.; Yun, Y. G.; Ahn, H. J.; Lee, J. S.; Sin, A. D. [Korea Institute of Nuclear Safety, Taejeon (Korea)

    2000-03-01

    The Nuclear Desalination Plant(NDP) is being developed to produce electricity and fresh water, and is expected to locate near population zone. In the aspect of safety, it is required to protect the public and environment from the possible releases of fission products and to prevent the fresh water from the contamination of radioactivity. Thus, in a present study, the safety characteristics of the integral reactor adopting passive and inherent safety features significantly different from existing nuclear power plants were investigated based on the design of foreign and domestic integral reactors. Also, safety requirements applicable to the NDP were analyzed based on the regulatory requirements for current and advanced reactor designs, and use requirements for small-medium size reactors. Based on these analyses, some safety concerns to be considered in the design stage have been identified. They includes the use of proven technology for new safety systems, the systematic classification and selection of design basis accidents, and the safety assurance of desalination-related systems. These efforts to identify and resolve the safety concerns in the design stage will provide the early confidence of SMART safety to designers, and the technical basis to evaluate the safety to reviewers in the future. 8 refs., 20 figs., 4 tabs. (Author)