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Sample records for plant design requirement

Advanced Neutron Sources: Plant Design Requirements

International Nuclear Information System (INIS)

1990-07-01

The Advanced Neutron Source (ANS) is a new, world class facility for research using hot, thermal, cold, and ultra-cold neutrons. At the heart of the facility is a 350-MW th , heavy water cooled and moderated reactor. The reactor is housed in a central reactor building, with supporting equipment located in an adjoining reactor support building. An array of cold neutron guides fans out into a large guide hall, housing about 30 neutron research stations. Office, laboratory, and shop facilities are included to provide a complete users facility. The ANS is scheduled to begin operation at the Oak Ridge National Laboratory at the end of the decade. This Plant Design Requirements document defines the plant-level requirements for the design, construction, and operation of the ANS. This document also defines and provides input to the individual System Design Description (SDD) documents. Together, this Plant Design Requirements document and the set of SDD documents will define and control the baseline configuration of the ANS
Safety of Nuclear Power Plants: Design. Specific Safety Requirements

International Nuclear Information System (INIS)

2012-01-01

On the basis of the principles included in the Fundamental Safety Principles, IAEA Safety Standards Series No. SF-1, this Safety Requirements publication establishes requirements applicable to the design of nuclear power plants. It covers the design phase and provides input for the safe operation of the power plant. It elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.
Safety of nuclear power plants: Design. Safety requirements

International Nuclear Information System (INIS)

2000-01-01

The present publication supersedes the Code on the Safety of Nuclear Power Plants: Design (Safety Series No. 50-C-D (Rev. 1), issued in 1988). It takes account of developments relating to the safety of nuclear power plants since the Code on Design was last revised. These developments include the issuing of the Safety Fundamentals publication, The Safety of Nuclear Installations, and the present revision of various safety standards and other publications relating to safety. Requirements for nuclear safety are intended to ensure adequate protection of site personnel, the public and the environment from the effects of ionizing radiation arising from nuclear power plants. It is recognized that technology and scientific knowledge advance, and nuclear safety and what is considered adequate protection are not static entities. Safety requirements change with these developments and this publication reflects the present consensus. This Safety Requirements publication takes account of the developments in safety requirements by, for example, including the consideration of severe accidents in the design process. Other topics that have been given more detailed attention include management of safety, design management, plant ageing and wearing out effects, computer based safety systems, external and internal hazards, human factors, feedback of operational experience, and safety assessment and verification. This publication establishes safety requirements that define the elements necessary to ensure nuclear safety. These requirements are applicable to safety functions and the associated structures, systems and components, as well as to procedures important to safety in nuclear power plants. It is expected that this publication will be used primarily for land based stationary nuclear power plants with water cooled reactors designed for electricity generation or for other heat production applications (such as district heating or desalination). It is recognized that in the case of
Safety of nuclear power plants: Design. Safety requirements

International Nuclear Information System (INIS)

2004-01-01

The present publication supersedes the Code on the Safety of Nuclear Power Plants: Design (Safety Series No. 50-C-D (Rev. 1), issued in 1988). It takes account of developments relating to the safety of nuclear power plants since the Code on Design was last revised. These developments include the issuing of the Safety Fundamentals publication, The Safety of Nuclear Installations, and the present revision of various safety standards and other publications relating to safety. Requirements for nuclear safety are intended to ensure adequate protection of site personnel, the public and the environment from the effects of ionizing radiation arising from nuclear power plants. It is recognized that technology and scientific knowledge advance, and nuclear safety and what is considered adequate protection are not static entities. Safety requirements change with these developments and this publication reflects the present consensus. This Safety Requirements publication takes account of the developments in safety requirements by, for example, including the consideration of severe accidents in the design process. Other topics that have been given more detailed attention include management of safety, design management, plant ageing and wearing out effects, computer based safety systems, external and internal hazards, human factors, feedback of operational experience, and safety assessment and verification. This publication establishes safety requirements that define the elements necessary to ensure nuclear safety. These requirements are applicable to safety functions and the associated structures, systems and components, as well as to procedures important to safety in nuclear power plants. It is expected that this publication will be used primarily for land based stationary nuclear power plants with water cooled reactors designed for electricity generation or for other heat production applications (such as district heating or desalination). It is recognized that in the case of
Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Spanish Edition)

International Nuclear Information System (INIS)

2012-01-01

This publication is a revision of Safety Requirements No. NS-R-1, Safety of Nuclear Power Plants: Design. It establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in the design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.
Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Russian Edition)

International Nuclear Information System (INIS)

2012-01-01

This publication is a revision of Safety Requirements No. NS-R-1, Safety of Nuclear Power Plants: Design. It establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in the design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.
Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Chinese Ed.)

International Nuclear Information System (INIS)

2012-01-01

On the basis of the principles included in the Fundamental Safety Principles, IAEA Safety Standards Series No. SF-1, this Safety Requirements publication establishes requirements applicable to the design of nuclear power plants. It covers the design phase and provides input for the safe operation of the power plant. It elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.
Safety of Nuclear Power Plants: Design. Specific Safety Requirements (French Ed.)

International Nuclear Information System (INIS)

2012-01-01

On the basis of the principles included in the Fundamental Safety Principles, IAEA Safety Standards Series No. SF-1, this Safety Requirements publication establishes requirements applicable to the design of nuclear power plants. It covers the design phase and provides input for the safe operation of the power plant. It elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.
Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Arabic Ed.)

International Nuclear Information System (INIS)

2012-01-01

On the basis of the principles included in the Fundamental Safety Principles, IAEA Safety Standards Series No. SF-1, this Safety Requirements publication establishes requirements applicable to the design of nuclear power plants. It covers the design phase and provides input for the safe operation of the power plant. It elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.
Safety of Nuclear Power Plants: Design. Specific Safety Requirements (French Edition)

International Nuclear Information System (INIS)

2017-01-01

This publication establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. A review of Safety Requirements publications was commenced in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan. The review revealed no significant areas of weakness and resulted in just a small set of amendments to strengthen the requirements and facilitate their implementation, which are contained in the present publication.
Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Russian Edition)

International Nuclear Information System (INIS)

2016-01-01

This publication establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. A review of Safety Requirements publications was commenced in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan. The review revealed no significant areas of weakness and resulted in just a small set of amendments to strengthen the requirements and facilitate their implementation, which are contained in the present publication.
Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Arabic Edition)

International Nuclear Information System (INIS)

2017-01-01

This publication establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. A review of Safety Requirements publications was commenced in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan. The review revealed no significant areas of weakness and resulted in just a small set of amendments to strengthen the requirements and facilitate their implementation, which are contained in the present publication.
Future CANDU nuclear power plant design requirements document executive summary

International Nuclear Information System (INIS)

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

1996-03-01

The future CANDU Requirements Document (FCRED) describes a clear and complete statement of utility requirements for the next generation of CANDU nuclear power plants including those in Korea. The requirements are based on proven technology of PHWR experience and are intended to be consistent with those specified in the current international requirement documents. Furthermore, these integrated set of design requirements, incorporate utility input to the extent currently available and assure a simple, robust and more forgiving design that enhances the performance and safety. The FCRED addresses the entire plant, including the nuclear steam supply system and the balance of the plant, up to the interface with the utility grid at the distribution side of the circuit breakers which connect the switchyard to the transmission lines. Requirements for processing of low level radioactive waste at the plant site and spent fuel storage requirements are included in the FCRED. Off-site waste disposal is beyond the scope of the FCRED. 2 tabs., 1 fig. (Author) .new
Magnetohydrodynamics (MHD) Engineering Test Facility (ETF) 200 MWe power plant. Design Requirements Document (DRD)

Science.gov (United States)

Rigo, H. S.; Bercaw, R. W.; Burkhart, J. A.; Mroz, T. S.; Bents, D. J.; Hatch, A. M.

1981-01-01

A description and the design requirements for the 200 MWe (nominal) net output MHD Engineering Test Facility (ETF) Conceptual Design, are presented. Performance requirements for the plant are identified and process conditions are indicated at interface stations between the major systems comprising the plant. Also included are the description, functions, interfaces and requirements for each of these major systems. The lastest information (1980-1981) from the MHD technology program are integrated with elements of a conventional steam electric power generating plant.
Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Russian Edition); Bezopasnost' atomnykh ehlektrostantsij: proektirovanie. Konkretnye trebovaniya bezopasnosti

Energy Technology Data Exchange (ETDEWEB)

NONE

2012-04-15

This publication is a revision of Safety Requirements No. NS-R-1, Safety of Nuclear Power Plants: Design. It establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in the design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.
Advanced Neutron Source: Plant Design Requirements

International Nuclear Information System (INIS)

1990-07-01

The Advanced Neutron Source will be a new world-class facility for research using hot, thermal, cold, and ultra-cold neutrons. The heart of the facility will be a 330-MW (fission), heavy-water cooled and heavy-water moderated reactor. The reactor will be housed in a central reactor building, with supporting equipment located in an adjoining reactor support building. An array of cold neutron guides will fan out into a large guide hall, housing about 30 neutron research stations. Appropriate office, laboratory, and shop facilities will be included to provide a complete facility for users. The ANS is scheduled to begin operation at the Oak Ridge National Laboratory early in the next decade. This PDR document defines the plant-level requirements for the design, construction, and operation of ANS. It also defines and provides input to the individual System Design Description (SDD) documents. Together, this PDR document and the set of SDD documents will define and control the baseline configuration of ANS
Design requirements, criteria and methods for seismic qualification of CANDU power plants

International Nuclear Information System (INIS)

Singh, N.; Duff, C.G.

1979-10-01

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

Energy Technology Data Exchange (ETDEWEB)

Stevenson, J D

1976-04-01

Since 1965, when extreme load requirements began to be considered explicitly in nuclear power plant design, there has been a gradual divergence in requirements imposed by national regulatory agencies. However, nuclear plant safety is an international problem because of the potential international effects of any postulated plant failure. For this reason this paper has been prepared in an attempt to highlight the differences in national criteria currently used in the extreme load design of nuclear plant facilities. No attempt has been made to evaluate the relative merit of the criteria established by the various national regulatory agencies. This paper presents the results of a recent survey made of national atomic energy regulatory agencies and major nuclear steam supply design agencies, which requested a summary of current licensing criteria associated with earthquake, extreme wind (tornado), flood, airplane crash and accident (pipe break) loads applicable within the various national jurisdictions. Also presented are a number of comparisons which are meant to illustrate the differences in national regulatory criteria.
Survey of extreme load design regulatory agency licensing requirements for nuclear power plants

International Nuclear Information System (INIS)

Stevenson, J.D.

1976-01-01

Since 1965, when extreme load requirements began to be considered explicitly in nuclear power plant design, there has been a gradual divergence in requirements imposed by national regulatory agencies. However, nuclear plant safety is an international problem because of the potential international effects of any postulated plant failure. For this reason this paper has been prepared in an attempt to highlight the differences in national criteria currently used in the extreme load design of nuclear plant facilities. No attempt has been made to evaluate the relative merit of the criteria established by the various national regulatory agencies. This paper presents the results of a recent survey made of national atomic energy regulatory agencies and major nuclear steam supply design agencies, which requested a summary of current licensing criteria associated with earthquake, extreme wind (tornado), flood, airplane crash and accident (pipe break) loads applicable within the various national jurisdictions. Also presented are a number of comparisons which are meant to illustrate the differences in national regulatory criteria. (Auth.)
Risk-informed assessment of regulatory and design requirements for future nuclear power plants. Annual report

International Nuclear Information System (INIS)

2000-01-01

OAK B188 Risk-informed assessment of regulatory and design requirements for future nuclear power plants. Annual report. The overall goal of this research project is to support innovation in new nuclear power plant designs. This project is examining the implications, for future reactors and future safety regulation, of utilizing a new risk-informed regulatory system as a replacement for the current system. This innovation will be made possible through development of a scientific, highly risk-formed approach for the design and regulation of nuclear power plants. This approach will include the development and/or confirmation of corresponding regulatory requirements and industry standards. The major impediment to long term competitiveness of new nuclear plants in the U.S. is the capital cost component--which may need to be reduced on the order of 35% to 40% for Advanced Light Water Reactors (ALWRS) such as System 80+ and Advanced Boiling Water Reactor (ABWR). The required cost reduction for an ALWR such as AP600 or AP1000 would be expected to be less. Such reductions in capital cost will require a fundamental reevaluation of the industry standards and regulatory bases under which nuclear plants are designed and licensed. Fortunately, there is now an increasing awareness that many of the existing regulatory requirements and industry standards are not significantly contributing to safety and reliability and, therefore, are unnecessarily adding to nuclear plant costs. Not only does this degrade the economic competitiveness of nuclear energy, it results in unnecessary costs to the American electricity consumer. While addressing these concerns, this research project will be coordinated with current efforts of industry and NRC to develop risk-informed, performance-based regulations that affect the operation of the existing nuclear plants; however, this project will go further by focusing on the design of new plants

Risk-informed assessment of regulatory and design requirements for future nuclear power plants. Annual report

Energy Technology Data Exchange (ETDEWEB)

NONE

2000-08-01

OAK B188 Risk-informed assessment of regulatory and design requirements for future nuclear power plants. Annual report. The overall goal of this research project is to support innovation in new nuclear power plant designs. This project is examining the implications, for future reactors and future safety regulation, of utilizing a new risk-informed regulatory system as a replacement for the current system. This innovation will be made possible through development of a scientific, highly risk-formed approach for the design and regulation of nuclear power plants. This approach will include the development and/or confirmation of corresponding regulatory requirements and industry standards. The major impediment to long term competitiveness of new nuclear plants in the U.S. is the capital cost component--which may need to be reduced on the order of 35% to 40% for Advanced Light Water Reactors (ALWRS) such as System 80+ and Advanced Boiling Water Reactor (ABWR). The required cost reduction for an ALWR such as AP600 or AP1000 would be expected to be less. Such reductions in capital cost will require a fundamental reevaluation of the industry standards and regulatory bases under which nuclear plants are designed and licensed. Fortunately, there is now an increasing awareness that many of the existing regulatory requirements and industry standards are not significantly contributing to safety and reliability and, therefore, are unnecessarily adding to nuclear plant costs. Not only does this degrade the economic competitiveness of nuclear energy, it results in unnecessary costs to the American electricity consumer. While addressing these concerns, this research project will be coordinated with current efforts of industry and NRC to develop risk-informed, performance-based regulations that affect the operation of the existing nuclear plants; however, this project will go further by focusing on the design of new plants.
Electrical design requirements for electrode boilers for nuclear plants

International Nuclear Information System (INIS)

Kempker, M.J.

1979-01-01

Medium-voltage steam electrode boilers, in the 20- to 50-MW range, have become an attractive alternative to comparable fossil-fueled boilers as a source of auxiliary steam during the startup and normal shutdown of nuclear power plants. The electrode boiler represents a favorable option because of environmental, fire protection, and licensing considerations. However, this electrical option brings some difficult design problems for which solutions are required in order to integrate the electrode boiler into the plant low resistance grounded power system. These considerations include the effects of an unbalanced electrode boiler on the performance of polyphase induction motors, boiler grounding for personnel safety, boiler neutral grounding, and ground relaying
Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Spanish Edition); Seguridad de las centrales nucleares: Diseno. Requisitos de seguridad especificos

Energy Technology Data Exchange (ETDEWEB)

NONE

2012-04-15

This publication is a revision of Safety Requirements No. NS-R-1, Safety of Nuclear Power Plants: Design. It establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in the design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.
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
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.)
Utility/user requirements for the MHTGR desalination plant

International Nuclear Information System (INIS)

Brown, S.J.; Snyder, G.M.

1989-01-01

This paper describes the approach used by Gas-Cooled Reactor Associates (GCRA) and the Metropolitan Water District of Southern California (MWD) in developing Utility/User (U/U) Requirements for the Modular High Temperature Gas-cooled Reactor (MHTGR) Desalination Plant. This is a cogeneration plant that produces fresh water from seawater, and electricity. The U/U requirements for the reference MHTGR plant are used except for those changes necessary to: provide low-grade heat to a seawater desalination process, enable siting in a Southern California coastal area, take advantage of reduced weather extremes where substantial cost reductions are expected, and use seawater cooling instead of a cooling tower. The resulting requirements and the differences from the reference MHTGR requirements are discussed. The nuclear portion of the design is essentially the same as that for the reference MHTGR design. The major differences occur in the turbine-generator and condenser, and for the most part, the design parameters for the reference plant are found to be conservative for the desalination plant. The most important difference in requirements is in the higher seismic levels required for a Southern California site, which requires reassessment and possible modification of the design of some reference plant equipment for use in the desalination plant. (author). 5 refs, 1 tab
Design concept and its requirements of the integrated SMART nuclear desalination plant

International Nuclear Information System (INIS)

Hwang, Young Dong; Kim, Young In; Chon, Bong Hyun; Lee, Doo Jung; Chang, Moon Hee

2001-02-01

The integrated SMART desalination plant consists of four(4) units of Multi Effect Distillation Process combined with Thermal-Vapor Compressor(MED-TVC) and coupled with the extracted steam from turbine through the steam transformer. Steam transformer produces the main pressure steam and supplies to the MED-TVC unit. Each distillation unit has the production the capacity of 10,000 m3/day of distilled water per day at top brine temperature of 65 deg C using the seawater supplied at temperature of 33 deg C. MED-TVC was selected as a desalination process coupled with SMART, since the thermal vapor compression is very effective where the steam is available at high temperature and pressure conditions than required in the evaporator. The MED-TVC unit is consisted of the steam supply system, vapor and condensate system, seawater supply system, brine system and chemical dosing system. The standard design of the SMART desalination plant is under development as a part of the SMART project. This report describes design concept of these systems and their requirements
Design concept and its requirements of the integrated SMART nuclear desalination plant

Energy Technology Data Exchange (ETDEWEB)

Hwang, Young Dong; Kim, Young In; Chon, Bong Hyun; Lee, Doo Jung; Chang, Moon Hee

2001-02-01

The integrated SMART desalination plant consists of four(4) units of Multi Effect Distillation Process combined with Thermal-Vapor Compressor(MED-TVC) and coupled with the extracted steam from turbine through the steam transformer. Steam transformer produces the main pressure steam and supplies to the MED-TVC unit. Each distillation unit has the production the capacity of 10,000 m3/day of distilled water per day at top brine temperature of 65 deg C using the seawater supplied at temperature of 33 deg C. MED-TVC was selected as a desalination process coupled with SMART, since the thermal vapor compression is very effective where the steam is available at high temperature and pressure conditions than required in the evaporator. The MED-TVC unit is consisted of the steam supply system, vapor and condensate system, seawater supply system, brine system and chemical dosing system. The standard design of the SMART desalination plant is under development as a part of the SMART project. This report describes design concept of these systems and their requirements.
Design and simulation of a plant control system for a GCFR demonstration plant

International Nuclear Information System (INIS)

Estrine, E.A.; Greiner, H.G.

1980-02-01

A plant control system is being designed for a 300 MW(e) Gas Cooled Fast Breeder Reactor (GCFR) demonstration plant. Control analysis is being performed as an integral part of the plant design process to ensure that control requirements are satisfied as the plant design evolves. Plant models and simulations are being developed to generate information necessary to further define control system requirements for subsequent plant design iterations
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
Operation and maintenance requirements of system design bases

International Nuclear Information System (INIS)

Banerjee, A.K.; Hanley, N.E.

1989-01-01

All system designs make assumptions about system operation testing, inspection, and maintenance. Existing industry codes and standards explicitly address design requirements of new systems, while issues related to system and plant reliability, life, design margins, effects of service conditions, operation, maintenance, etc., usually are implicit. However, system/component design documents of existing power plants often address the code requirements without considering the operation, maintenance, inspection, and testing (OMIT) requirements. The nuclear industry is expending major efforts at most nuclear power plants to reassemble and/or reconstitute system design bases. Stone ampersand Webster Engineering Corporation (SWEC) recently addressed the OMIT requirements of system/component design as an integral part of a utility's preventive maintenance program. For each component, SWEC reviewed vendor recommendations, NPRDS data/industry experience, the existing maintenance program, component service conditions, and actual plant experience. A maintenance program that considers component service conditions and plant experience ensures a connection between maintenance and design basis. Root cause analysis of failure and engineering evaluation of service condition are part of the program. System/component OMIT requirements also are compared against system design, service condition, degradation mechanism, etc., through system/component life-cycle evaluation
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.)
Systems engineering requirements impacting MHTGR circulator design

International Nuclear Information System (INIS)

Chi, H.W.; Baccaglini, G.M.; Potter, R.C.; Shenoy, A.S.

1988-01-01

At the initiation of the MHTGR program, an important task involved translating the plant users' requirements into design conditions. This was particularly true in the case of the heat transport and shutdown cooling systems since these embody many components. This paper addresses the two helium circulators in these systems. An integrated approach is being used in the development of design and design documentation for the MHTGR plant. It is an organized and systematic development of plant functions and requirements, determined by top-down design, performance, and cost trade-off studies and analyses, to define the overall plant systems, subsystems, components, and human actions. These studies, that led to the identification of the major design parameters for the two circulators, are discussed in this paper. This includes the performance information, steady state and transient data, and the various interface requirements. The design of the circulators used in the MHTGR is presented. (author). 1 ref., 17 figs
Preparation of plant and system design description documents

International Nuclear Information System (INIS)

1989-01-01

This standard prescribes the purpose, scope, organization, and content of plant design requirements (PDR) documents and system design descriptions (SDDs), to provide a unified approach to their preparation and use by a project as the principal means to establish the plant design requirements and to establish, describe, and control the individual system designs from conception and throughout the lifetime of the plant. The Electric Power Research Institute's Advanced Light Water Reactor (LWR) Requirements Document should be considered for LWR plants
NRC review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document - Evolutionary plant designs, Chapter 1, Project No. 669

International Nuclear Information System (INIS)

1992-08-01

The staff of the US Nuclear Regulatory Commission has prepared Volume 2 (Parts 1 and 2) of a safety evaluation report (SER), ''NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document -- Evolutionary Plant Designs,'' to document the results of its review of the Electric Power Research Institute's ''Advanced Light Water Reactor Utility Requirements Document.'' This SER gives the results of the staff's review of Volume II of the Requirements Document for evolutionary plant designs, which consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant (approximately 1300 megawatts-electric)
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
Current fusion power plant design concepts

International Nuclear Information System (INIS)

Gore, B.F.; Murphy, E.S.

1976-09-01

Nine current U.S. designs for fusion power plants are described in this document. Summary tabulations include a tenth concept, for which the design document was unavailable during preparation of the descriptions. The information contained in the descriptions was used to define an envelope of fusion power plant characteristics which formed the basis for definition of reference first commercial fusion power plant design. A brief prose summary of primary plant features introduces each of the descriptions contained in the body of this document. In addition, summary tables are presented. These tables summarize in side-by-side fashion, plant parameters, processes, combinations of materials used, requirements for construction materials, requirements for replacement materials during operation, and production of wastes
Considerations on the Application of the IAEA Safety Requirements for the Design of Nuclear Power Plants

International Nuclear Information System (INIS)

2016-05-01

Revised to take into consideration findings from the Fukushima Daiichi nuclear power plant accident, IAEA Safety Standards Series No. SSR-2/1 (Rev. 1), Safety of Nuclear Power Plants: Design, has introduced some new concepts with respect to the earlier safety standard published in the year 2000. The preparation of SSR-2/1 (Rev. 1) was carried out with constant and intense involvement of IAEA Member States, but some new requirements, because of the novelty of the concepts introduced and the complexity of the issues, are not always interpreted in a unique way. The IAEA is confident that a complete clarification and a full understanding of the new requirements will be available when the supporting safety guides for design and safety assessment of nuclear power plants are prepared. The IAEA expects that the effort devoted to the preparation of this publication, which received input and comments from several Member States and experts, will also facilitate and harmonize the preparation or revision of these supporting standards
Design requirements for the new reactor

International Nuclear Information System (INIS)

Koski, S.

2005-01-01

This presentation deals with the safety related design requirements specified for the new nuclear power plant to be built in Finland (FINS). The legislation, codes and standards, on which the design requirements are based, can be arranged into a hierarchical pyramid as follows: The safety related design criteria are based on the three uppermost hierarchical levels: Finnish legislation (e.g. decisions of the State Council) Basic Regulations (75-INSAG-3, USNRC General Design Criteria) Process oriented nuclear documents (YVL- guides or corresponding US/German rules). The European Utility Requirements (EUR) document was used as the starting point for the writing of the design requirements document. The structure and headlines of EUR could be kept, but in many cases the contents had to be deleted and rewritten to correspond to the requirement level of the above codes and standards. This was the case, for example, with the requirements concerning safety classification or application of failure criteria. In the presentation, the most important safety related design criteria are reviewed, with an emphasis on those requirements which exceed the requirement level applied on the existing plant units. Some hints are also given on the main differences between Finnish and international safety requirements. (orig.)
EU-APR Design in compliance with EUR Grid Requirement

Energy Technology Data Exchange (ETDEWEB)

Kim, Dong-Hwan; Lee, Keun-Sung [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

European Utility Requirements (EUR) provides technical requirements for the generation III nuclear power plant in the European countries. EUR grid requirements present the plant requirements to satisfy the needs of the grid network. The grid requirements are the precondition for the operation of a generating plant on the network. This paper describes EU-APR design which has taken account of EUR grid requirements. In this paper, EU-APR designs according to the EUR grid requirements were described. EU-APR was designed in compliance with the voltage and frequency operation field and also designed to have the capability of load following such as primary control, secondary control, and daily load following. Consequently, the EU-APR design according to the EUR grid requirements is expected to get competitiveness and enhance the license feasibility in the European nuclear market.

Safety requirements applicable to the SMART design

International Nuclear Information System (INIS)

Seul, Kwang Won; Kim, Wee Kyong; Kim, Hho Jung

1999-01-01

The 330 MW thermal power of integral reactor, named SMART (System integrated Modular Advanced ReacTor), is under development at KAERI for seawater desalination application and electricity generation. The final product of nuclear desalination plant (NDP) is electricity and fresh water. Thus, in addition to the protection of the public around the plant facility from the possible release of radioactive materials, the fresh water should be prevented from radioactivity contamination. In this study, to ensure the safety of SMART reactor in the early stage of design development, the safety requirements applicable to the SMART design were investigated, based on the current regulatory requirements for the existing NPPs and the advanced light water reactor (LWR) designs. The interface requirements related to the desalination facility were also investigated, based on the recent IAEA research activities pertaining to the NDP. As a result, it was found that the current regulatory requirements and guidance for the existing NPPs and advanced LWR designs are applicable to the SMART design and its safety evaluation. However, the safety requirements related to the SMART-specific design and the desalination plant are needed to develop in the future to assure the safety of the SMART reactor
Nuclear Energy Research Initiative. Risk Informed Assessment of Regulatory and Design Requirements for Future Nuclear Power Plants. Annual Report

International Nuclear Information System (INIS)

Ritterbusch, S.E.

2000-01-01

The overall goal of this research project is to support innovation in new nuclear power plant designs. This project is examining the implications, for future reactors and future safety regulation, of utilizing a new risk-informed regulatory system as a replacement for the current system. This innovation will be made possible through development of a scientific, highly risk-informed approach for the design and regulation of nuclear power plants. This approach will include the development and.lor confirmation of corresponding regulatory requirements and industry standards. The major impediment to long term competitiveness of new nuclear plants in the U.S. is the capital cost component--which may need to be reduced on the order of 35% to 40% for Advanced Light Water Reactors (ALWRs) such as System 80+ and Advanced Boiling Water Reactor (ABWR). The required cost reduction for an ALWR such as AP600 or AP1000 would be expected to be less. Such reductions in capital cost will require a fundamental reevaluation of the industry standards and regulatory bases under which nuclear plants are designed and licensed. Fortunately, there is now an increasing awareness that many of the existing regulatory requirements and industry standards are not significantly contributing to safety and reliability and, therefore, are unnecessarily adding to nuclear plant costs. Not only does this degrade the economic competitiveness of nuclear energy, it results in unnecessary costs to the American electricity consumer. While addressing these concerns, this research project will be coordinated with current efforts of industry and NRC to develop risk-informed, performance-based regulations that affect the operation of the existing nuclear plants; however, this project will go farther by focusing on the design of new plants
Nuclear Energy Research Initiative. Risk Informed Assessment of Regulatory and Design Requirements for Future Nuclear Power Plants. Annual Report

Energy Technology Data Exchange (ETDEWEB)

Ritterbusch, S.E.

2000-08-01

The overall goal of this research project is to support innovation in new nuclear power plant designs. This project is examining the implications, for future reactors and future safety regulation, of utilizing a new risk-informed regulatory system as a replacement for the current system. This innovation will be made possible through development of a scientific, highly risk-informed approach for the design and regulation of nuclear power plants. This approach will include the development and.lor confirmation of corresponding regulatory requirements and industry standards. The major impediment to long term competitiveness of new nuclear plants in the U.S. is the capital cost component--which may need to be reduced on the order of 35% to 40% for Advanced Light Water Reactors (ALWRs) such as System 80+ and Advanced Boiling Water Reactor (ABWR). The required cost reduction for an ALWR such as AP600 or AP1000 would be expected to be less. Such reductions in capital cost will require a fundamental reevaluation of the industry standards and regulatory bases under which nuclear plants are designed and licensed. Fortunately, there is now an increasing awareness that many of the existing regulatory requirements and industry standards are not significantly contributing to safety and reliability and, therefore, are unnecessarily adding to nuclear plant costs. Not only does this degrade the economic competitiveness of nuclear energy, it results in unnecessary costs to the American electricity consumer. While addressing these concerns, this research project will be coordinated with current efforts of industry and NRC to develop risk-informed, performance-based regulations that affect the operation of the existing nuclear plants; however, this project will go farther by focusing on the design of new plants.
NRC review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document - Evolutionary plant designs, Chapters 2--13, Project No. 669

International Nuclear Information System (INIS)

1992-08-01

The staff of the US Nuclear Regulatory Commission has prepared Volume 2 (Parts 1 and 2) of a safety evaluation report (SER), ''NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document -- Evolutionary Plant Designs,'' to document the results of its review of the Electric Power Research Institute's ''Advanced Light Water Reactor Utility Requirements Document.'' This SER gives the results of the staff's review of Volume II of the Requirements Document for evolutionary plant designs, which consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant (approximately 1300 megawatts-electric)
Status for seismic design requirements of nuclear power plants in Japan

International Nuclear Information System (INIS)

Takahashi, H.

1977-01-01

The fundamental purpose for the aseismic design of the nuclear power plants is to protect the inhabitants near the plant from radiation accidents during and after earthquake vibrations. In order to achieve the above purpose, the following considerations have been made. All buidlings, structures, system and components are classified into three Classes A, B and C according to their degree of importance for plant safety, and are designed to meet the requirements specified for each class, respectively. Magnitude and epicenter of the design basis earthquake are determined based upon seismological and geological investigations and observation of ground motion in the site, and the maximum ground acceleration which could be expected can be calculated empirically. With respect to time history waves, more than three are selected referring to dynamic characteristic of base rock in the site, observed ground motion records in the site or other strong motion seismographs.The figures of horizontal seismic coefficients to be used in determining design forces on Class A buildings and structures are 3 Co (where Co. is as defined in the Japan Building Standard Law). On the other hand the horizontal design force should not be less than those determined as the results of the dynamic analyses based on DEGM (Design Earthquake Ground Motion). The figures of horizontal seismic coefficient and forces for Class A system and components are usually determined based on the dynamic analyses for DEGM. The buildings and structures treated as an elastic column system with masses, and the bottom mass is supported by elastic springs representing the soil-foundation interaction characteristics. DEGM is used as the input disturbance in the dynamic response analysis, and the model analysis or time history method is worked out. System and components are modeled as elastic bars with lumped masses of 3 dimensional degree of freedom, and the response analysis is carried out using floor respone spectra
Safeguards Guidance for Designers of Commercial Nuclear Facilities – International Safeguards Requirements for Uranium Enrichment Plants

Energy Technology Data Exchange (ETDEWEB)

Philip Casey Durst; Scott DeMuth; Brent McGinnis; Michael Whitaker; James Morgan

2010-04-01

For the past two years, the United States National Nuclear Security Administration, Office of International Regimes and Agreements (NA-243), has sponsored the Safeguards-by-Design Project, through which it is hoped new nuclear facilities will be designed and constructed worldwide more amenable to nuclear safeguards. In the course of this project it was recognized that commercial designer/builders of nuclear facilities are not always aware of, or understand, the relevant domestic and international safeguards requirements, especially the latter as implemented by the International Atomic Energy Agency (IAEA). To help commercial designer/builders better understand these requirements, a report was prepared by the Safeguards-by-Design Project Team that articulated and interpreted the international nuclear safeguards requirements for the initial case of uranium enrichment plants. The following paper summarizes the subject report, the specific requirements, where they originate, and the implications for design and construction. It also briefly summarizes the established best design and operating practices that designer/builder/operators have implemented for currently meeting these requirements. In preparing the subject report, it is recognized that the best practices are continually evolving as the designer/builder/operators and IAEA consider even more effective and efficient means for meeting the safeguards requirements and objectives.
Steam generator design requirements for ACR-1000

International Nuclear Information System (INIS)

Subash, S.; Hau, K.

2006-01-01

Atomic Energy of Canada Limited (AECL) has developed the ACR-1000 (Advanced CANDU Reactor-1000 ) to meet market expectations for enhanced safety of plant operation, high capacity factor, low operating cost, increased operating life, simple component replacement, reduced capital cost, and shorter construction schedule. The ACR-1000 design is based on the use of horizontal fuel channels surrounded by a heavy water moderator, the same feature as in all CANDU reactors. The major innovation in the ACR-1000 is the use of low enriched uranium fuel, and light water as the coolant, which circulates in the fuel channels. This results in a compact reactor core design and a reduction of heavy water inventory, both contributing to a significant decrease in capital cost per MWe produced. The ACR-1000 plant is a two-unit, integrated plant with each unit having a nominal gross output of about 1165 MWe with a net output of approximately 1085 MWe. The plant design is adaptable to a single unit configuration, if required. This paper focuses on the technical considerations that went into developing some of the important design requirements for the steam generators for the ACR-1000 plant and how these requirements are specified in the Technical Specification, which is the governing document for the steam generator (SG) detail design. Layout of these SGs in the plant is briefly described and their impacts on the SG design. (author)
Quality assurance in the design of nuclear power plants

International Nuclear Information System (INIS)

1981-01-01

This Safety Guide provides the requirements and recommendations related to the establishment and implementation of quality assurance for design of items for a nuclear power plant. The requirements of this Guide shall be applied to the extent necessary during all constituent activities of the nuclear power plant project, such as design, manufacture, construction, commissioning and operations. Its requirements and recommendations shall be implemented, as appropriate, by the responsible organization or by its designated representatives: by plant designers, architect-engineers or manufacturers, when involved in performing design activities related to items to be manufactured; by site constructors, when involved in field engineering activities; by plant operators and other organizations, when involved in design activities related to plant modifications or to selection of spare or replacement parts; and by design consultants and other technical organizations, when performing any engineering activity that affects the work of other design organizations during various stages of nuclear power plant projects
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
Current USAEC seismic requirements for nuclear power plants

International Nuclear Information System (INIS)

Mehta, D.S.

1975-01-01

The principal seismic and geologic considerations which guide the USAEC in its evaluation of the suitability of proposed sites for nuclear power plants and plant design bases are set forth as design criteria in the AEC regulatory guides. The basic requirements of seismic design and analysis for seismic Category I structures, components, and systems important to public safety have been established in the USAEC regulatory guides and Code of Federal Regulations. It is pointed out that the current state-of-art techniques, best available technology, and additional studies in the field of earthquake engineering can be utilized to resolve seismic concerns. The seismic design requirements for nuclear plants to withstand postulated earthquakes can be standardized and this will be a significant milestone in the continuation of the Nuclear Standardization Program. (author)
Designing nuclear power plants for improved operation and maintenance

International Nuclear Information System (INIS)

1996-09-01

The purpose of this publication is to compile demonstrated, experience based design guidelines for improving the operability and maintainability of nuclear power plants. The guidelines are for use principally in the design of new nuclear power plants, but should also be useful in upgrading existing designs. The guidelines derive from the experience of operating and maintaining existing nuclear power plants as well as from the design of recent plants. In particular these guidelines are based on and consistent with both the EPRI advanced Light Water Reactor Utility Requirements Document, Volume 1, and the European Utility Requirements for LWR Nuclear Power Plants. 6 refs, 1 fig
Designing nuclear power plants for improved operation and maintenance

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-09-01

The purpose of this publication is to compile demonstrated, experience based design guidelines for improving the operability and maintainability of nuclear power plants. The guidelines are for use principally in the design of new nuclear power plants, but should also be useful in upgrading existing designs. The guidelines derive from the experience of operating and maintaining existing nuclear power plants as well as from the design of recent plants. In particular these guidelines are based on and consistent with both the EPRI advanced Light Water Reactor Utility Requirements Document, Volume 1, and the European Utility Requirements for LWR Nuclear Power Plants. 6 refs, 1 fig.
Plant design and layout of the different buildings with respect to safety, operational and maintenance requirements

International Nuclear Information System (INIS)

Linder, C.

1981-01-01

Design and layout of the buildings of a nuclear power plant are governed by the safety requirements regarding nearby population as called for by government regulations as well as by operational and maintenance requirements called for by the power utilities in order to assure smooth operation and easy service conditions. The lecture will focus on the different functional circumstances to be considered, their relative importance, criteria to be applied, pertinent regulations etc. and also give examples on the solutions to the above requirements. Main topics to be covered will be those circumstances that impose the highest demands on the civil engineering layout and design: airplane impact, earthquake, loss of coolant accident, pipe whipping, fuel cask transfer, annual overhaul, leak detection etc. (orig./RW)
Comparison of implementation of selected TMI action plan requirements on operating plants designed by Babcock and Wilcox

International Nuclear Information System (INIS)

Thoma, J.O.

1984-05-01

This report provides the results of a study conducted by the US Nuclear Regulatory Commission staff to compare the degree to which eight Babcock and Wilcox (B and W) designed licensed nuclear power plants have complied with the requirements in NUREG-0737, Clarification of TMI Action Plan Requirements. The eight licensed operating plants examined are as follows: Arkansas Nuclear One Unit 1 (ANO-1), Crystal River Unit 3, Davis Besse, Oconee Units 1, 2, and 3, Rancho Seco, and Three Mile Island Unit 1 (TMI-1). The purpose of this audit was to establish the progress of the TMI-1 licensee, General Public Utilities (GPU) Nuclear Corporation, in completing the long-term requirements in NUREG-0737 relative to the other B and W licensees examined
Fire protection requirements of the insurance industry and their impact on nuclear power plant design and construction

International Nuclear Information System (INIS)

Deitchman, J.V.; King, W.T. Jr.; Nashman, T.A.

1976-01-01

The insurance industry, with its wealth of knowledge and experience in the fire protection area and with preservation of its funds at stake, has always been heavily involved in the fire protection programs of nuclear power plants. Since it was concerned with property preservation in addition to nuclear safety, the insurance industry placed more detailed emphasis on fire protection requirements than did the nuclear regulatory bodies. Since the Browns Ferry fire, however, the insurance industry, the Nuclear Regulatory Commission, the Advisory Committee on Reactor Safeguards and the utilities themselves have re-examined their approaches to fire protection. A more coordinated approach seems to have emerged, which is based largely upon insurance industry specifications and guidelines. The paper briefly summarizes the fire protection requirements of the insurance industry as they apply to nuclear power plants. Some of the ways these requirements affect project planning, plant design, and construction timing are reviewed, as well as some of the more controversial fire protection areas
Current trends in codal requirements for safety in operation of nuclear power plants

International Nuclear Information System (INIS)

Srivasista, K.; Shah, Y.K.; Gupta, S.K.

2006-01-01

The Code of practice on safety in nuclear power plant operation states the requirements to be met during operation of a nuclear power plant for assuring safety. Among various stages of authorization, regulatory body issues authorization for operation of a nuclear power plant, monitors and enforces regulatory requirements. The responsible organization shall have overall responsibility and the plant management shall have the primary responsibility for ensuring safe and efficient operation of its nuclear power plants. A set of codal requirements covering technical and administrative aspects are mandatory for the plant management to implement to ensure that the nuclear power plant is operated in accordance with the design intent. Requirements on operating procedures and instructions establish operation and maintenance, inspection and testing of the plant in a planned and systematic way. The requirements on emergency preparedness programme establish with a reasonable assurance that, in the event of an emergency situation, appropriate measures can be taken to mitigate the consequences. Commissioning requirements verify performance criteria during commissioning to ensure that the design intent and QA requirements are met. Several modifications in systems important to safety required during operation of a nuclear power plant are regulated. However new operational codal requirements arising out of periodic safety review, operational experience feedback, life management, probabilistic safety assessment, physical security, safety convention and obligations and decommissioning are not covered in the present code of practice for safety in nuclear power plant operation. Codal provisions on 'Review by operating organization on aspects of design having implications on operability' are also required to be addressed. The merits in developing such a methodology include acceptance of the design by operating organization, ensuring maintainability, proper layout etc. in the new designs
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)
Test requirements for the integral effect test to simulate Korean PWR plants

Energy Technology Data Exchange (ETDEWEB)

Song, Chul Hwa; Park, C. K.; Lee, S. J.; Kwon, T. S.; Yun, B. J.; Chung, M. K

2001-02-01

In this report, the test requirements are described for the design of the integral effect test facility to simulate Korean PWR plants. Since the integral effect test facility should be designed so as to simulate various thermal hydraulic phenomena, as closely as possible, to be occurred in real plants during operation or anticipated transients, the design and operational characteristics of the reference plants (Korean Standard Nuclear Plant and Korean Next Generation Reactor)were analyzed in order to draw major components, systems, and functions to be satisfied or simulated in the test facility. The test matrix is set up by considering major safety concerns of interest and the test objectives to confirm and enhance the safety of the plants. And the analysis and prioritization of the test matrix leads to the general design requirements of the test facility. Based on the general design requirements, the design criteria is set up for the basic and detailed design of the test facility. And finally it is drawn the design requirements specific to the fluid system and measurement system of the test facility. The test requirements in this report will be used as a guideline to the scaling analysis and basic design of the test facility. The test matrix specified in this report can be modified in the stage of main testing by considering the needs of experiments and circumstances at that time.
Test requirements for the integral effect test to simulate Korean PWR plants

International Nuclear Information System (INIS)

Song, Chul Hwa; Park, C. K.; Lee, S. J.; Kwon, T. S.; Yun, B. J.; Chung, M. K.

2001-02-01

In this report, the test requirements are described for the design of the integral effect test facility to simulate Korean PWR plants. Since the integral effect test facility should be designed so as to simulate various thermal hydraulic phenomena, as closely as possible, to be occurred in real plants during operation or anticipated transients, the design and operational characteristics of the reference plants (Korean Standard Nuclear Plant and Korean Next Generation Reactor)were analyzed in order to draw major components, systems, and functions to be satisfied or simulated in the test facility. The test matrix is set up by considering major safety concerns of interest and the test objectives to confirm and enhance the safety of the plants. And the analysis and prioritization of the test matrix leads to the general design requirements of the test facility. Based on the general design requirements, the design criteria is set up for the basic and detailed design of the test facility. And finally it is drawn the design requirements specific to the fluid system and measurement system of the test facility. The test requirements in this report will be used as a guideline to the scaling analysis and basic design of the test facility. The test matrix specified in this report can be modified in the stage of main testing by considering the needs of experiments and circumstances at that time
Neutronics requirements for a DEMO fusion power plant

Energy Technology Data Exchange (ETDEWEB)

Fischer, U., E-mail: ulrich.fischer@kit.edu [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Bachmann, C. [EUROfusion Consortium , Boltzmannstraße 2, 85748 Garching (Germany); Palermo, I. [Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT), Madrid (Spain); Pereslavtsev, P. [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Villari, R. [ENEA UT-FUS C.R. Frascati, Via E. Fermi 45, 00044 Frascati (Italy)

2015-10-15

Highlights: • Discussion and specification of neutronic requirements for a DEMO power plant. • TBR uncertainties are reviewed/discussed and design margins are elaborated. • Limits are given for radiation loads to super-conducting magnets and steel structural components. • Available DEMO results are compared to recommended limits and TBR design target. - Abstract: This paper addresses the neutronic requirements a DEMO fusion power plant needs to fulfil for a reliable and safe operation. The major requirement is to ensure Tritium self-sufficiency taking into account the various uncertainties and plant-internal losses that occur during DEMO operation. A further major requirement is to ensure sufficient protection of the superconducting magnets against the radiation penetrating in-vessel components and vessel. Reliable criteria for the radiation loads need to be defined and verified to ensure the reliable operation of the magnets over the lifetime of DEMO. Other issues include radiation induced effects on structural materials such as the accumulated displacement damage, the generation of gases such as helium which may deteriorate the material performance. The paper discusses these issues and their impact on design options for DEMO taking into account results obtained in the frame of European Power Plant Physics and Technology (PPPT) 2013 programme activities with DEMO models employing the helium cooled pebble bed (HCPB), the helium cooled lithium lead (HCLL), and the water-cooled (WCLL) blanket concepts.

System 80+ integrated design of a complete plant

International Nuclear Information System (INIS)

Turk, R.S.; Stamm, S.L.; Fox, W.A.

1992-01-01

In 1985, ABB-Combustion Engineering Nuclear Power (ABB-CENP) and elements of Duke Power Company [now Duke Engineering ampersand Services (DE ampersand S)] joined forces under the aegis of the Electric Power Research Institute (EPRI) Advanced Light Water Reactor (ALWR) Program to develop, with the sponsoring utilities, the design requirements for the next generation of nuclear power plants. With support from the US Department of Energy, ABB-CENP and DE ampersand S again teamed up the following year to initiate a project to design and license the System 80+ standard plant design, an advanced pressurized water reactor that meets these utility requirements. A distinguishing feature of the System 80+ standard design is that it is an essentially complete plant, predesigned and prelicensed to ensure rapid and economical construction. This is in stark contrast to typical prior conduct, where the reactor vendor offered only the nuclear steam supply system and the plant was built on a design-as-you-go basis with constant pressure to release individual elements of the plant design for construction or procurement as soon as possible. Now, however, the design process can be integrated over the total plant, ensuring that the goals set for ALWRs can be met. This integrated design process is manifested in several ways: (1) broad-based participation during the design process by involving designers, analysts, suppliers, constructors, and operators; (2) use of probabilistic risk assessment (PRA) as a design tool to aid in evaluating design features on a total-plant basis; (3) application of human factors engineering methods to a total plant distributed control system to improve the human-machine interface in the design; and (4) use of computer-aided design to enhance assessment of interactions and impacts of all aspects of the total plant. Each of these aspects of integrated plant design is discussed in this paper
An approach to nuclear plant design and modification support for Russian-designed plants in Eastern Europe

International Nuclear Information System (INIS)

Ioannidi, J.; Akins, M.J.

2002-01-01

The Western nuclear countries have embarked on numerous programs to improve the safety of the Russian-designed nuclear power plants. In Russian-designed plants in Eastern Europe, plant management is being asked for the first time to decide which safety projects to implement and is finding itself lacking in nuclear safety analytical tools and practices, funds, and experience with project management and project engineering skills and tools. Some of the major areas where assistance is needed are: 1) Defining plant weaknesses toward nuclear safety. 2) Evaluating and grading the importance to safety of proposed modification. 3) Project Planning and Scheduling using computer based scheduling software. 4) Project Finance Development and Management using well defined cash flow management techniques. 5) Contract Management and Change Control. 6) Interface Management. Each of these areas requires a significant amount of discussion to understand the issues and problems associated with them. However, this paper is limited to the Project Management areas. This paper encourages the use of a design engineering firm experienced in safety practices and associated management and technical skills to serve as the Owner's Engineer/Project Management Consultant for the program period for a Russian-designed plants located outside Russia. This approach would allow for the availability and transfer of knowledge of safety practices to plant personnel and owners engineers at nuclear plants outside Russia, improving their nuclear safety culture. The plant personnel would control plant modernizations and upgrades based upon a proven and well-defined process for detailed project definition, configuration change control, and project management. This offers the opportunity to enhance the long-term safety culture by developing plant personnel knowledgeable of the safety practices, plant design basis, developing a modification control process enabling them to control the design basis through future
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
Considerations for Applying Design Extension Conditions to Domestic Nuclear Power Plants

International Nuclear Information System (INIS)

Ryu, Yongho

2013-01-01

The concept is designed to include more serious accidents than the existing design basis accidents considering additional failures. Design extension conditions can be derived based on engineering judgments, deterministic analysis or probabilistic analysis of the nuclear power plants. They are used to secure practical response capabilities to prevent or mitigate accidents. They may also require the deployment of additional safety equipment for existing nuclear power plants currently in operation. Though the general requirements of design extension conditions are described under the IAEA standards, no specific guidelines have been presented as required for their actual application to the nuclear power plant design. Furthermore, there is great variation between countries in implementing the requirements of design extension conditions. Therefore, for the actual application, considerable effort should be made among relevant organizations to establish detailed requirements of the design extension conditions. Such activities could constitute a part of the efforts of the nuclear community to meet the general public's expectations concerning the safety of nuclear power plants. The introduction of design extension conditions is expected to be a means of systematically enhancing the safety of nuclear power plants. Yet, there exists great differences in terms of the scope of analysis and the acceptance criteria, as no uniform practices have yet been established in applying the specific requirements for design extension conditions. A careful review is required in terms of the technical basis for setting the requirements, including those pertaining to the scope of analysis and the acceptance criteria. The introduction of these new requirements to Korean nuclear power plants may cause unexpected problems. Therefore, it is desirable for the regulatory agency to systematically assess the impact of design extension conditions and to discuss the arising issues with the stake holder
New design system for nuclear power plant

International Nuclear Information System (INIS)

Kakuta, Masataka; Yoshinaga, Toshiaki; Yoshida, Ikuzo; Tokumasu, Shinji.

1980-01-01

As for the machine and equipment layout and the piping design for nuclear power plants, the multilateral coordination and study on such factors as functions, installation, radiation exposure and maintenance are required, and the high reliability is demanded. On the other hand, the quantity of things handled is enormous, therefore it is difficult to satisfy completely the above described requirements and to make plant planning which is completely free from the mutual interference of machines, equipments and pipings by the ordinary design with drawings only. Thereupon, the following new device was adopted to the design method for the purposes of improving the quality and shortening the construction period. Namely at the time of designing new plants, the rationalization of plant planning method was attempted by introducing color composite drawings and the technique of model engineering, at the same time, the newly developed design system for pipings was applied with a computer, thus the large accomplishment was able to be obtained regarding the improvement of reliability and others by making the check-up of the propriety. The design procedures of layout and piping, the layout design and general coordination in nuclear power stations with models and color composite drawings and the design system are explained. (Kako, I.)
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.)
The System 80+ Standard Plant design control document. Volume 11

International Nuclear Information System (INIS)

1997-01-01

This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume covers parts 6 and 7 and appendix 7A for section 7 (Instrumentation and Control) of the ADM Design and Analysis. The topics covered by these are: other systems required for safety; control systems not required by safety; and CMF evaluation of limiting faults. Parts 1--3 of section 8 (Electric Power) of the ADM are also included in this volume. Topics covered by these parts are: introduction; offsite power system; and onsite power system
Nuclear power. Volume 1. Nuclear power plant design

International Nuclear Information System (INIS)

Pedersen, E.S.

1978-01-01

NUCLEAR POWER PLANT DESIGN is intended to be used as a working reference book for management, engineers and designers, and as a graduate-level text for engineering students. The book is designed to combine theory with practical nuclear power engineering and design experience, and to give the reader an up-to-date view of the status of nuclear power and a basic understanding of how nuclear power plants function. Volume 1 contains the following chapters; (1) nuclear reactor theory; (2) nuclear reactor design; (3) types of nuclear power plants; (4) licensing requirements; (5) shielding and personnel exposure; (6) containment and structural design; (7) main steam and turbine cycles; (8) plant electrical system; (9) plant instrumentation and control systems; (10) radioactive waste disposal (waste management) and (11) conclusion
Plant aging and design bases documentation

International Nuclear Information System (INIS)

Kelly, J.

1985-01-01

As interest in plant aging and lifetime extension continues to grow, the need to identify and capture the original design bases for the plant becomes more urgent. Decisions on lifetime extension and availability must be based on a rational understanding of design input, assumptions, and objectives. As operating plant time accumulates, the history of the early design begins to fade. The longer the utility waits, the harder it will be to re-establish the original design bases. Therefore, the time to develop this foundation is now. This paper demonstrates the impact that collecting and maintaining the original design bases of the plant can have on a utility's lifetime extension program. This impact becomes apparent when considering the technical, regulatory and financial aspects of lifetime extension. It is not good enough to know that the design information is buried somewhere in the corporate archives, and that given enough time, it could be retrieved. To be useful to the lifetime extension program, plant design information must be concise, readily available (i.e., retrievable), and easy to use. These objectives can only be met through a systematic program for collecting and presenting plant design documentation. To get the maximum benefit from a lifetime extension program, usable design bases documentation should be available as early in the plant life as possible. It will help identify areas that require monitoring today so that data is available to make rational decisions in the future
A direct methodology to establish design requirements for human–system interface (HSI) of automatic systems in nuclear power plants

International Nuclear Information System (INIS)

Anuar, Nuraslinda; Kim, Jonghyun

2014-01-01

Highlights: • A systematic method to identify the design requirements for human–system interface is proposed. • Eight combinations of control agents in each control stage (levels of automation) are defined. • The use of Itemized Sequence Diagram (ISD) is discussed for task allocation to control agents. • The design requirements of human–system interface are established based on the produced ISD. - Abstract: This paper suggests a systematic approach to establish design requirements for the human–system interface (HSI) between operators and automatic systems. The role of automation in the control of a nuclear power plant (NPP) operation is to support the human operator and act as an efficient team player to help reduce the human operator’s workload. Some of the problems related to the interaction between the human operator and automation are out-of-the-loop performance, mode errors, role change to supervisory role and final authority issues. Therefore, the design of HSI is critical to avoiding breakdowns in communication between the human operator and the system. In this paper, the design requirements for human–system interface of automatic systems are constructed with the help of a tool called Itemized Sequence Diagram (ISD). Eight levels of automation (LOA) are initially defined in the function allocation and an ISD is drawn for each of the LOA for task allocation. The ISD is a modified version of sequence diagram, which is widely used in systems engineering as well as software engineering. The ISD elements of arrows, messages, actors and alternative boxes collectively show the interactions between the control agents, which are decomposed into four different roles: information acquiring, plant diagnosing, response selecting and response implementing. Eleven design requirements to optimize the human–automation interaction are suggested by using this method. The design requirements produced from the identified interaction points in the ISD are
Engineering Design of ITER Prototype Fast Plant System Controller

Science.gov (United States)

Goncalves, B.; Sousa, J.; Carvalho, B.; Rodrigues, A. P.; Correia, M.; Batista, A.; Vega, J.; Ruiz, M.; Lopez, J. M.; Rojo, R. Castro; Wallander, A.; Utzel, N.; Neto, A.; Alves, D.; Valcarcel, D.

2011-08-01

The ITER control, data access and communication (CODAC) design team identified the need for two types of plant systems. A slow control plant system is based on industrial automation technology with maximum sampling rates below 100 Hz, and a fast control plant system is based on embedded technology with higher sampling rates and more stringent real-time requirements than that required for slow controllers. The latter is applicable to diagnostics and plant systems in closed-control loops whose cycle times are below 1 ms. Fast controllers will be dedicated industrial controllers with the ability to supervise other fast and/or slow controllers, interface to actuators and sensors and, if necessary, high performance networks. Two prototypes of a fast plant system controller specialized for data acquisition and constrained by ITER technological choices are being built using two different form factors. This prototyping activity contributes to the Plant Control Design Handbook effort of standardization, specifically regarding fast controller characteristics. Envisaging a general purpose fast controller design, diagnostic use cases with specific requirements were analyzed and will be presented along with the interface with CODAC and sensors. The requirements and constraints that real-time plasma control imposes on the design were also taken into consideration. Functional specifications and technology neutral architecture, together with its implications on the engineering design, were considered. The detailed engineering design compliant with ITER standards was performed and will be discussed in detail. Emphasis will be given to the integration of the controller in the standard CODAC environment. Requirements for the EPICS IOC providing the interface to the outside world, the prototype decisions on form factor, real-time operating system, and high-performance networks will also be discussed, as well as the requirements for data streaming to CODAC for visualization and
The System 80+ Standard Plant design control document. Volume 19

International Nuclear Information System (INIS)

1997-01-01

This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume contains five technical specification bases that are part of Appendix 16 A of the ADM Design and Analysis. They are: TS B3.3 Instrumentation Bases; TS B3.4 RCS Bases; TS B3.5 ECCS Bases; TS B3.6 Containment Systems Bases; and TS B3.7 Plant Systems Bases
The System 80+ Standard Plant design control document. Volume 2

International Nuclear Information System (INIS)

1997-01-01

This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume covers the following information of the CDM: (2.8) Steam and power conversion; (2.9) Radioactive waste management; (2.10) Tech Support Center; (2.11) Initial test program; (2.12) Human factors; and sections 3, 4, and 5. Also covered in this volume are parts 1--6 of section 1 (General Plant Description) of the ADM Design and Analysis
Considering plant life management influences on new plant design

International Nuclear Information System (INIS)

Dam, R.F.; Choy, E.; Soulard, M.; Nickerson, J.H.; Hopwood, J.

2003-01-01

After operating successfully for more than half their design life, owners of CANDU reactors are now engaging in Plant Life Management (PLiM) activities to ensure not only life attainment, but also life extension. For several years, Atomic Energy of Canada Ltd. (AECL) has been working with domestic and offshore CANDU utilities on a comprehensive and integrated CANDU PLiM program that will see existing CANDU plants successfully and reliably operate through their design life and beyond. To support the PLiM program development, a significant level of infrastructure has been, and continues to be, developed at AECL. This includes the development of databases that document relevant knowledge and background to allow for a more accessible and complete understanding of degradation issues and the strategies needed to deal with these issues. As the level of integration with various project, services and R and D activities in AECL increases, this infrastructure is growing to encompass a wider range of design, operations and maintenance details to support comprehensive and quantitative assessment of CANDU stations. With the maturation of the PLiM program, these processes were adapted for application to newer plants. In particular, a fully integrated program was developed that interrelates the design basis, operations, safety, and reliability and maintenance strategies, as applied to meet plant design goals. This has led to the development of the maintenance-based design concept. The various PLiM technologies, developed and applied in the above programs with operating stations, are being modified and tailored to assist with the new plant design processes to assure that ACR- Advanced CANDU Reactor meets its targets for operation, maintenance, and lifetime performance. Currently, the ACR, developed by Atomic Energy of Canada Ltd. (AECL), is being designed with features to increase capacity factors, to reduce the risk of major equipment failures, to improve access to key components
The System 80+ Standard Plant design control document. Volume 10

International Nuclear Information System (INIS)

1997-01-01

This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume contains Appendices 6A, 6B, and 6C for section 6 (Engineered Safety Features) of the ADM Design and Analysis. Also, parts 1--5 of section 7 (Instrumentation and Control) of the ADM Design and Analysis are covered. The following information is covered in these parts: introduction; reactor protection system; ESF actuation system; system required for safe shutdown; and safety-related display instrumentation
An integrated translation of design data of a nuclear power plant from a specification-driven plant design system to neutral model data

International Nuclear Information System (INIS)

Mun, Duhwan; Yang, Jeongsam

2010-01-01

How to efficiently integrate and manage lifecycle data of a nuclear power plant has gradually become an important object of study. Because plants usually have a very long period of operation and maintenance, the plant design data need to be presented in a computer-interpretable form and to be independent of any commercial systems. The conversion of plant design data from various design systems into neutral model data is therefore an important technology for the effective operation and maintenance of plants. In this study, a neutral model for the efficient integration of plant design data is chosen from among the currently available options and extended in order to cover the information model requirements of nuclear power plants in Korea. After the mapping of the neutral model and the data model of a specification-driven plant design system, a plant data translator is also implemented in accordance with the schema mapping results.
An integrated translation of design data of a nuclear power plant from a specification-driven plant design system to neutral model data

Energy Technology Data Exchange (ETDEWEB)

Mun, Duhwan, E-mail: dhmun@moeri.re.k [Marine Safety and Pollution Response Research Department, Maritime and Ocean Engineering Research Institute, KORDI, 171 Jang-dong, Yuseong-gu, Daejeon 305-343 (Korea, Republic of); Yang, Jeongsam, E-mail: jyang@ajou.ac.k [Division of Industrial and Information Systems Engineering, Ajou University, San 5, Wonchun-dong, Yeongtong-gu, Suwon 443-749 (Korea, Republic of)

2010-03-15

How to efficiently integrate and manage lifecycle data of a nuclear power plant has gradually become an important object of study. Because plants usually have a very long period of operation and maintenance, the plant design data need to be presented in a computer-interpretable form and to be independent of any commercial systems. The conversion of plant design data from various design systems into neutral model data is therefore an important technology for the effective operation and maintenance of plants. In this study, a neutral model for the efficient integration of plant design data is chosen from among the currently available options and extended in order to cover the information model requirements of nuclear power plants in Korea. After the mapping of the neutral model and the data model of a specification-driven plant design system, a plant data translator is also implemented in accordance with the schema mapping results.
Study on design method for seismically isolated FBR plants

International Nuclear Information System (INIS)

Hirata, Kazuta; Yabana, Shuichi; Ohtori, Yasuki; Ishida, Katsuhiko; Sawada, Yoshihiro; Shiojiri; Hiroo; Mazda, Taiji

1998-01-01

CRIEPI conducted 'Demonstration test on FBR seismic isolation system' from 1987 to 1996 under contract with Ministry of International Trade and Industry, Japan. In the demonstration test, base isolation technologies are prepared and demonstrated to apply to FBR and the design guidelines are proposed. In this report overall contents of the design guidelines entitled Design guidelines for seismically base isolated FBR plants' are included. The design guidelines, as a rule, are limited to apply to FBR plants where entire reactor building is isolated in the horizontal direction using laminated rubber bearings as isolators. The design guidelines and its concepts, however, will be useful for the development of similar guidelines for other isolation systems using different type of isolation methods and other nuclear facilities. The design guidelines consist of three parts and appendices. The first part is 'Policy for Safety Design of Base Isolated FBR Plants' specifying the principles and the requirements in the planning and the design for the safety of base isolated FBR plants. The second part is Policy for Seismic Design of Base Isolated FBR' describing the principles and the requirements in the seismic design and the evaluation of safety for base isolated FBR plants. The third part is 'Design Methods for Seismic Isolated FBR Plants' detailing the methods, procedures and parameters to be used in the design and the evaluation of safety fro base isolated FBR plants. In appendices examples of design procedures for base isolated reactor building and laminated rubber bearings as well as various test data on laminated rubber bearings, etc. are shown. (author)
Regulatory requirements on the design and construction of nuclear power plant control and instrumentation systems in Finland

International Nuclear Information System (INIS)

Heikkila, M.A.

1978-01-01

The Department of Reactor Safety of the Institute of Radiation Protection, being the nuclear regulatory authority in Finland, has set up regulations which govern the design and construction of NPP systems and components. The regulations are partly compiled from existing codes and standards, published primarily in the United States and Federal Republic of Germany, and partly worked out at the Institute. The regulations are collected to a special set of YVL guides (guides for nuclear power plants), and one of these gives requirements on the design and construction of NPPCI systems and components. The scope of the requirements is based on the safety classification of the CI systems and components. Three safety classes have been singled out: the first for CI systems which take part in reactor protection, the second for other directly safety related, and the third for remaining CI systems important enough to deserve supervision. The safety class for CI components is inherited from the system they belong to. The safety classification of IC systems has direct bearing on the initial assumptions of plant accident analysis. The design principles of IC systems are inspected as part of the preliminary and final safety reports. Focus is directed on the principles of redundancy, separation, diversity, testability, etc. The requirements on IC components are directed to different stages of manufacture, installation and operation. The type tests shall be adequate and acceptably documented. The manufacture of components is followed, the test reports reviewed and the efficiency of manufacturers quality assurance program evaluated. Further requirements concern the installation phase and tests at the end of it, and finally guides include directions for maintenance and testing during the operations phase. (author)
Safeguards Guidance Document for Designers of Commercial Nuclear Facilities: International Nuclear Safeguards Requirements and Practices For Uranium Enrichment Plants

Energy Technology Data Exchange (ETDEWEB)

Robert Bean; Casey Durst

2009-10-01

This report is the second in a series of guidelines on international safeguards requirements and practices, prepared expressly for the designers of nuclear facilities. The first document in this series is the description of generic international nuclear safeguards requirements pertaining to all types of facilities. These requirements should be understood and considered at the earliest stages of facility design as part of a new process called “Safeguards-by-Design.” This will help eliminate the costly retrofit of facilities that has occurred in the past to accommodate nuclear safeguards verification activities. The following summarizes the requirements for international nuclear safeguards implementation at enrichment plants, prepared under the Safeguards by Design project, and funded by the U.S. Department of Energy (DOE) National Nuclear Security Administration (NNSA), Office of NA-243. The purpose of this is to provide designers of nuclear facilities around the world with a simplified set of design requirements and the most common practices for meeting them. The foundation for these requirements is the international safeguards agreement between the country and the International Atomic Energy Agency (IAEA), pursuant to the Treaty on the Non-proliferation of Nuclear Weapons (NPT). Relevant safeguards requirements are also cited from the Safeguards Criteria for inspecting enrichment plants, found in the IAEA Safeguards Manual, Part SMC-8. IAEA definitions and terms are based on the IAEA Safeguards Glossary, published in 2002. The most current specification for safeguards measurement accuracy is found in the IAEA document STR-327, “International Target Values 2000 for Measurement Uncertainties in Safeguarding Nuclear Materials,” published in 2001. For this guide to be easier for the designer to use, the requirements have been restated in plainer language per expert interpretation using the source documents noted. The safeguards agreement is fundamentally a

The research for the design verification of nuclear power plant based on VR dynamic plant

International Nuclear Information System (INIS)

Wang Yong; Yu Xiao

2015-01-01

This paper studies a new method of design verification through the VR plant, in order to perform verification and validation the design of plant conform to the requirements of accident emergency. The VR dynamic plant is established by 3D design model and digital maps that composed of GIS system and indoor maps, and driven by the analyze data of design analyzer. The VR plant could present the operation conditions and accident conditions of power plant. This paper simulates the execution of accident procedures, the development of accidents, the evacuation planning of people and so on, based on VR dynamic plant, and ensure that the plant design will not cause bad effect. Besides design verification, simulated result also can be used for optimization of the accident emergency plan, the training of accident plan and emergency accident treatment. (author)
B plant standards/requirements identification document (S/RID)

Energy Technology Data Exchange (ETDEWEB)

Maddox, B.S., Westinghouse Hanford

1996-07-29

This Standards/Requirements Identification Document (S/RID) set forth the Environmental Safety and Health (ES{ampersand}H) standards/requirements for the B Plant. This S/RID is applicable to the appropriate life cycle phases of design, construction,operation, and preparation for decommissioning. These standards/requirements are adequate to ensure the protection of the health and safety of workers, the public, and the environment.
Safety of nuclear power plants: Operation. Safety requirements

International Nuclear Information System (INIS)

2004-01-01

The safety of a nuclear power plant is ensured by means of its proper siting, design, construction and commissioning, followed by the proper management and operation of the plant. In a later phase, proper decommissioning is required. This Safety Requirements publication supersedes the Code on the Safety of Nuclear Power Plants: Operation, which was issued in 1988 as Safety Series No. 50-C-O (Rev. 1). The purpose of this revision was: to restructure Safety Series No. 50-C-O (Rev. 1) in the light of the basic objectives, concepts and principles in the Safety Fundamentals publication The Safety of Nuclear Installations. To be consistent with the requirements of the International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources. And to reflect current practice and new concepts and technical developments. Guidance on fulfillment of these Safety Requirements may be found in the appropriate Safety Guides relating to plant operation. The objective of this publication is to establish the requirements which, in the light of experience and the present state of technology, must be satisfied to ensure the safe operation of nuclear power plants. These requirements are governed by the basic objectives, concepts and principles that are presented in the Safety Fundamentals publication The Safety of Nuclear Installations. This publication deals with matters specific to the safe operation of land based stationary thermal neutron nuclear power plants, and also covers their commissioning and subsequent decommissioning
Safety of nuclear power plants: Operation. Safety requirements

International Nuclear Information System (INIS)

2003-01-01

The safety of a nuclear power plant is ensured by means of its proper siting, design, construction and commissioning, followed by the proper management and operation of the plant. In a later phase, proper decommissioning is required. This Safety Requirements publication supersedes the Code on the Safety of Nuclear Power Plants: Operation, which was issued in 1988 as Safety Series No. 50-C-O (Rev. 1). The purpose of this revision was: to restructure Safety Series No. 50-C-O (Rev. 1) in the light of the basic objectives, concepts and principles in the Safety Fundamentals publication The Safety of Nuclear Installations. To be consistent with the requirements of the International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources. And to reflect current practice and new concepts and technical developments. Guidance on fulfillment of these Safety Requirements may be found in the appropriate Safety Guides relating to plant operation. The objective of this publication is to establish the requirements which, in the light of experience and the present state of technology, must be satisfied to ensure the safe operation of nuclear power plants. These requirements are governed by the basic objectives, concepts and principles that are presented in the Safety Fundamentals publication The Safety of Nuclear Installations. This publication deals with matters specific to the safe operation of land based stationary thermal neutron nuclear power plants, and also covers their commissioning and subsequent decommissioning
Safety of nuclear power plants: Operation. Safety requirements

International Nuclear Information System (INIS)

2000-01-01

The safety of a nuclear power plant is ensured by means of its proper siting, design, construction and commissioning, followed by the proper management and operation of the plant. In a later phase, proper decommissioning is required. This Safety Requirements publication supersedes the Code on the Safety of Nuclear Power Plants: Operation, which was issued in 1988 as Safety Series No. 50-C-O (Rev. 1). The purpose of this revision was: to restructure Safety Series No. 50-C-O (Rev. 1) in the light of the basic objectives, concepts and principles in the Safety Fundamentals publication The Safety of Nuclear Installations; to be consistent with the requirements of the International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources; and to reflect current practice and new concepts and technical developments. Guidance on fulfillment of these Safety Requirements may be found in the appropriate Safety Guides relating to plant operation. The objective of this publication is to establish the requirements which, in the light of experience and the present state of technology, must be satisfied to ensure the safe operation of nuclear power plants. These requirements are governed by the basic objectives, concepts and principles that are presented in the Safety Fundamentals publication The Safety of Nuclear Installations. This publication deals with matters specific to the safe operation of land based stationary thermal neutron nuclear power plants, and also covers their commissioning and subsequent decommissioning
The System 80+ Standard Plant design control document. Volume 1

International Nuclear Information System (INIS)

1997-01-01

This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume covers the DCD introduction and contains sections 1 and parts 1--7 of section 2 of the CDM. Parts 1--7 included the following: (2.1) Design of SSC; (2.2) Reactor; (2.3) RCS and connected systems; (2.4) Engineered Safety Features; (2.5) Instrumentation and Control; (2.6) Electric Power; and (2.7) Auxiliary Systems
The basic design and requirement for plant tissue culture laboratory in MINT

International Nuclear Information System (INIS)

Azraf Azman; Rosli Darmawan; Rusli Ibrahim; Mohd Nazir Basiran; Azhar Mohamad; Mohamed Najli Mohamed Yasin; Shuhaimi Shamsuddin

2005-01-01

The production of multiple species plantlets involves a relatively complex process and it is a highly specialized operation. Tissue culture technology is rapidly becoming a commercialized method for propagating new cultivars, rare species and difficult-to-propagate plant. Not only are skills and knowledge essential but the laboratory itself also plays an important role to ensure the successful growth of the plantlets. To produce quality plantlets, plant tissue culture laboratories should fulfill the basic requirements. The laboratory should have proper building and layout which comprise of media preparation and washing room, sterilization or autoclave room, transfer room and culture or growth room. The scope of this paper is to compare these fundamental requirements with the plant tissue culture laboratory in MINT. All the basic needs and differences will be discussed and the proposal for corrective actions will be presented. (Author)
Deriving human resource requirements for new nuclear plants

International Nuclear Information System (INIS)

Goodnight, Ch.T.

2007-01-01

For those contemplating the deployment of a new nuclear plant, critical issues include the development of the operational staff and its organization. This paper will discuss the key elements of deriving an appropriate staff plan for steady state operations. There are five areas that must be analyzed, and each area has unique requirements. These key areas are 1) Operations, 2) Engineering, 3) Maintenance, 4) Regulatory/Oversight, and 5) Site Support. After the analysis for each area is complete, and the human resource requirements are identified, an organizational structure must be developed to support the necessary management, potential centralization, and appropriate functional alignments for effective and safe plant operation. Four organizational design principals have been defined. The first design principal relates to the organizational structure: no more than 7 layers of management between an individual contributor and the senior nuclear manager in the NPP. The second design principal relates to the grouping of activities in order to ensure appropriate management of related or supporting activities. The third design principal relates to out-sourcing support activities when these activities comply with particular conditions for instance when they are not mission critical to day-to day plant operations. The fourth design principal relates to the centralization of some activities when there is more than one NPP operated by the same parent company
The System 80+ Standard Plant design control document. Volume 24

International Nuclear Information System (INIS)

1997-01-01

This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume contains sections 7--11 of the ADM Emergency Operations Guidelines. Topics covered are: excess steam demand recovery; loss of all feedwater; loss of offsite power; station blackout recovery; and functional recovery guideline. Appendix A Severe Accident Management Guidelines and Appendix B Lower Mode Operational Guidelines are also included
The System 80+ Standard Plant design control document. Volume 15

International Nuclear Information System (INIS)

1997-01-01

This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume contains all five parts of section 12 (Radiation Protection) of the ADM Design and Analysis. Topics covered are: ALARA exposures; radiation sources; radiation protection; dose assessment; and health physics program. All six parts and appendices A and B for section 13 (Conduct of Operations) of the ADM Design and Analysis are also contained in this volume. Topics covered are: organizational structure; training program; emergency planning; review and audit; plant procedures; industrial security; sabotage protection (App 13A); and vital equipment list (App 13B)
The System 80+ Standard Plant design control document. Volume 18

International Nuclear Information System (INIS)

1997-01-01

This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume contains the following technical specifications of section 16 (Technical Specifications) of the ADM Design and Analysis: TS 3.3 Instrumentation; TS 3.4 Reactor Coolant System; TS 3.5 Emergency Core Cooling System; TS 3.6 Containment Systems; TS 3.7 Plant Systems; TS 3.8 Electrical Power Systems; TS 3.9 Refueling Operations; TS 4.0 Design Features; TS 5.0 Administrative Controls. Appendix 16 A Tech Spec Bases is also included. It contains the following: TS B2.0 Safety Limits Bases; TS B3.0 LCO Applicability Bases; TS B3.1 Reactivity Control Bases; TS B3.2 Power Distribution Bases
Review of nuclear piping seismic design requirements

International Nuclear Information System (INIS)

Slagis, G.C.; Moore, S.E.

1994-01-01

Modern-day nuclear plant piping systems are designed with a large number of seismic supports and snubbers that may be detrimental to plant reliability. Experimental tests have demonstrated the inherent ruggedness of ductile steel piping for seismic loading. Present methods to predict seismic loads on piping are based on linear-elastic analysis methods with low damping. These methods overpredict the seismic response of ductile steel pipe. Section III of the ASME Boiler and Pressure Vessel Code stresses limits for piping systems that are based on considerations of static loads and hence are overly conservative. Appropriate stress limits for seismic loads on piping should be incorporated into the code to allow more flexible piping designs. The existing requirements and methods for seismic design of piping systems, including inherent conservations, are explained to provide a technical foundation for modifications to those requirements. 30 refs., 5 figs., 3 tabs
Recommended general safety requirements for nuclear power plants

International Nuclear Information System (INIS)

1983-06-01

This report presents recommendations for a set of general safety requirements that could form the basis for the licensing of nuclear power plants by the Atomic Energy Control Board. In addition to a number of recommended deterministic requirements the report includes criteria for the acceptability of the design of such plants based upon the calculated probability and consequence (in terms of predicted radiation dose to members of the public) of potential fault sequences. The report also contains a historical review of nuclear safety principles and practices in Canada
Control room design and human engineering in power plants

International Nuclear Information System (INIS)

Herbst, L.; Hinz, W.

1982-01-01

The concept for modern plant control rooms is primary influenced by: The automation of protection, binary control and closed loop control functions; organization employing functional areas; computer based information processing; human engineered design. Automation reduces the human work load. Employment of functional areas permits optimization of operational sequences. Computer based information processing makes it possible to output information in accordance with operating requirements. Design based on human engineering principles assures the quality of the interaction between the operator and the equipment. The degree to which these conceptional features play a role in design of power plant control rooms depends on the unit rating, the mode of operation and on the requirements respecting safety and availability of the plant. (orig.)
System 80+ Design and Licensing : Improving Plant Reliability

International Nuclear Information System (INIS)

Newman, Robert E.

1989-01-01

The U. S. nuclear industry is striving to improve plant reliability and availability through improved plant design, component designs and plant maintenance. In an effort to improve safety and to demonstrate that commercial nuclear power is economically competitive with other energy sources, the utilities, nuclear vendors, architect engineers and constructors, and component suppliers are all participating in an industry-wide effort to develop improved Light Water Reactor (LWR) designs that are based upon the many years of successful LWR operation. In an age when the world faces the environmental pressures of the greenhouse effect and acid rain, electricity generated from nuclear energy must play an increasing role in the energy picture of Korea, the United States and the rest of the world. This paper discusses the plant availability requirement that has been established by the industry-wide effort mentioned above. After briefly describing Combustion Engineering's program for development of the System 80 Plus standard design and the participation of the Korea Advanced Energy Research Institute (KAERI) in the program, the paper then describes the design features that are being incorporated into System 80+. The industry ALRR Program has established a very ambitious criterion of 87% for the plant availability of future nuclear units. To satisfy such a requirement, the next generation of nuclear plants will include a great many design improvements that reflect the hundreds of years of operating experience that we have accrued. C-ESA's System 80+ will include a number of design changes that improve operating margins and make the plant easier to operate and maintain. Not surprisingly, there is a great deal of overlap between improved safety and improved reliability. In the end, our design will satisfy the future needs of the utilities, the regulators, and the public. C-E is very pleased that KAERI is working with US to achieve these important goals
Design quality assurance for nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

NONE

1986-07-01

This Standard contains the requirements for the quality assurance program applicable to the design phase of a nuclear plant, and is applicable to the design of safety-related equipment, systems, and structures, as identified by the owner. 1 fig.
Design quality assurance for nuclear power plants

International Nuclear Information System (INIS)

1986-07-01

This Standard contains the requirements for the quality assurance program applicable to the design phase of a nuclear plant, and is applicable to the design of safety-related equipment, systems, and structures, as identified by the owner. 1 fig
The System 80+ Standard Plant design control document. Volume 23

International Nuclear Information System (INIS)

1997-01-01

This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume contains part 16 References and Appendix 19 A Design Alternatives for section 19 (Probabilistic Risk Assessment) of the ADM Design and Analysis. Also covered is section 20 Unresolved Safety Issues of the ADM Design and Analysis. Finally sections 1--6 of the ADM Emergency Operations Guidelines are contained in this volume. Information covered in these sections include: standard post-trip actions; diagnostic actions; reactor trip recovery guideline; LOCA recovery; SG tube rupture recovery
Virtual environments for nuclear power plant design

International Nuclear Information System (INIS)

Brown-VanHoozer, S.A.; Singleterry, R.C. Jr.; King, R.W.

1996-01-01

In the design and operation of nuclear power plants, the visualization process inherent in virtual environments (VE) allows for abstract design concepts to be made concrete and simulated without using a physical mock-up. This helps reduce the time and effort required to design and understand the system, thus providing the design team with a less complicated arrangement. Also, the outcome of human interactions with the components and system can be minimized through various testing of scenarios in real-time without the threat of injury to the user or damage to the equipment. If implemented, this will lead to a minimal total design and construction effort for nuclear power plants (NPP)
The System 80+ Standard Plant design control document. Volume 21

International Nuclear Information System (INIS)

1997-01-01

This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume contains parts 1--10 of section 19 (Probabilistic Risk Assessment) of the ADM Design and Analysis. Topics covered are: methodology; initiating event evaluation; accident sequence determination; data analysis; systems analysis; external events analysis; shutdown risk assessment; accident sequence quantification; and sensitivity analysis. Also included in this volume are Appendix 19.8A Shutdown Risk Assessment and Appendix A to Appendix 19.8A Request for Information

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)
The System 80+ Standard Plant design control document. Volume 20

International Nuclear Information System (INIS)

1997-01-01

This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume contains 2 technical specifications bases as part of Appendix 16 A Tech Spec Bases. They are TS B3.8 Electrical Power Technical Systems Bases and TS B3.9 Refueling Operations Bases. All 3 parts of section 17 (QA) and all 10 parts of section 18 (Human Factors) of the ADM Design and Analysis are contained in this volume. Topics covered in section 17 are: design phase QA; operations phase QA; and design phase reliability assurance. Topics covered by section 18 are: design team organization; design goals; design process; functional task analysis; control room configuration; information presentation; control and monitoring; verification and validation; and review documents
Improvement of layout and piping design for PWR nuclear power plants

International Nuclear Information System (INIS)

Nozue, Kosei; Waki, Masato; Kashima, Hiroo; Yoshioka, Tsuyoshi; Obara, Ichiro.

1983-01-01

For a nuclear power plant, a period of nearly ten years is required from the initial planning stage to commencement of transmission after passing through the design, manufacturing, installation and trial running stages. In the current climate there is a trend that the time required for nuclear power plant construction will further increase when locational problems, thorough explanation to residents in the neighborhood of the construction site and their under-standing, subsequent safety checks and measures to be taken in compliance with various controls and regulations which get tighter year after year, are taken into account. Under such circumstances, in order to satisfy requirements such as improving the reliability of the nuclear power plant design, manufacturing and construction departments, improvements in the economy as well as the quality and shortening of construction periods, the design structure for Mitsubishi PWR nuclear power plants was thoroughly consolidated with regard to layout and piping design. At the same time, diversified design improvements were made with the excellent domestic technology based on plant designs imported from the U.S.A. An outline of the priority items is introduced in this paper. (author)
Nuclear power plant system environmental design and decision methodology

International Nuclear Information System (INIS)

Zendehrouh, Z.; Shinozuka, M.; Schauer, F.P.

1975-01-01

The methodology described is concerned with a system reliability analysis by which the correlation among the level of design for the environmental and natural phenomena (earthquake, flood, tornado, etc.), reasonable practical measure of safety (such as conventional safety factor), and damage (radioactivity release) probability are established. In fact, the methodology indicates how the risk of environmental and natural hazard is combined with a specific design in order to evaluate damage probability associated with the design. This leads to the optimum design decision when combined further with the cost considerations involving the radioactivity release. This fundamental approach is essential in the design of nuclear plant structures, because, unlike the convential structures, the architectural considerations and structural analysis requirements alone cannot, by themselves, result in a balanced design in the framework of social requirements. The proposed methodology incorporates the different methods of environmental load determinations with their respective probabilistic formulations as well as detailed and advanced multi-discipline (structural, mechanical, soil, nuclear physics, biology, etc.) theoretical and empirical analysis including the effect of probabilistic nature of design variables, to establish a sound and reasonable design decision model for nuclear power plants. The information required for the analysis is also described and the areas for which further research is desirable are pointed out. Furthermore, the proposed methodology can very well be utilized to determine the requirements of standardized plants to facilitate the speed of their design and review process
Design requirements for new nuclear reactor facilities in Canada

International Nuclear Information System (INIS)

Shim, S.; Ohn, M.; Harwood, C.

2012-01-01

The Canadian Nuclear Safety Commission (CNSC) has been establishing the regulatory framework for the efficient and effective licensing of new nuclear reactor facilities. This regulatory framework includes the documentation of the requirements for the design and safety analysis of new nuclear reactor facilities, regardless of size. For this purpose, the CNSC has published the design and safety analysis requirements in the following two sets of regulatory documents: 1. RD-337, Design of New Nuclear Power Plants and RD-310, Safety Analysis for Nuclear Power Plants; and 2. RD-367, Design of Small Reactor Facilities and RD-308, Deterministic Safety Analysis for Small Reactor Facilities. These regulatory documents have been modernized to document past practices and experience and to be consistent with national and international standards. These regulatory documents provide the requirements for the design and safety analysis at a high level presented in a hierarchical structure. These documents were developed in a technology neutral approach so that they can be applicable for a wide variety of water cooled reactor facilities. This paper highlights two particular aspects of these regulatory documents: The use of a graded approach to make the documents applicable for a wide variety of nuclear reactor facilities including nuclear power plants (NPPs) and small reactor facilities; and, Design requirements that are new and different from past Canadian practices. Finally, this paper presents some of the proposed changes in RD-337 to implement specific details of the recommendations of the CNSC Fukushima Task Force Report. Major changes were not needed as the 2008 version of RD-337 already contained requirements to address most of the lessons learned from the Fukushima event of March 2011. (author)
Design basis reconstitution and configuration management of nuclear power plants

International Nuclear Information System (INIS)

Smith, P.R.

1989-01-01

The major design requirements of nuclear power plant components, systems, and structures are found in the plant's licensing commitments documented in the Final Safety Analysis Report and in the technical specification commitments of the plant. These specifications consider the original design and its degradation by in-service use. Before a nuclear power plant begins operation, the plant systems, structures, and organizational elements are functionally arranged to operate in a particular way. This functional arrangement is specified by the plant's design requirements and is called its configuration. The paper discusses configuration management and information management for configuration management. The management of large amounts of information and the various information systems associated with nuclear generating facilities is an ever-growing challenge for utilities. Plant operations involve a complex interrelation among data elements, especially in relation to design modifications and operational changes. Consequently, the operation of these data systems is interrelated and, as a result, redundant data items may exist. Thus, in view of the need to control and manage the plant configuration baseline, managers are striving to streamline their information management programs, which usually involves the integration of data-base systems
Analysis of effect of safety classification on DCS design in nuclear power plants

International Nuclear Information System (INIS)

Gou Guokai; Li Guomin; Wang Qunfeng

2011-01-01

By analyzing the safety classification for the systems and functions of nuclear power plants based on the general design requirements for nuclear power plants, especially the requirement of availability and reliability of I and C systems, the characteristics of modem DCS technology and I and C products currently applied in nuclear power field are interpreted. According to the requirements on the safety operation of nuclear power plants and the regulations for safety audit, the effect of different safety classifications on DCS design in nuclear power plants is analyzed, by considering the actual design process of different DCS solutions in the nuclear power plants under construction. (authors)
Designing for nuclear power plant maintainability and operability

International Nuclear Information System (INIS)

Pedersen, T.J.

1998-01-01

Experience has shown that maintenance and operability aspects must be addressed in the design work. ABB Atom has since long an ambition of achieving optimised, overall plant designs, and efficient feedback of growing operating experience has stepwise eliminated shortcomings, and yielded better and better plant operating performances. The records of the plants of the latest design versions are very good; four units in Sweden have operated at an energy availability of 90.1%, and the two Olkiluoto units in Finland at a load factor of 92.7%, over the last decade. The occupational radiation exposures have also been at a low level. The possibilities for implementing 'lessons learned' in existing plants are obviously limited by practical constraints. In Finland and Sweden, significant modernisations are still underway, however, involving replacement of mechanical equipment, and upgrading and backfitting of I and C systems on a large scale, in most of the plants. The BWR 90 design focuses on meeting requirements from utilities as well as new regulatory requirements, with a particular emphasis on the consequences of severe accidents; there shall be no large releases to the environment. Other design improvements involve: all-digital I and C systems and enhanced human factors engineering to improve work environment for operators, optimisation of buildings and containment to decrease construction time and costs, and selection of materials as well as maintenance of operating procedures to reduce radiation exposures even further. The BWR 90 design was offered to Finland in the early 1990s, but development work continues. It has been selected by a number of European utilities for assessing its conformance with the European Utility Requirements (EUR), aiming at a specific EUR Volume 3 for the BWR 90. Some characteristics of the ABB BWRs, with emphasis on features of importance for achieving improved economy and enhanced safety, are described below. (author)
Development of regulatory requirements/guides for desalination unit coupled with nuclear plant

International Nuclear Information System (INIS)

Jo, Jong Chull; Yune, Young Gill; Kim, Woong Sik

2005-10-01

The basic design of System-integrated Modular Advanced Reactor (SMART), a small-to-medium sized integral type pressurized water reactor (PWR) with the capacity of 330MWth, has been developed in Korea. In order to demonstrate the safety and performance of the SMART design, 'Development Project of SMART-P (SMART-Pilot Plant)' has been being performed as one of the 'National Mid and Long-term Atomic Energy R and D Programs', which includes design, construction, and start-up operation of the SMART-P with the capacity of 65MWth, a 1/5 scaled-down design of the SMART. At the same time, a study on the development of regulatory requirements/guides for the desalination unit coupled with nuclear plant has been carried out by KINS in order to prepare for the forthcoming SMART-P licensing. The results of this study performed from August of 2002 to October of 2005 can be summarized as follows: (1) The general status of desalination technologies has been survey. (2) The design of the desalination plant coupled with the SMART-P has been investigated. (3) The regulatory requirements/guides relevant to a desalination unit coupled with a nuclear plant have been surveyed. (4) A direction on the development of domestic regulatory requirements/guides for a desalination unit has been established. (5) A draft of regulatory requirements/guides for a desalination unit has been developed. (6) Expert technical reviews have been performed for the draft regulatory requirements/guides for a desalination unit. The draft regulatory requirements/guides developed in this study will be finalized and can be applied directly to the licensing of the SMART-P and SMART. Furthermore, it will be also applied to the licensing of the desalination unit coupled with the nuclear plant
The System 80+ Standard Plant design control document. Volume 17

International Nuclear Information System (INIS)

1997-01-01

This Design Control Document (DCD) is a repository of information comprising the System 80+trademark Standard Plant Design. The DCD also provides that design-related information to be incorporated by reference in the design certification rule for the System 80+ Standard Plant Design. Applicants for a combined license pursuant to 10 CFR 52 must ensure that the final Design Certification Rule and the associated Statements of Consideration are used when making all licensing decisions relevant to the System 80+ Standard Plant Design. The Design Control Document contains the DCD introduction, The Certified Design Material (CDM) [i.e., ''Tier 1''] and the Approved Design Material (ADM) [i.e., ''Tier 2''] for the System 80+ Standard Plant Design. The CDM includes the following sections: (1) Introductory material; (2) Certified Design Material for System 80+ systems and structures; (3) Certified Design Material for non-system-based aspects of the System 80+ Certified design; (4) Interface requirements; and (5) Site parameters. The ADM, to the extent applicable for the System 80+ Standard Plant Design, includes: (1) the information required for the final safety analysis report under 20 CFR 50.34; (2) other relevant information required by 10 CFR 52.47; and (3) emergency operations guidelines. This volume contains parts 2-7 and appendix 15A for section 15 (Accident Analysis) of the ADM Design and Analysis. Topics covered in these parts are: decrease in heat removal; decrease in RCS flow rate; power distribution anomalies; increase in RCS inventory; decrease in RCS inventory; release of radioactive materials. The appendix covers radiological release models. Also contained here are five technical specifications for section 16 (Technical Specifications) of the ADM Design and Analysis. They are: TS 1.0 Use and Applications; TS 2.0 Safety Limits; TS 3.0 LCO Availability; TS 3.1 Reactivity Control; and TS 3.2 Power Distribution
QA manpower requirement for nuclear power plants

International Nuclear Information System (INIS)

Link, M.

1980-01-01

To ensure the quality of the plant, QA activities are to be performed by the owner, the main contractor, the subcontractors and the Licensing Authority. The responsibilities of the QA-personnel of these organizations comprise as a minimum the control of the quality assurance systems and the proof of the quality requirements. Numbers of the required QA-personnel, designated for different tasks and recommended educational levels and professional qualifications will be given. (orig./RW)
GenoCAD Plant Grammar to Design Plant Expression Vectors for Promoter Analysis.

Science.gov (United States)

Coll, Anna; Wilson, Mandy L; Gruden, Kristina; Peccoud, Jean

2016-01-01

With the rapid advances in prediction tools for discovery of new promoters and their cis-elements, there is a need to improve plant expression methodologies in order to facilitate a high-throughput functional validation of these promoters in planta. The promoter-reporter analysis is an indispensible approach for characterization of plant promoters. It requires the design of complex plant expression vectors, which can be challenging. Here, we describe the use of a plant grammar implemented in GenoCAD that will allow the users to quickly design constructs for promoter analysis experiments but also for other in planta functional studies. The GenoCAD plant grammar includes a library of plant biological parts organized in structural categories to facilitate their use and management and a set of rules that guides the process of assembling these biological parts into large constructs.
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)
Conceptual design of small-sized HTGR system (4). Plant design and technical feasibility

International Nuclear Information System (INIS)

Ohashi, Hirofumi; Sato, Hiroyuki; Yan, Xing L.; Sumita, Junya; Nomoto, Yasunobu; Tazawa, Yujiro; Noguchi, Hiroki; Imai, Yoshiyuki; Tachibana, Yukio

2013-09-01

Japan Atomic Energy Agency (JAEA) has started a conceptual design of a 50MWt small-sized high temperature gas cooled reactor (HTGR) for steam supply and electricity generation (HTR50S), which is a first-of-kind of the commercial plant or a demonstration plant of a small-sized HTGR system for steam supply to the industries and district heating and electricity generation by a steam turbine, to deploy in developing countries in the 2020s. HTR50S was designed for steam supply and electricity generation by the steam turbine with the reactor outlet temperature of 750degC as a reference plant configuration. On the other hand, the intermediate heat exchanger (IHX) will be installed in the primary loop to demonstrate the electricity generation by the helium gas turbine and hydrogen production by thermochemical water splitting by utilizing the secondary helium loop with the reactor outlet temperature of 900degC as a future plant configuration. The plant design of HTR50S for the steam supply and electricity generation was performed based on the plant specification and the requirements for each system taking into account for the increase of the reactor outlet coolant temperature from 750degC to 900degC and the installation of IHX. The technical feasibility of HTR50S was confirmed because the designed systems (i.e., reactor internal components, reactor pressure vessel, vessel cooling system, shutdown cooling system, steam generator (SG), gas circulator, SG isolation and drainage system, reactor containment vessel, steam turbine and heat supply system) satisfies the design requirements. The conceptual plant layout was also determined. This paper provides the summary of the plan design and technical feasibility of HTR50S. (author)
System 80+TM standard plant: Design and operations overview

International Nuclear Information System (INIS)

Matzie, R.A.; Ritterbusch, S.E.

1999-01-01

The System 80+ Standard Plant Design is a 1400 MWe evolutionary Advanced Light Water Reactor (ALWR), designed to meet the Electric Power Research Institute (EPRI) ALWR Utility Requirements Document (URD) and the demands of the international market for nuclear power plants which are not only safer but also more economical to maintain and operate. ABB Combustion Engineering Nuclear Power used a defense-in-depth process that (1) adds design margin to basic components to improve performance during normal operation and to decrease the likelihood of an unanticipated transient or an accident, (2) improves the redundancy and diversity of safety systems in order to mitigate design basis accidents and prevent severe accidents, and (3) improves severe accident mitigation capability. This paper describes the most important improved systems and components with emphasis on severe accident prevention and mitigation capability. The improved design features were implemented in an evolutionary manner using proven components. This approach ensures that the plant operates safely and economically, as demonstrated by operating plants in the US and the Republic of Korea. Detailed studies, summarized in this paper, have shown that the System 80+ plant availability is expected to exceed the ALWR requirement of 87% and that the annual operations and maintenance costs are expected to be reduced by $14 million. (author)
Control room design and human engineering in power plants

International Nuclear Information System (INIS)

Herbst, L.; Hinz, W.

1981-01-01

Automation reduces the human work load. Employment of functional areas permits optimization of operational sequences. Computer based information processing makes it possible to output information in accordance with operating requirements. Design based on human engineering principles assures the quality of the interaction between the operator and the equipment. The degree to which these conceptional features play a role in design of power plant control rooms depends on the unit rating, the mode of operation and on the requirements respecting safety and availability of the plant. (orig./RW)
Construction plant requirements for nuclear sites

International Nuclear Information System (INIS)

Tatum, C.B.; Harris, J.A.

1981-01-01

Planning and developing the temporary construction plant facilities for a nuclear project is equivalent to providing utility services for a small city. Provision of adequate facilities is an important factor in the productivity of both the manual and non-manual work force. This paper summarizes construction facility requirements for a two unit (1300 MWe each) nuclear project. Civil, mechanical and electrical facilities are described, including design, installation and operation. Assignment of responsibility for specific work tasks regarding the construction plant is also discussed. In presenting this data, the authors seek to transfer experience and assist in the provision of adequate facilities on future projects
System Definition and Analysis: Power Plant Design and Layout

International Nuclear Information System (INIS)

1996-01-01

This is the Topical report for Task 6.0, Phase 2 of the Advanced Turbine Systems (ATS) Program. The report describes work by Westinghouse and the subcontractor, Gilbert/Commonwealth, in the fulfillment of completing Task 6.0. A conceptual design for critical and noncritical components of the gas fired combustion turbine system was completed. The conceptual design included specifications for the flange to flange gas turbine, power plant components, and balance of plant equipment. The ATS engine used in the conceptual design is an advanced 300 MW class combustion turbine incorporating many design features and technologies required to achieve ATS Program goals. Design features of power plant equipment and balance of plant equipment are described. Performance parameters for these components are explained. A site arrangement and electrical single line diagrams were drafted for the conceptual plant. ATS advanced features include design refinements in the compressor, inlet casing and scroll, combustion system, airfoil cooling, secondary flow systems, rotor and exhaust diffuser. These improved features, integrated with prudent selection of power plant and balance of plant equipment, have provided the conceptual design of a system that meets or exceeds ATS program emissions, performance, reliability-availability-maintainability, and cost goals
Safety criteria for design of nuclear power plants

International Nuclear Information System (INIS)

1997-01-01

In Finland the general safety requirements for nuclear power plants are presented in the Council of State Decision (395/91). In this guide, safety principles which supplement the Council of State Decision and which are to be used in the design of nuclear power plants are defined
Human factor engineering applied to nuclear power plant design

International Nuclear Information System (INIS)

Manrique, A.; Valdivia, J.C.

2007-01-01

Advantages of implementing adequate Human Factor Engineering techniques in the design of nuclear reactors have become not only a fact recognized by the majority of engineers and operators but also an explicit requirement regulated and mandatory for the new designs of the so called advanced reactors. The first step for this is preparing a plan to incorporate all the Human Factor Engineering principles and developing an integral design of the Instrumentation and Control and Man-machine interface systems. Such a plan should state: -) Activities to be performed, and -) Creation of a Human Factor Engineering team adequately qualified. The Human Factor Engineering team is an integral part of the design team and is strongly linked to the engineering organizations but simultaneously has independence to act and is free to evaluate designs and propose changes in order to enhance human behavior. TECNATOM S.A. (a Spanish company) has been a part of the Design and Human Factor Engineering Team and has collaborated in the design of an advanced Nuclear Power Plant, developing methodologies and further implementing those methodologies in the design of the plant systems through the development of the plant systems operational analysis and of the man-machine interface design. The methodologies developed are made up of the following plans: -) Human Factor Engineering implementation in the Man-Machine Interface design; -) Plant System Functional Requirement Analysis; -) Allocation of Functions to man/machine; -) Task Analysis; -) Human-System Interface design; -) Control Room Verification and -) Validation

Code on the safety of nuclear power plants: Design

International Nuclear Information System (INIS)

1988-01-01

This Code is a compilation of nuclear safety principles aimed at defining the essential requirements necessary to ensure nuclear safety. These requirements are applicable to structures, systems and components, and procedures important to safety in nuclear power plants embodying thermal neutron reactors, with emphasis on what safety requirements shall be met rather than on specifying how these requirements can be met. It forms part of the Agency's programme for establishing Codes and Safety Guides relating to land based stationary thermal neutron power plants. The document should be used by organizations designing, manufacturing, constructing and operating nuclear power plants as well as by regulatory bodies
Design requirements of instrumentation and control systems for next generation reactor

International Nuclear Information System (INIS)

Koo, In Soo; Lee, Byung Sun; Park, Kwang Hyun; Park, Heu Yoon; Lee, Dong Young; Kim, Jung Taek; Hwang, In Koo; Chung, Chul Hwan; Hur, Seop; Kim, Chang Hoi; Na, Nan Ju

1994-03-01

In this report, the basic design requirements of Instrumentation and Control systems for next generation reactor are described, which are top-tier level, to support the advanced I and C systems. It contains the requirements in accordance with the plant reliability, the plant performance, the operator's aid functions, the features for maintenance and testing, licensing issues for I and C systems. Advanced I and C systems are characterized such as the application of the digital and the human engineering technologies. To development of this requirements, the I and C systems for the foreign passive and the evolutionary types of reactor and the domestic conventional reators were reviewed and anlysed. At the detail design stage, these requirements will be used for top-tier requirements. To develop the detail design requirements in the future, more quantitive and qualitive analyses are need to be added. (Author) 44 refs
Design requirements of instrumentation and control systems for next generation reactor

Energy Technology Data Exchange (ETDEWEB)

Koo, In Soo; Lee, Byung Sun; Park, Kwang Hyun; Park, Heu Yoon; Lee, Dong Young; Kim, Jung Taek; Hwang, In Koo; Chung, Chul Hwan; Hur, Seop; Kim, Chang Hoi; Na, Nan Ju [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1994-03-01

In this report, the basic design requirements of Instrumentation and Control systems for next generation reactor are described, which are top-tier level, to support the advanced I and C systems. It contains the requirements in accordance with the plant reliability, the plant performance, the operator`s aid functions, the features for maintenance and testing, licensing issues for I and C systems. Advanced I and C systems are characterized such as the application of the digital and the human engineering technologies. To development of this requirements, the I and C systems for the foreign passive and the evolutionary types of reactor and the domestic conventional reators were reviewed and anlysed. At the detail design stage, these requirements will be used for top-tier requirements. To develop the detail design requirements in the future, more quantitive and qualitive analyses are need to be added. (Author) 44 refs.
Flood control design requirements and flood evaluation methods of inland nuclear power plant

International Nuclear Information System (INIS)

Zhang Ailing; Wang Ping; Zhu Jingxing

2011-01-01

Effect of flooding is one of the key safety factors and environmental factors in inland nuclear power plant sitting. Up to now, the rule of law and standard systems are established for the selection of nuclear power plant location and flood control requirements in China. In this paper flood control standards of China and other countries are introduced. Several inland nuclear power plants are taken as examples to thoroughly discuss the related flood evaluation methods. The suggestions are also put forward in the paper. (authors)
Conceptual design of a laser fusion power plant. Part I. An integrated facility

International Nuclear Information System (INIS)

1981-07-01

This study is a new preliminary conceptual design and economic analysis of an inertial confinement fusion (ICF) power plant performed by Bechtel under the direction of Lawrence Livermore National Laboratory (LLNL). The purpose of a new conceptual design is to examine alternatives to the LLNL HYLIFE power plant and to incorporate information from the recent liquid metal cooled power plant conceptual design study (CDS) into the reactor system and balance of plant design. A key issue in the design of a laser fusion power plant is the degree of symmetry in the illumination of the target that will be required for a proper burn. Because this matter is expected to remain unresolved for some time, another purpose of this study is to determine the effect of symmetry requirements on the total plant size, layout, and cost
European utility requirements: common rules to design next LWR plants in an open electricity market

International Nuclear Information System (INIS)

Berbey, Pierre; Ingemarsson, Karl-Fredrik

2004-01-01

The major European electricity producers want to keep able to build new nuclear power plants and they believe 3. generation LWRs would be the most adapted response to their needs in the first decades of this century. Producing a common European Utility Requirement (EUR) document has been one of the basic tasks towards this objective. In this common frame, standardized and competitive LWR NPPs could be developed and offered to the investors. This idea is now well supported by all the other actors on the European electricity market: vendors, regulators, grid managers, administrations although in the competitive and unified European electricity market that is emerging, the electricity producers' stakes are more and more different from the other electricity business actors'. The next term objectives of the electricity producers involved in EUR are focused on negotiating common rules of the game together with the regulators. This covers the nuclear safety approaches, the conditions requested to connect a plant to a HV grid, as well as the design standards. Discussions are going on between the EUR organization and all the corresponding bodies to develop stabilized and predictable design rules that would meet the constraints of nuclear electricity generation in this new environment. Finally there cannot be competition without competitors. The EUR organization has proven to be the right place to establish trustful relationship between the vendors and their potential customers, through fair assessment of the proposed designs performance vs. the utility needs. This will be continued and developed with the main vendors present in Europe, so as to keep alive a list of 4 to 6 designs 'qualified', i.e. showing an acceptable score of non-compliance vs. EUR. (authors)
Technical design considerations in the provision of a commercial MOX plant

International Nuclear Information System (INIS)

Elliott, M.F.

1997-01-01

The Sellafield MOX Plant (SMP) has a design production target of 120 t/year Heavy Metal of mixed uranium dioxide and plutonium dioxided (MOX) fuel. It will have the capability to produce fuel with fissile enrichments up to 10%. The feed materials are those arising from reprocessing operations on the Sellafield site, although the plant also has the capability to receive and process plutonium from overseas reprocessing plants. The ability to produce 10% enriched fuels, together with the requirement to use high burn-up feed has posed a number of design challenges to prevent excessive powder temperatures within the plant. As no stimulants are available to represent the heat generating nature of plutonium powders, it is difficult to prove equipment design by experiment. Extensive use has therefore been made of finite element analysis techniques. The requirement to process material of low burn-up (i.e. high fissile enrichment) has also impacted on equipment design in order to ensure that criticality limits are not exceeded. This has been achieved where possible by 'safe by geometry' design and, where appropriate, by high integrity protection systems. SMP has been designed with a high plant availability but at minimum cost. The requirement to minimize cost has meant that high availability must be obtained with the minimum of equipment. This had led to major challenges for equipment designers in terms of both the reliability and also the maintainability of equipment. Extensive use has been made of theoretical modelling techniques which have given confidence that plant throughput can be achieved. (author). 1 fig
An Approach to Establish Design Requirements for Human-System Interface (HSI) of Automatic Systems in Nuclear Power Plants

Energy Technology Data Exchange (ETDEWEB)

Nuraslinda, Anuar; Kim, Jonghyun [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2013-05-15

This paper aims to demonstrate an approach to establish the design requirements for automatic systems in nuclear power plant (NPP) by using a powerful tool called Itemized Sequence Diagram (ISD). The process starts with function allocation by defining a set of levels of automation (LOAs). Then, task allocation is done using the ISD and finally the design requirements are established by examining the interaction points between human operator and automation, which are all located on the interface as modeled in the ISD. The strengths of this approach are discussed and a suggestion to integrate with that of the methodology employed to produce the existing guidelines or guidance is included in this paper. Some issues of automation have been addressed earlier in this paper and 12 design requirements that address human-system interaction were suggested by using the ISD as a tool to identify the interaction points between human operator and automation. The integration of the proposed approach in this paper with that of existing guidance could result in the new issue identification that would call for the establishment of new guidance. For example, Requirement 11 states that the HSI should provide the means for take-over from automatic to manual control was not mentioned in the existing guidance.
General requirements for concrete containment structures for CANDU nuclear power plants

International Nuclear Information System (INIS)

1993-07-01

This standard provides the general requirements used in the design, construction, testing, and commissioning of concrete containment structures for CANDU nuclear power plants designated as class containment and is directed to the owners, designers, manufacturers, fabricators, and constructors of the concrete components and parts
Rule-Based Design of Plant Expression Vectors Using GenoCAD.

Science.gov (United States)

Coll, Anna; Wilson, Mandy L; Gruden, Kristina; Peccoud, Jean

2015-01-01

Plant synthetic biology requires software tools to assist on the design of complex multi-genic expression plasmids. Here a vector design strategy to express genes in plants is formalized and implemented as a grammar in GenoCAD, a Computer-Aided Design software for synthetic biology. It includes a library of plant biological parts organized in structural categories and a set of rules describing how to assemble these parts into large constructs. Rules developed here are organized and divided into three main subsections according to the aim of the final construct: protein localization studies, promoter analysis and protein-protein interaction experiments. The GenoCAD plant grammar guides the user through the design while allowing users to customize vectors according to their needs. Therefore the plant grammar implemented in GenoCAD will help plant biologists take advantage of methods from synthetic biology to design expression vectors supporting their research projects.
Design of chemical plant

International Nuclear Information System (INIS)

Lee, Dong Il; Kim, Seung Jae; Yang, Jae Ho; Ryu, Hwa Won

1993-01-01

This book describes design of chemical plant, which includes chemical engineer and plan for chemical plant, development of chemical process, cost engineering pattern, design and process development, general plant construction plan, project engineering, foundation for economy on assets and depreciation, estimation for cost on capital investment and manufacturing cost, design with computers optimal design and method like fluid mechanics design chemical device and estimation for cost, such as dispatch of material and device writing on design report and appendixes.
Architect engineer balance-of-plant radiological design considerations

International Nuclear Information System (INIS)

Piccot, A.R.

1975-01-01

Methods which are or may be used by Architect Engineers in dealing with the problems of radiological safety in the design of a nuclear power plant are discussed. The bases and basic requirements for a radiation protection program are briefly noted. Requirements in the areas of planning, organization, responsibilities and implementation of radiation protection are discussed. Lists of safety tasks which should be performed during the various design phases are presented
Plant dynamics studies towards design of plant protection system for PFBR

Energy Technology Data Exchange (ETDEWEB)

Natesan, K., E-mail: natesan@igcar.gov.in [Nuclear and Safety Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Kasinathan, N.; Velusamy, K.; Selvaraj, P.; Chellapandi, P. [Nuclear and Safety Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India)

2012-09-15

Highlights: Black-Right-Pointing-Pointer Analysis of various design basis events in a fast breeder reactor towards design of plant protection system. Black-Right-Pointing-Pointer Plant dynamic modeling of a sodium cooled fast breeder reactor. Black-Right-Pointing-Pointer Selection of optimum set of plant parameters for considering best plant availability. - Abstract: Prototype fast breeder reactor (PFBR) is a 500 MWe (1250 MWt) liquid sodium cooled pool type reactor currently under construction in India. For a safe and efficient operation of the plant, it is necessary that the reactor is protected from all the transients that may occur in the plant. In order to accomplish this, adequate number of SCRAM parameters is required in the plant protection system with reliable instrumentation. For identifying the SCRAM parameters, the neutronic and thermal hydraulic responses of the plant for various possible events need to be established. Towards this, a one dimensional plant dynamics code DYANA-P has been developed with thermal hydraulic models for reactor core, hot and cold pools, intermediate heat exchangers, pipelines, steam generator, primary sodium circuits and secondary sodium circuits. The code also incorporates neutron kinetics and reactivity feedback models. By a comprehensive plant dynamics study an optimum list of SCRAM parameters and the maximum permissible response time for various instruments used for deriving them have been arrived at.
Plutonium Finishing Plant (PFP) Standards/Requirements Identification Document (S/RID)

Energy Technology Data Exchange (ETDEWEB)

Maddox, B.S.

1996-01-01

This Standards/Requirements Identification Document (S/RID) sets forth the Environmental Safety and Health (ESH) standards/requirements for the Plutonium Finishing Plant (PFP). This S/RID is applicable to the appropriate life cycle phases of design, construction, operation, and preparation for decommissioning. These standards/requirements are adequate to ensure the protection of the health and safety of workers, the public, and the environment.
Plutonium Finishing Plant (PFP) Standards/Requirements Identification Document (S/RID)

International Nuclear Information System (INIS)

Maddox, B.S.

1996-01-01

This Standards/Requirements Identification Document (S/RID) sets forth the Environmental Safety and Health (ESH) standards/requirements for the Plutonium Finishing Plant (PFP). This S/RID is applicable to the appropriate life cycle phases of design, construction, operation, and preparation for decommissioning. These standards/requirements are adequate to ensure the protection of the health and safety of workers, the public, and the environment
Impact of extreme load requirements and quality assurance on nuclear power plant costs

International Nuclear Information System (INIS)

Stevenson, J.D.

1993-01-01

Definitive costs, applicable to nuclear power plant concrete structures, as a function of National Regulatory Requirements, standardization, the effect of extreme load design associated with both design basis accidents and extreme external events and quality assurance are difficult to develop since such effects are interrelated and not only differ widely from country to country, project to project but also vary in time. Table 1 shows an estimate of the of the overall plant cost effects of external event extreme load design on nuclear power plant design for the U.S -and selected foreign countries for which experience with LWRs exist- Germany is the most expensive primarily due to a military aircraft crash resistance. However, the German requirement for 4 safeguards trains rather than 2 and the containment design requirement to consider one Steam Generator blowdown concurrent with a RCS blowdown. This presentation will concentrate on the direct current impact extreme load design and quality assurance have on concrete structures, systems and components for nuclear plants. This presentation is considered timely due to the increased interest in the c potential backfit of Eastern European nuclear power stations of the WWER 440 and WWER 1000 types which typically did not consider the extreme loads identified in Table 1 and accident loads in Table 3 and quality assurance in Table 5 in their original design. Concrete structures in particular are highlighted because they typically form the last barrier to radioactive release from the containment and other Safety Related Structures
An AI-based layout design system for nuclear power plants

International Nuclear Information System (INIS)

Fujita, Kikuo; Akagi, Shinsuke; Nakatogawa, Tetsundo; Tanaka, Kazuo; Takeuchi, Makoto.

1991-01-01

An AI-based layout design system for nuclear power plants has been developed. The design of the layout of nuclear power plants is a time-consuming task requiring expertise, in which a lot of machinery and equipment must be arranged in a plant building considering various kinds of design constraints, i.e. spatial, functional, economical etc. Computer aided layout design systems have been widely expected and the application of AI technology is expected as a promising approach for the synthesis phase of this task. In this paper, we present an approach to the layout design of nuclear power plants based on a constraint-directed search; one of the AI techniques. In addition, we show how it was implemented with an object-oriented programming technique and give an example of its application. (author)
Overall quality assurance program requirements for nuclear power plants

International Nuclear Information System (INIS)

1992-09-01

This standard contains the requirements for the owner's overall quality assurance program for a nuclear power plant. This program encompasses all phases of a nuclear power plant life cycle, including site evaluation, design, procurement, manufacturing, construction and installation, commissioning, operation, and decommissioning. It covers the activities associated with specifying, directing, and administering the work to be done during these phases, and the evaluation and integrated of the activities and programs of participants
LMFBR plant design features for sodium spill and fire protection

International Nuclear Information System (INIS)

Palm, R.E.

1982-01-01

Design features have been developed for an LMFBR plant to protect the concrete structures from potential liquid spills and fires and prevent sodium-concrete reactions. The inclusion of these features in the plant design reduces the severity of design basis accident conditions imposed on containment and other critical plant structures. Steel liners are provided in cells containing radioactive sodium systems, and catch pans are located in non-radioactive sodium system cells. The design requirements and descriptions of each of these protective features are presented. The loading conditions, analytical approach and numerical results are also included. Design of concrete cell structures that are subject to high temperature effects from sodium spills is discussed. The structural design considers the influence of high temperature on design properties of concrete and carbon steel materials based on results of a comprehensive test program. The development of these design features and high temperature design considerations for the Clinch River Breeder Reactor Plant (CRBRP) are presented in this paper
ASME section XI - design and access requirements for in-service inspection

International Nuclear Information System (INIS)

Davis, D.D.

1982-01-01

The Owner of a nuclear power plant has the regulatory commitment to perform Section XI in-service inspection throughout the service life of a plant. In anticipation of what will be needed to perform adequately the required examinations and tests, sub-article IWA-1500 of Section XI not only requires that sufficient access be provided to accommodate equipment and inspection personnel but also requires that other provisions be considered such as: component surface preparations, material selections, shielding, removal and storage of hardware, handling equipment, and provisions for repairs and replacements. It is, therefore, the owner's and the architect engineer's responsibility to ensure that proper design and access provisions are incorporated to enable the owner to meet his commitments. Since the architect engineer usually has the prime responsibility for the implementation of design criteria, the owner must ensure that these provisions be considered in each phase of design and construction. The benefits of this can result in shorter outages, more meaningful examinations and tests and less radiation exposure of inspection personnel. This paper will address in detail those topics that affect design and access provisions which need to be considered during the design and construction of a nuclear power plant. (author)

Materials for Nuclear Plants From Safe Design to Residual Life Assessments

CERN Document Server

Hoffelner, Wolfgang

2013-01-01

The clamor for non-carbon dioxide emitting energy production has directly impacted on the development of nuclear energy. As new nuclear plants are built, plans and designs are continually being developed to manage the range of challenging requirement and problems that nuclear plants face especially when managing the greatly increased operating temperatures, irradiation doses and extended design life spans. Materials for Nuclear Plants: From Safe Design to Residual Life Assessments provides a comprehensive treatment of the structural materials for nuclear power plants with emphasis on advanced design concepts. Materials for Nuclear Plants: From Safe Design to Residual Life Assessments approaches structural materials with a systemic approach. Important components and materials currently in use as well as those which can be considered in future designs are detailed, whilst the damage mechanisms responsible for plant ageing are discussed and explained. Methodologies for materials characterization, material...
Design of XML-based plant data model

International Nuclear Information System (INIS)

Nair, Preetha M.; Padmini, S.; Gaur, Swati; Diwakar, M.P.

2013-01-01

XML has emerged as an open standard for exchanging structured data on various platforms to handle rich, nested, complex data structures. XML with its flexible tree-like data structure allows a more natural representation as compared to traditional databases. In this paper we present data model for plant data acquisition systems captured using XML technologies. Plant data acquisition systems in a typical Nuclear Power Plant consists of embedded nodes at the first tier and operator consoles at the second tier for operator operation, interaction and display of Plant parameters. This paper discusses a generic data model that was designed to capture process, network architecture, communication/interface protocol and diagnostics aspects required for a Nuclear Power Plant. (author)
User requirements and conceptual design of the ITER Electron Cyclotron Control System

Energy Technology Data Exchange (ETDEWEB)

Carannante, Giuseppe, E-mail: Giuseppe.Carannante@F4E.europa.eu [Fusion for Energy, Josep Pla 2, Barcelona 08019 (Spain); Cavinato, Mario [Fusion for Energy, Josep Pla 2, Barcelona 08019 (Spain); Gandini, Franco [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Granucci, Gustavo [Istituto di Fisica del Plasma ENEA-CNR-EURATOM, via Cozzi 53, 20125 Milano (Italy); Henderson, Mark; Purohit, Dharmesh [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Saibene, Gabriella; Sartori, Filippo [Fusion for Energy, Josep Pla 2, Barcelona 08019 (Spain); Sozzi, Carlo [Istituto di Fisica del Plasma ENEA-CNR-EURATOM, via Cozzi 53, 20125 Milano (Italy)

2015-10-15

The ITER Electron Cyclotron (EC) plant is a complex system, essential for plasma operation. The system is being designed to supply up to 20 MW of power at 170 GHz; it consists of 24 RF sources (or Gyrotrons) connected by switchable transmission lines to four upper and one equatorial launcher. The complexity of the EC plant requires a Plant Controller, which provides the functional and operational interface with CODAC and the Plasma Control System and coordinates the various Subsystem Control Units, i.e. the local controllers of power supplies, Gyrotrons, transmission lines and launchers. A conceptual design of the Electron Cyclotron Control System (ECCS) was developed, starting from the collection of the user requirements, which have then been organized as a set of operational scenarios exploiting the EC system. The design consists in a thorough functional analysis, including also protection functions, and in the development of a conceptual I&C architecture. The main aim of the work was to identify the physics requirements and to translate them into control system requirements, in order to define the interfaces within the components of the ECCS. The definition of these interfaces is urgent because some of the subsystems are already in an advanced design phase. The present paper describes both the methodology used and the resulting design.
Development of computer-aided design and production system for nuclear power plant

International Nuclear Information System (INIS)

Ishii, Masanori

1983-01-01

The technically required matters related to the design and production of nuclear power stations tended to increase from the viewpoint of the safety and reliability, and it is indispensable to cope with such technically required matters skillfully for the rationalization of the design and production and for the construction of highly reliable plants. Ishikawajima Harima Heavy Industries Co., Ltd., has developed the computer-aided design data information and engineering system which performs dialogue type design and drawing, and as the result, the design-production consistent system is developed to do stress analysis, production design, production management and the output of data for numerically controlled machine tools consistently. In this paper, mainly the consistent system in the field of plant design centering around piping and also the computer system for the design of vessels and others are outlined. The features of the design works for nuclear power plants, the rationalization of the design and production management of piping and vessels, and the application of the CAD system to other general equipment and improvement works are reported. This system is the powerful means to meet the requirement of heightening quality and reducing cost. (Kako, I.)
Support to design and construction of the PBMR plant

International Nuclear Information System (INIS)

Cazorla, F.; Moron, P.; Gonzalez, J. I.

2010-01-01

Developing the new reactor design to a licensable state for constructing a pilot plant is a tough task require specific resources, concerning knowledge and previous experience, which trespassing the pure scientific or technologic knowledge linked to the reactor conceptual design. Taking into consideration the experience derived from the collaboration between the South African company PBMR (PTY) Ltd.; the Pebble Bed Modular Reactor Designer, and Tecnatom SA, the article presents some of the aspects in which the companies or organization in charge of the design can demand external support to license and construct the pilot plants with guaranteed success. (Author)
Design of nuclear power generation plants adopting model engineering method

International Nuclear Information System (INIS)

Waki, Masato

1983-01-01

The utilization of model engineering as the method of design has begun about ten years ago in nuclear power generation plants. By this method, the result of design can be confirmed three-dimensionally before actual production, and it is the quick and sure method to meet the various needs in design promptly. The adoption of models aims mainly at the improvement of the quality of design since the high safety is required for nuclear power plants in spite of the complex structure. The layout of nuclear power plants and piping design require the model engineering to arrange rationally enormous quantity of things in a limited period. As the method of model engineering, there are the use of check models and of design models, and recently, the latter method has been mainly taken. The procedure of manufacturing models and engineering is explained. After model engineering has been completed, the model information must be expressed in drawings, and the automation of this process has been attempted by various methods. The computer processing of design is in progress, and its role is explained (CAD system). (Kako, I.)
Information management system for design, construction and operation of nuclear power plants

International Nuclear Information System (INIS)

Bolch, M.C.; Jones, C.R.

1990-01-01

This paper describes the principal requirements and features of a computerized information management system (IMS) believed to be a necessary part of the program to design, build and operate the next generation of nuclear power plants in the United States. This way a result of extensive review and input from an industry group studying future nuclear power plant construction improvements. The needs of the power plant constructor, owner and operator for such a computerized technical data base are described in terms of applications and scope and timing of turnover of the IMS by the plant designer. The applications cover the full life cycle of the plant including project control, construction activities, quality control, maintenance and operation. The scope of the IMS is also described in terms of the technical data to be included, hardware and software capabilities and training. The responsibilities of the plant designer for developing the IMS and generating the technical data base is defined as part of the plant process. The requirements to be met include a comprehensive plant data model and computer system hardware and software
Information management system for design, construction and operation of nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Bolch, M.C. (Duke Power Co. (US)); Jones, C.R. (S. Levy Inc. (US))

1990-01-01

This paper describes the principal requirements and features of a computerized information management system (IMS) believed to be a necessary part of the program to design, build and operate the next generation of nuclear power plants in the United States. This way a result of extensive review and input from an industry group studying future nuclear power plant construction improvements. The needs of the power plant constructor, owner and operator for such a computerized technical data base are described in terms of applications and scope and timing of turnover of the IMS by the plant designer. The applications cover the full life cycle of the plant including project control, construction activities, quality control, maintenance and operation. The scope of the IMS is also described in terms of the technical data to be included, hardware and software capabilities and training. The responsibilities of the plant designer for developing the IMS and generating the technical data base is defined as part of the plant process. The requirements to be met include a comprehensive plant data model and computer system hardware and software.
International requirements for life extension of nuclear power plants

International Nuclear Information System (INIS)

Wernicke, Robert

2009-01-01

Lifetime extension or long-term operation of nuclear facilities are topics of great international significance against the backdrop of a fleet of nuclear power plants of which many have reached 2/3 of their planned life. The article deals with the conditions for, and the specific requirements of, seeking long-term operation of nuclear power plants as established internationally and on the basis of IAEA collections. Technically, long-term operation is possible for many of the nuclear power plants in the world because, normally, they were built on the basis of conservative rules and regulations and, as a consequence, incorporate significant additional safety. Application of requirements to specific plants implies assessments of technical safety which show that conservative design philosophies created reserves and, as a consequence, there is an adequate level of safety also in long-term plant operation. For this purpose, the technical specifications must be revised, necessary additions made, and (international) operating experience taken into account and management of aging established. Two examples are presented to show how the approach to long-term plant operation is put into practice on a national level. (orig.)
MPD3: a useful medicinal plants database for drug designing.

Science.gov (United States)

Mumtaz, Arooj; Ashfaq, Usman Ali; Ul Qamar, Muhammad Tahir; Anwar, Farooq; Gulzar, Faisal; Ali, Muhammad Amjad; Saari, Nazamid; Pervez, Muhammad Tariq

2017-06-01

Medicinal plants are the main natural pools for the discovery and development of new drugs. In the modern era of computer-aided drug designing (CADD), there is need of prompt efforts to design and construct useful database management system that allows proper data storage, retrieval and management with user-friendly interface. An inclusive database having information about classification, activity and ready-to-dock library of medicinal plant's phytochemicals is therefore required to assist the researchers in the field of CADD. The present work was designed to merge activities of phytochemicals from medicinal plants, their targets and literature references into a single comprehensive database named as Medicinal Plants Database for Drug Designing (MPD3). The newly designed online and downloadable MPD3 contains information about more than 5000 phytochemicals from around 1000 medicinal plants with 80 different activities, more than 900 literature references and 200 plus targets. The designed database is deemed to be very useful for the researchers who are engaged in medicinal plants research, CADD and drug discovery/development with ease of operation and increased efficiency. The designed MPD3 is a comprehensive database which provides most of the information related to the medicinal plants at a single platform. MPD3 is freely available at: http://bioinform.info .
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)
An estimate and evaluation of design error effects on nuclear power plant design adequacy

International Nuclear Information System (INIS)

Stevenson, J.D.

1984-01-01

An area of considerable concern in evaluating Design Control Quality Assurance procedures applied to design and analysis of nuclear power plant is the level of design error expected or encountered. There is very little published data 1 on the level of error typically found in nuclear power plant design calculations and even less on the impact such errors would be expected to have on overall design adequacy of the plant. This paper is concerned with design error associated with civil and mechanical structural design and analysis found in calculations which form part of the Design or Stress reports. These reports are meant to document the design basis and adequacy of the plant. The estimates contained in this paper are based on the personal experiences of the author. In Table 1 is a partial listing of the design docummentation review performed by the author on which the observations contained in this paper are based. In the preparation of any design calculations, it is a utopian dream to presume such calculations can be made error free. The intent of this paper is to define error levels which might be expected in a competent engineering organizations employing currently technically qualified engineers and accepted methods of Design Control. In addition, the effects of these errors on the probability of failure to meet applicable design code requirements also are estimated
System design methodology of non-fossil fuel fired power plants

International Nuclear Information System (INIS)

Mohamed, J.A.; Guven, H.M.

1992-01-01

In most thermal system designs, economic and thermodynamic aspects of the design are not studied simultaneously early on in the design process. As a result, the economic ramification of thermodynamic changes to the system configuration, and vice versa, are not immediately apparent to the designer or the performance, involving both thermal and economic aspects of the plant. In this study, a rational approach is presented to formalize the design process of small power plants, typically, burning non-conventional fuel sources such as wood residues, tires, biofuels, etc. The method presented in this paper allows for handling of process information, both qualitative and quantitative, to enable the designer to change his design in an optimal manner. A two-level design structure (macro-level and micro-level), is introduced to enable the designer to adapt his design in an efficient manner to the available (or required) technology-level, type of application, economic factors, O and M requirements, etc. At the macro-level of design, economic feasibility (business) decisions are made, while at the micro-level of design, technical feasibility (engineering) decisions are made
Fukushima, two years later, modification requirements in nuclear power plants

International Nuclear Information System (INIS)

Sanchez J, J.; Camargo C, R.; Nunez C, A.; Mendoza F, J. E.; Salmeron V, J. A.

2013-10-01

The occurred events in the nuclear power plant of Fukushima Daiichi as consequence of the strong earthquake of 9 grades in the Richter scale and the later tsunami with waves estimated in more than 14 meters high began a series of important questions about the safety of the nuclear power plants in operation and of the new designs. Firstly, have allowed to be questioned on the magnitudes and consequences of the extreme external natural events; that can put in risk the integrity of the safety barriers of a nuclear power plant when being presented in a multiple way. As consequence of the events of the Fukushima Daiichi NPP, the countries with NPPs in operation and /or construction carried out evaluations about their safety operation. They have also realized evaluations about accidents and their impact in the safety, analysis and studies too that have forced to the regulatory bodies to continue a systematic and methodical revision of their procedures and regulations, to identify the possible improvements to the safety in response to the events happened in Japan; everything has taken it to determine the necessity to incorporate additional requirements to the nuclear power plants to mitigate events Beyond the Design Base. Due to Mexico has the nuclear power plant of Laguna Verde, with two units of BWR-5 type with contention Mark III, some the modifications can be applicable to these units to administrate and/or to mitigate the consequences of the possible occurrence of an accident Beyond the Design Base and that could generate a severe accident. In this work an exposition is presented on the modification requirements to confront external natural events Beyond the Design Base, and its application in our country. (Author)
Nuclear energy research initiative, an overview of the cooperative program for the risk-informed assessment of regulatory and design requirements for future nuclear power plants

International Nuclear Information System (INIS)

Ritterbusch, Stanley E.

2000-01-01

EPRI sstudies have shown that nuclear plant capital costs will have to decrease by about 35% to 40% to be competitive with fossil-generated electricity in the Unite States. Also, the ''first concrete'' to fuel load construction schedule will have to be decreased to less than 40 months. Therefore, the U. S. Department of Energy (DOE) initiate the Nuclear Energy Research Initiative (NERI) and ABB CENP proposed a cooperative program with Sandia National Laboratory (SNL) and Duke Engineering and Services (DE and S) to begin an innovative research effort to drastically cut the cost of new nuclear power plant construction for the U. S. de-regulated market place. This program was approved by the DOE through three separate but coordinated ''cooperative agreements.'' They are the ''Risk-Informed Assessment of Regulatory and Design Requirements for Future Nuclear Power Plants'' (Risk-Informed NPP), the ''Smart Nuclear Power Plant Program'' (Smart-NPP), and ''Design, Procure, Construct, Install and Test'' (DPCIT) Program. DOE funded the three cooperative agreements at a level of $2.6 million for the first year of the program. Funding for the complete program is durrently at a level $6.9 million, however, ABB CENP and all partners anticipate that the scope of the NERI program will be increased as a result of the overall importance of NERI to the U. S. Government. The Risk-Informed NPP program, which is aimed at revising costly regularory and design requirements without reducing overall plant safety, has two basic tasks: ''development of Risk-Informed Methods'' and ''strengthening the Reliability Database.'' The overall objective of the first task is to develop a scientific, risk-informed approach for identifying and simplifying deterministic industry standards, regulatory requirements, and safety systems that do not significantly contribute to nuclear power plant reliability and safety. The second basic task is to develop a means for strengthening the reliability database
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.
Safety requirements to the operation of hydropower plants; Sicherheit beim Betrieb von Wasserkraftwerken

Energy Technology Data Exchange (ETDEWEB)

Lux, Reinhard [Berufsgenossenschaft Energie Textil Elektro Medienerzeugnisse (BG ETEM), Koeln (Germany)

2011-07-01

Employers have to take into account various safety and health requirements relating to the design, construction, operation and maintenance of hydropower plants. Especially the diversity of the hydropower plant components requires the consideration of different safety and health aspects. In 2011 the ''Fachausschuss Elektrotechnik'' (expert committee electro-technics) of the institution for statutory accident insurance and prevention presented a new ''BG-Information'' dealing with ''Safe methods operating hydropower plants''. The following article gives an introduction into the conception and the essential requirements of this new BG-Information. (orig.)
Safety and regulatory requirements of nuclear power plants

International Nuclear Information System (INIS)

Kumar, S.V.; Bhardwaj, S.A.

2000-01-01

A pre-requisite for a nuclear power program in any country is well established national safety and regulatory requirements. These have evolved for nuclear power plants in India with participation of the regulatory body, utility, research and development (R and D) organizations and educational institutions. Prevailing international practices provided a useful base to develop those applicable to specific system designs for nuclear power plants in India. Their effectiveness has been demonstrated in planned activities of building up the nuclear power program as well as with unplanned activities, like those due to safety related incidents etc. (author)
Design of control rooms and ergonomics in power plants

International Nuclear Information System (INIS)

Herbst, L.; Hinz, W.

1981-01-01

Modern power plant control rooms are characterized by automation of protection and control functions, subdivision according to functions, computer-aided information processing, and ergonomic design. Automation relieves the personnel of stress. Subdivision according to functions permits optimized procedures. Computer-aided information processing results in variable information output tailored to the actual needs. Ergonomic design assures qualified man-machine interaction. Of course, these characteristics will vary between power plants in dependence of unit power, mode of operation, and safety and availability requirements. (orig.) [de
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

Safety Design Approach for the Development of Safety Requirements for Design of Commercial HTGR

International Nuclear Information System (INIS)

Ohashi, Hirofumi; Sato, Hiroyuki; Nakagawa, Shigeaki; Tachibana, Yukio; Nishihara, Tetsuo; Yan, Xing; Sakaba, Nariaki; Kunitomi, Kazuhiko

2014-01-01

The research committee on “Safety requirements for HTGR design” was established in 2013 under the Atomic Energy Society of Japan to develop the draft safety requirements for the design of commercial High Temperature Gas-cooled Reactors (HTGRs), which incorporate the HTGR safety features demonstrated using the High Temperature Engineering Test Reactor (HTTR), lessons learned from the accident of Fukushima Daiichi Nuclear Power Station and requirements for the integration of the hydrogen production plants. The safety design approach for the commercial HTGRs which is a basement of the safety requirements is determined prior to the development of the safety requirements. The safety design approaches for the commercial HTGRs are to confine the radioactive materials within the coated fuel particles not only during normal operation but also during accident conditions, and the integrity of the coated fuel particles and other requiring physical barriers are protected by the inherent and passive safety features. This paper describes the main topics of the research committee, the safety design approaches and the safety functions of the commercial HTGRs determined in the research committee. (author)
Research requirements for improved design of reinforced concrete containment structures

International Nuclear Information System (INIS)

Banerjee, A.K.; Holley, M.J. Jr.

1978-01-01

Reinforced concrete is a competitive material for the construction of nuclear power plant containment structures. However, the designer is constrained by limited data on the behavior of certain construction details which require him to use what may be excessive rebar quantities and lead to difficult and costly construction. This paper discusses several design situations where research is recommended to increase the designer's options, to facilitate construction, and to extend the applicability of reinforced concrete to such changing containment requirements as may be imposed by an evolving nuclear technology. (Auth.)
Design aspects of water usage in the Windscale nuclear fuel reprocessing plants

International Nuclear Information System (INIS)

Wharton, J.; Bullock, M.J.

1982-01-01

The safeguard requirements of a nuclear fuel reprocessing plant place unique constraints on a designer which, in turn, affect the scope for the exercise of water economy. These constraints are examined within the context of the British Nuclear Fuels Limited reprocessing plants at Windscale and indicate the scope for water conservation. The plants and their design principles are described with particular reference to water services and usage. Progressive design development is discussed to illustrate the increasing importance of water economy. (author)
Design considerations for an integrated safeguards system for fuel-reprocessng plants

International Nuclear Information System (INIS)

Cartan, F.O.

1982-05-01

This report presents design ideas for safeguards systems in nuclear fuels reprocessing plants. The report summarizes general safeguards requirements and describes a safeguards system concept being developed and tested at the Idaho Chemical Processing Plant. The report gives some general concepts intended for design consideration and a checklist of specific problems that should be considered. The report is intended as an aid for the safeguards system designer and as a source of useful information
Design of a chemical batch plant : a study of dedicated parallel lines with intermediate storage and the plant performance

OpenAIRE

Verbiest, Floor; Cornelissens, Trijntje; Springael, Johan

2016-01-01

Abstract: Production plants worldwide face huge challenges in satisfying high service levels and outperforming competition. These challenges require appropriate strategic decisions on plant design and production strategies. In this paper, we focus on multiproduct chemical batch plants, which are typically equipped with multiple production lines and intermediate storage tanks. First we extend the existing MI(N) LP design models with the concept of parallel production lines, and optimise the as...
Seismic considerations in the design of atomic power plants

International Nuclear Information System (INIS)

Arya, A.S.; Chandrasekaran, A.R.; Thakkar, S.K.

1975-01-01

A seismic design is one of the most important factors for the safety of nuclear power plants constructed in seismic areas. The various considerations in the design of atomic power plant structures and components to achieve high degree (near absolute) of safety during future probable earthquakes is described as follows: (a) determination of design earthquake parameters for SSE and OBE (b) fixing time history accelerograms and acceleration response spectra (c) mathematical modelling of the reactor building considering soil-structure interaction (d) deciding allowable stresses, damping factors and serviceability limits like drift, displacements and crack widths (e) tests for determining stiffness and damping characteristics of components in-situ before commissioning of plant. The main questions that arise under various items requiring further research investigations or development work are pointed out for discussion. (author)
Analysis of regulatory requirement for beyond design basis events of SMART

International Nuclear Information System (INIS)

Kim, W. S.; Seol, K. W.

2000-01-01

To enhance the safety of SMART reactor, safety and regulatory requirements associated with beyond design basis events (beyond BDE), which were developed and applied to advanced light water reactor designs, were analyzed along with a design status of passive reactor. And, based on these requirements, their applicability on the SMART design was evaluated. In the design aspect, severe accident prevention and mitigation features, containment performance, and accident management were analyzed. The evaluation results show that the requirement related to beyond DBE such as ATWS, loss of residual heat removal during shutdown operation, station blackout, fire, inter-system LOCA, and well-known events from severe accident phenomena is applicable to the SMART design. However, comprehensive approach against beyond DBE is not yet provided in the SMART design, and then it is required to designate and analyze the beyond DBE-related features. This study is expected to contribute to efforts to improve plant safety and to establish regulatory requirements for safety review
Discussion of important safety requirements for new nuclear power plants

International Nuclear Information System (INIS)

Zhang Lin; Jia Xiang; Yan Tianwen; Li Wenhong; Li Chun

2014-01-01

This paper presents the analysis of several important safety requirements and improvement direction. Technical view of security goals on site safety evaluation, internal and external events fortification, serious accident prevention and mitigation, as well as the core, containment system and instrument control system design and engineering optimization, and etc are indicated. It will be useful for new plant design, construction and safety improvement. (authors)
Development of risk benefit structural design method for innovative reactor plants

International Nuclear Information System (INIS)

Yoshio Kamishima; Tai Asayama; Yukio Takahashi; Masanori Tashimo; Hideo Machida; Yomomi Otani; Yasuharu Chuman

2005-01-01

The development of innovative nuclear plants where the energy in the future is carried out in Japan. The design method based on a risk benefit of having maintained mitigation of a risk and the improvement in economy is called for, in order to realize the national innovative nuclear plants. Main key technologies of the risk benefit structural design method are crack propagation evaluation technology and structural reliability evaluation technology. This research aims at pulling up these two technologies on an engineering practical use level. In this paper, requirements from the design of typical innovative nuclear plants and research plan are shown.(authors)
Project designing of Temelin nuclear power plant

International Nuclear Information System (INIS)

Krychtalek, Z.; Linek, V.

1989-01-01

The geological and seismic parameters are listed of the Temelin nuclear power plant. The division of the site in building zones is described. The main zones consist of the power generation unit zone with the related auxiliary buildings of hot plants and of the auxiliary buildings of the nonactive part with industrial buildings. The important buildings are interconnected with communication and technology bridges. Cooling towers and spray pools and the entrance area are part of the urbanistic design. The architectonic design of the buildings uses standard building elements and materials. The design of the buildings is based on the requirements on their function and on structural load and on the demands of maximal utilization of the type of the reinforced concrete prefab structure system. The structure is made of concrete or steel cells. The project design is based on Soviet projects. The layout is shown of the main power generation units and a section is presented of a 1,000 MW unit. (J.B.). 2 figs
Structural materials requirements for in-vessel components of fusion power plants

International Nuclear Information System (INIS)

Schaaf, B. van der

2000-01-01

The economic production of fusion energy is determined by principal choices such as using magnetic plasma confinement or generating inertial fusion energy. The first generation power plants will use deuterium and tritium mixtures as fuel, producing large amounts of highly energetic neutrons resulting in radiation damage in materials. In the far future the advanced fuels, 3 He or 11 B, determine power plant designs with less radiation damage than in the first generation. The first generation power plants design must anticipate radiation damage. Solid sacrificing armour or liquid layers could limit component replacements costs to economic levels. There is more than radiation damage resistance to determine the successful application of structural materials. High endurance against cyclic loading is a prominent requirement, both for magnetic and inertial fusion energy power plants. For high efficiency and compactness of the plant, elevated temperature behaviour should be attractive. Safety and environmental requirements demand that materials have low activation potential and little toxic effects under both normal and accident conditions. The long-term contenders for fusion power plant components near the plasma are materials in the range from innovative steels, such as reduced activation ferritic martensitic steels, to highly advanced ceramic composites based on silicon carbide, and chromium alloys. The steels follow an evolutionary path to basic plant efficiencies. The competition on the energy market in the middle of the next century might necessitate the riskier but more rewarding development of SiCSiC composites or chromium alloys
Liquid and solid rad waste treatment in advanced nuclear power plants. Application to the SBWR design

International Nuclear Information System (INIS)

Tielas Reina, M.; Asuar Alonso, O.

1994-01-01

Rad waste treatment requirements for the new generation of American advanced passive and evolutionary power plants are listed in the URD (Utility Requirements Document) of the EPRI (Electrical Power Research Institute). These requirements focus on: - Minimization of shipped solid wastes - Minimization of liquid effluents - Simplification of design and operation, with emphasis not only on waste treatment system design but also on general plant design and operation These objectives are aimed at: - Reducing and segregating wastes at source - Minimizing chemical contamination of these wastes System design simplification is completed by providing free space in the building for the use of mobile plants, either for special services not considered in the basic design or to accommodate future technical advances. (Author)
Nuclear design and technical development required for ISER plant

International Nuclear Information System (INIS)

Yokoyama, Takashi; Yamano, Naoki.

1987-01-01

The report outlines some results of a study carried out by the ISER (intrinsically safe and economical reactor) Investigation Group. In particular, nuclear design concepts are examined in relation to the fuel cycle. Discussion is also made on technical development efforts to be made for realizing an ISER. Calculation of some basic nuclear design parameters is performed and results are used to examine the reactor core and fuel for ISER. As a result, it is indicated that a high-burnup type reactor core should be used on a 4.5-batch replacement, 15 EFPM (effective full power month) scheme to optimize an ISER. For technical development, consideration is made on various tests to be performed with an experimental reactor, called ISER-E, as well as other tests to provide basic data required for demonstrating the inherent safety of ISER. The study also deals with the possibility of the application of currently available light water reactor techniques to experiments with a critical assembly, non-nuclear test loop, and the experimental reactor ISER-E. It is revealed that many of the required experiments can be carried out by using test facilities and light water reactor techniques which are currently, or will be readily, available. It is stressed that international cooperation is necessary to accomplish these tests. (Nogami, K.)
Revolution of Nuclear Power Plant Design Through Digital Technology

International Nuclear Information System (INIS)

Zhang, L.; Shi, J.; Chen, W.

2015-01-01

In the digital times, digital technology has penetrated into every industry. As the highest safety requirement standard, nuclear power industry needs digital technology more to breed high quality and efficiency. Digital power plant is derived from digital design and the digitisation of power plant transfer is an inevitable trend. This paper introduces the technical solutions and features of digital nuclear power plant construction by Shanghai Nuclear Engineering Research & Design Institute, points out the key points and technical difficulties that exist in the process of construction and can serve as references for further promoting construction of digital nuclear power plant. Digital technology is still flourishing. Although many problems will be encountered in construction, it is believed that digital technology will make nuclear power industry more safe, cost-effective and efficient. (author)
Accounting for maintenance in the design of nuclear power plants

International Nuclear Information System (INIS)

Meuwisse, C.; Martin-Mattei, C.; Hamon, L.

1997-01-01

The objective of the CIDEM project (French acronym for Design Integrating Availability, Operating Experience and Maintenance) is to control the per-kW production cost of future Electricite de France REP 2000 nuclear plants. In particular, such cost control requires accounting for maintenance and logistic support from the time of design of the future installations. This technical and economic optimization is based on assessment and comparison of possible choices in terms of materials and maintenance, and on the search for potential improvements. In the Basic Design phase, the method involves identifying reference components which are supposed highly similar to the components to be designed. In the Construction phase, it culminates in the early definition of a detailed maintenance and support plant. To be effective, the approach requires realistic tailoring of the studies to be undertaken, and the participation of multidisciplinary teams working in the framework of concurrent engineering. (author)
How to design electrical systems with central control capability for industrial plants

Energy Technology Data Exchange (ETDEWEB)

Cigolini, S.; Galati, G.; Lionetto, P.F.; Stiz, M. (Siemens, Milan (Italy) Centro Elettrotecnico Sperimentale Italiano, Milan (Italy))

1991-12-01

The modern centralized control system, incorporating microprocessors, constitutes an extremely efficacious instrument for the management of an industrial plant's electrical system and provides the performance, reliability, flexibility and safety features required by today's technologically advanced plant processes. The use of intelligent centralized control systems, capable of autonomous operation and dialoguing with industrial plant electrical systems, simplifies the design of the overall plant. This paper reviews the main design criteria for the automated systems and gives examples of some suitable commercially available intelligent systems.
Advanced Light Water Reactor Plants System 80+trademark Design Certification Program

International Nuclear Information System (INIS)

Davis, G.A.

1992-01-01

Since 1985, ABB Combustion Engineering Nuclear Power (CENP) and Duke Engineering ampersand Services, Inc. (DE ampersand S) have been developing the next generation of pressurized water reactor (PWR) plant for worldwide deployment. The goal is to make available a pre-licensed, standardized plant design that can satisfy the need for a reliable and economic supply of electricity for residential, commercial and industrial use. To ensure that such a design is available when needed, it must be based on proven technology and established licensing criteria. These requirements dictate development of nuclear technology that is advanced, yet evolutionary in nature. This has been achieved with the System 80+ Standard Plant Design
Optimal design of regional wastewater pipelines and treatment plant systems.

Science.gov (United States)

Brand, Noam; Ostfeld, Avi

2011-01-01

This manuscript describes the application of a genetic algorithm model for the optimal design of regional wastewater systems comprised of transmission gravitational and pumping sewer pipelines, decentralized treatment plants, and end users of reclaimed wastewater. The algorithm seeks the diameter size of the designed pipelines and their flow distribution simultaneously, the number of treatment plants and their size and location, the pump power, and the required excavation work. The model capabilities are demonstrated through a simplified example application using base runs and sensitivity analyses. Scaling of the proposed methodology to real life wastewater collection and treatment plants design problems needs further testing and developments. The model is coded in MATLAB using the GATOOL toolbox and is available from the authors.
Electrical systems design applications on Japanese PWR plants in light of the Fukushima Daiichi Accident

International Nuclear Information System (INIS)

Nomoto, Tsutomu

2015-01-01

After the Fukushima Daiichi nuclear power plant (1F-NPP) accident (i.e. Station Blackout), several design enhancements have been incorporated or are under considering to Mitsubishi PWR plants' design of not only operational plants' design but also new plants' design. Especially, there are several important enhancements in the area of the electrical system design. In this presentation, design enhancements related to following electrical systems/equipment are introduced; - Offsite Power System; - Emergency Power Source; - Safety-related Battery; - Alternative AC Power Supply Systems. In addition, relevant design requirements/conditions which are or will be considered in Mitsubishi PWR plants are introduced. (authors)
Seismic design features of the ACR Nuclear Power Plant

International Nuclear Information System (INIS)

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

2003-01-01

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

Osiris and SOMBRERO inertial confinement fusion power plant designs

International Nuclear Information System (INIS)

Meier, W.R.; Bieri, R.L.; Monsler, M.J.

1992-03-01

The primary objective of the of the IFE Reactor Design Studies was to provide the Office of Fusion Energy with an evaluation of the potential of inertial fusion for electric power production. The term reactor studies is somewhat of a misnomer since these studies included the conceptual design and analysis of all aspects of the IFE power plants: the chambers, heat transport and power conversion systems, other balance of plant facilities, target systems (including the target production, injection, and tracking systems), and the two drivers. The scope of the IFE Reactor Design Studies was quite ambitious. The majority of our effort was spent on the conceptual design of two IFE electric power plants, one using an induction linac heavy ion beam (HIB) driver and the other using a Krypton Fluoride (KrF) laser driver. After the two point designs were developed, they were assessed in terms of their (1) environmental and safety aspects; (2) reliability, availability, and maintainability; (3) technical issues and technology development requirements; and (4) economics. Finally, we compared the design features and the results of the assessments for the two designs
The effect of regulatory requirements on the control and instrumentation system designer

International Nuclear Information System (INIS)

Golder, J.A.

1978-01-01

The difficulties encountered by the designer of control and protection systems for nuclear plant in attempting to satisfy the large number of imprecise regulations and recommendations which exist are described. The absence of fundamental quantitative safety requirements of international acceptability is deplored and the adoption of a major incident criteria expressed in quantitative terms as the basis for the derivation of target design criteria for protection systems and plant components is suggested. (author)
Advanced plant design recommendations from Cook Nuclear Plant experience

International Nuclear Information System (INIS)

Zimmerman, W.L.

1993-01-01

A project in the American Electric Power Service Corporation to review operating and maintenance experience at Cook Nuclear Plant to identify recommendations for advanced nuclear plant design is described. Recommendations so gathered in the areas of plant fluid systems, instrument and control, testing and surveillance provisions, plant layout of equipment, provisions to enhance effective maintenance, ventilation systems, radiological protection, and construction, are presented accordingly. An example for a design review checklist for effective plant operations and maintenance is suggested
Safety design requirements for safety systems and components of JSFR

International Nuclear Information System (INIS)

Kubo, Shigenobu; Shimakawa, Yoshio; Yamano, Hidemasa; Kotake, Shoji

2011-01-01

Safety design requirements for JSFR were summarized taking the development targets of the FaCT project and design feature of JSFR into account. The related safety principle and requirements for Monju, CRBRP, PRISM, SPX, LWRs, IAEA standards, goals of GIF, basic principle of INPRO etc. were also taken into account so that the safety design requirements can be a next-generation global standard. The development targets for safety and reliability are set based on those of FaCT, namely, ensuring safety and reliability equal to future LWR and related fuel cycle facilities. In order to achieve these targets, the defence-in-depth concept is used as the basic safety design principle. General features of the safety design requirements are 1) Achievement of higher reliability, 2) Achievement of higher inspectability and maintainability, 3) Introduction of passive safety features, 4) Reduction of operator action needs, 5) Design consideration against Beyond Design Basis Events, 6) In-Vessel Retention of degraded core materials, 7) Prevention and mitigation against sodium chemical reactions, and 8) Design against external events. The current specific requirements for each system and component are summarized taking the basic design concept of JSFR into account, which is an advanced loop-type large-output power plant with a mixed-oxide-fuelled core. (author)
Design of the reactor coolant system and associated systems in nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2008-01-01

This Safety Guide was prepared under the IAEA programme for establishing safety standards for nuclear power plants. The basic requirements for the design of safety systems for nuclear power plants are established in the Safety Requirements publication, Safety Standards Series No. NS-R-1 on Safety of Nuclear Power Plants: Design, which it supplements. This Safety Guide describes how the requirements for the design of the reactor coolant system (RCS) and associated systems in nuclear power plants should be met. 1.2. This publication is a revision and combination of two previous Safety Guides, Safety Series No. 50-SG-D6 on Ultimate Heat Sink and Directly Associated Heat Transport Systems for Nuclear Power Plants (1981), and Safety Series No. 50-SG-D13 on Reactor Coolant and Associated Systems in Nuclear Power Plants (1986), which are superseded by this new Safety Guide. 1.3. The revision takes account of developments in the design of the RCS and associated systems in nuclear power plants since the earlier Safety Guides were published in 1981 and 1986, respectively. The other objectives of the revision are to ensure consistency with Ref., issued in 2000, and to update the technical content. In addition, an appendix on pressurized heavy water reactors (PHWRs) has been included
Advanced design nuclear power plants: Competitive, economical electricity. An analysis of the cost of electricity from coal, gas and nuclear power plants

International Nuclear Information System (INIS)

1992-06-01

This report presents an updated analysis of the projected cost of electricity from new baseload power plants beginning operation around the year 2000. Included in the study are: (1) advanced-design, standardized nuclear power plants; (2) low emissions coal-fired power plants; (3) gasified coal-fired power plants; and (4) natural gas-fired power plants. This analysis shows that electricity from advanced-design, standardized nuclear power plants will be economically competitive with all other baseload electric generating system alternatives. This does not mean that any one source of electric power is always preferable to another. Rather, what this analysis indicates is that, as utilities and others begin planning for future baseload power plants, advanced-design nuclear plants should be considered an economically viable option to be included in their detailed studies of alternatives. Even with aggressive and successful conservation, efficiency and demand-side management programs, some new baseload electric supply will be needed during the 1990s and into the future. The baseload generating plants required in the 1990s are currently being designed and constructed. For those required shortly after 2000, the planning and alternatives assessment process must start now. It takes up to ten years to plan, design, license and construct a new coal-fired or nuclear fueled baseload electric generating plant and about six years for a natural gas-fired plant. This study indicates that for 600-megawatt blocks of capacity, advanced-design nuclear plants could supply electricity at an average of 4.5 cents per kilowatt-hour versus 4.8 cents per kilowatt-hour for an advanced pulverized-coal plant, 5.0 cents per kilowatt-hour for a gasified-coal combined cycle plant, and 4.3 cents per kilowatt-hour for a gas-fired combined cycle combustion turbine plant
Integrated structural design of nuclear power plants for high seismic areas

International Nuclear Information System (INIS)

Rieck, P.J.

1979-01-01

A design approach which structurally interconnects NPP buildings to be located in high seismic areas is described. The design evolution of a typical 600 MWe steel cylindrical containment PWR is described as the plant is structurally upgraded for higher seismic requirements, while maintaining the original plant layout. The plant design is presented as having separate reactor building and auxiliary structures for a low seismic area (0.20 g) and is structurally combined at the foundation for location in a higher seismic area (0.30 g). The evolution is completed by a fully integrated design which structurally connects the reactor building and auxiliary structures at superstructure elevations as well as foundation levels for location in very severe seismic risk areas (0.50 g). (orig.)
Research needs and improvement of standards for nuclear power plant design

International Nuclear Information System (INIS)

Chen, C.; Moreadith, F.L.

1978-01-01

The need for research and improvement of code requirements, for both economy and safety reasons is discussed for the following topics relevant to nuclear power plant structural analysis: Earthquake definition; dynamic behavior of reinforced concrete structures under impact loads; design for postulated pipe rupture; code requirements for loading combinations for concrete structures, reinforcing steel splicing, reinforced concrete structural design for thermal effects. (Auth.)
Design Provisions for Withstanding Station Blackout at Nuclear Power Plants

International Nuclear Information System (INIS)

2015-08-01

International operating experience has shown that the loss of off-site power supply concurrent with a turbine trip and unavailability of the standby alternating current power system is a credible event. Lessons learned from the past and recent station blackout events, as well as the analysis of the safety margins performed as part of the ‘stress tests’ conducted on European nuclear power plants in response to the Fukushima Daiichi accident, have identified the station blackout event as a limiting case for most nuclear power plants. The magnitude 9.0 earthquake and consequential tsunami which occurred in Fukushima, Japan, in March 2011, led to a common cause failure of on-site alternating current electrical power supply systems at the Fukushima Daiichi nuclear power plant as well as the off-site power grid. In addition, the resultant flooding caused the loss of direct current power supply, which further exacerbated an already critical situation at the plant. The loss of electrical power resulted in the meltdown of the core in three reactors on the site and severely restricted heat removal from the spent fuel pools for an extended period of time. The plant was left without essential instrumentation and controls, and this made accident management very challenging for the plant operators. The operators attempted to bring and maintain the reactors in a safe state without information on the vital plant parameters until the power supply was eventually restored after several days. Although the Fukushima Daiichi accident progressed well beyond the expected consequences of a station blackout, which is the complete loss of all alternating current power supplies, many of the lessons learned from the accident are valid. A failure of the plant power supply system such as the one that occurred at Fukushima Daiichi represents a design extension condition that requires management with predesigned contingency planning and operator training. The extended loss of all power at a
P.D.M.S. a cad software for the design of new power plants

International Nuclear Information System (INIS)

Le Lous, Y.

1982-01-01

P.D.M.S. (''Plant Design Management System'') is a computer based management system designed to assist the engineer, with no previous computer knowledge, to solve the problems associated with plant and piping design. The essential feature of P.D.M.S. is that it provides the user with the ability to create a 3D model of his complete plant, by making use of a graphic terminal connected to a computer. The system gives the engineer the powerful advantage over existing techniques that any part of the plant information, which may be required for a specific function, may be retrieved and presented to him in the form most suited to his requirements (i.e. lists of items or fully annotated drawings). P.D.M.S. incorporates advanced facilities to enable engineers to analyse the information for design accuracy and consistency. The project manager can ensure that no errors in the total design due to integration of disciplines within the project, or due to the amalgamation of the work of many designers, who possibly operate in different design centres. P.D.M.S., implemented on an IBM machine of the computer center of Clamart, is being used by the equipment Direction of EDF for the design of new power plants [fr
A proposal for safety design philosophy of HTGR for coupling hydrogen production plant

International Nuclear Information System (INIS)

Sato, Hiroyuki; Ohashi, Hirofumi; Tazawa, Yujiro; Imai, Yoshiyuki; Nakagawa, Shigeaki; Tachibana, Yukio; Kunitomi, Kazuhiko

2013-06-01

Japan Atomic Energy Agency (JAEA) has been conducting research and development for hydrogen production utilizing heat from High Temperature Gas-cooled Reactors (HTGRs). Towards the realization of nuclear hydrogen production, coupled hydrogen production plants should not be treated as an extension of a nuclear plant in order to open the door for the entry of non-nuclear industries as well as assuring reactor safety against postulated abnormal events initiated in the hydrogen production plants. Since hydrogen production plant utilizing nuclear heat has never been built in the world, little attention has been given to the establishment of a safety design for such system including the High Temperature engineering Test Reactor (HTTR). In the present study, requirements in order to design, construct and operate hydrogen production plants under conventional chemical plant standards are identified. In addition, design considerations for safety design of nuclear facility are suggested. Furthermore, feasibility of proposed safety design and design considerations are evaluated. (author)
Safety requirement of the nuclear power plants, after TMI-2 accident and their possible implementation on Bushehr NPP

International Nuclear Information System (INIS)

Mirhabibi, N.; Tochai, M.T.M.; Ashrafi, A.; Farnoudi, E.

1985-01-01

Based on the lessons learned from the TMI-2 accident and other research and developments, many improvements have been required for the design, manufacturing and operation of nuclear power plants in recent years. These requirements have already been implemented to the plants in operation and considered as new safety requirements for new plants. In the present paper these requirements and their possible implementation on Bushehr NPP are discussed. (Author)
Man-machine considerations in nuclear power plant control room design

International Nuclear Information System (INIS)

Tennant, D.V.

1987-01-01

Although human factors is a subject that has been around for a number of years, this area of design has only recently become known to the power industry. As power plants have grown in size and complexity, the instrumentation required to control and monitor plant processes has increased tremendously. This has been especially true in nuclear power facilities. Although operators are better trained and qualified, very little consideration has been devoted to man-machine interface and the limitations of human operators. This paper explores the historic aspects and design philosophy associated with nuclear plant control rooms. Current problems and solutions are explored along with the components of a control room review. Finally, a survey of future advances in control room design are offered. This paper is concerned with instrumentation, controls, and displays
Methodology for Plantwide Design and Optimization of Wastewater Treatment Plants

DEFF Research Database (Denmark)

Maria Dragan, Johanna; Zubov, Alexandr; Sin, Gürkan

2017-01-01

Design of Wastewater Treatment Plants (WWTPs) is a complex engineering task which requires integration of knowledge and experience from environmental biotechnology, process engineering, process synthesis and design as well as mathematical programming. A methodology has been formulated and applied...... for the systematic analysis and development of plantwide design of WWTPs using mathematical optimization and statistical methods such as sensitivity and uncertainty analyses....
Balance of Plant Requirements for a Nuclear Hydrogen Plant

Energy Technology Data Exchange (ETDEWEB)

Bradley Ward

2006-04-01

This document describes the requirements for the components and systems that support the hydrogen production portion of a 600 megawatt thermal (MWt) Next Generation Nuclear Plant (NGNP). These systems, defined as the "balance-of-plant" (BOP), are essential to operate an effective hydrogen production plant. Examples of BOP items are: heat recovery and heat rejection equipment, process material transport systems (pumps, valves, piping, etc.), control systems, safety systems, waste collection and disposal systems, maintenance and repair equipment, heating, ventilation, and air conditioning (HVAC), electrical supply and distribution, and others. The requirements in this document are applicable to the two hydrogen production processes currently under consideration in the DOE Nuclear Hydrogen Initiative. These processes are the sulfur iodide (S-I) process and the high temperature electrolysis (HTE) process. At present, the other two hydrogen production process - the hybrid sulfur-iodide electrolytic process (SE) and the calcium-bromide process (Ca-Br) -are under flow sheet development and not included in this report. While some features of the balance-of-plant requirements are common to all hydrogen production processes, some details will apply only to the specific needs of individual processes.
Evaluation of the applicability of existing nuclear power plant regulatory requirements in the U.S. to advanced small modular reactors.

Energy Technology Data Exchange (ETDEWEB)

LaChance, Jeffrey L.; Wheeler, Timothy A.; Farnum, Cathy Ottinger; Middleton, Bobby D.; Jordan, Sabina Erteza; Duran, Felicia Angelica; Baum, Gregory A.

2013-05-01

The current wave of small modular reactor (SMR) designs all have the goal of reducing the cost of management and operations. By optimizing the system, the goal is to make these power plants safer, cheaper to operate and maintain, and more secure. In particular, the reduction in plant staffing can result in significant cost savings. The introduction of advanced reactor designs and increased use of advanced automation technologies in existing nuclear power plants will likely change the roles, responsibilities, composition, and size of the crews required to control plant operations. Similarly, certain security staffing requirements for traditional operational nuclear power plants may not be appropriate or necessary for SMRs due to the simpler, safer and more automated design characteristics of SMRs. As a first step in a process to identify where regulatory requirements may be met with reduced staffing and therefore lower cost, this report identifies the regulatory requirements and associated guidance utilized in the licensing of existing reactors. The potential applicability of these regulations to advanced SMR designs is identified taking into account the unique features of these types of reactors.
Job training planning and design for process plant operators

International Nuclear Information System (INIS)

Wirstad, J.

1983-01-01

A method is presented by which process plant operators for nuclear power plants are trained in Sweden. It works by a top-down method of systems analysis which can be integrated into the analysis, specification, and design of the process automation system. The training methods can also be adapted to existing automation systems and operating schedules. The author's method is based on the principle that training programs should be based on job requirements, e.g. operator tasks in common, less frequent, and rare operating conditions. Procedures have been tested for the following steps: Job analysis, analysis of knowledge and experience required, analysis of operator training requirements, set-up and organisation of the training programme, achievement control, evaluation of the training programme. (orig./HP) [de
Job training planning and design for process plant operators

Energy Technology Data Exchange (ETDEWEB)

Wirstad, J.

1983-01-01

A method is presented by which process plant operators for nuclear power plants are trained in Sweden. It works by a top-down method of systems analysis which can be integrated into the analysis, specification, and design of the process automation system. The training methods can also be adapted to existing automation systems and operating schedules. The author's method is based on the principle that training programs should be based on job requirements, e.g. operator tasks in common, less frequent, and rare operating conditions. Procedures have been tested for the following steps: Job analysis, analysis of knowledge and experience required, analysis of operator training requirements, set-up and organisation of the training programme, achievement control, evaluation of the training programme.
Design control for standard U.S. EPRTM plants

International Nuclear Information System (INIS)

Mathews, Toney A.; Miller, Matthew P.

2009-01-01

The U.S. EPR TM design is being reviewed by the U.S. Nuclear Regulatory Commission (NRC) for reference by utility applicants to build and operate EPR TM nuclear reactors in the United States. While the U.S. EPR TM Design Certification and utility Combined License Applications are being reviewed by the NRC, the AREVA-Bechtel Consortium for Engineering Procurement and Construction is proceeding with developing the detailed design. Multiple, parallel regulatory and engineering activities require carefully prepared documents and rigorous design control processes. This paper will review the design control processes used by the AREVA-Bechtel Consortium. Design control must consider the basic design processes required to achieve an integrated, functional design, as well as design change control. Sources of change and the need to keep design bases and licensing bases consistent must be thoroughly understood. An objective of the U.S. EPR TM reactor deployment program for the United States is to achieve maximum standardization of common features of the plant. Such standardization is necessary for economics, speed-of-construction, and operational efficiencies available from a 'fleet' approach to deployment. (author)
External Events Excluding Earthquakes in the Design of Nuclear Power Plants. Safety Guide

International Nuclear Information System (INIS)

2008-01-01

This Safety Guide provides recommendations and guidance on design for the protection of nuclear power plants from the effects of external events (excluding earthquakes), i.e. events that originate either off the site or within the boundaries of the site but from sources that are not directly involved in the operational states of the nuclear power plant units. In addition, it provides recommendations on engineering related matters in order to comply with the safety objectives and requirements established in the IAEA Safety Requirements publication, Safety of Nuclear Power Plants: Design. It is also applicable to the design and safety assessment of items important to the safety of land based stationary nuclear power plants with water cooled reactors. Contents: 1. Introduction; 2. Application of safety criteria to the design; 3. Design basis for external events; 4. Aircraft crash; 5. External fire; 6. Explosions; 7. Asphyxiant and toxic gases; 8. Corrosive and radioactive gases and liquids; 9. Electromagnetic interference; 10. Floods; 11. Extreme winds; 12. Extreme meteorological conditions; 13. Biological phenomena; 14. Volcanism; 15. Collisions of floating bodies with water intakes and UHS components; Annex I: Aircraft crashes; Annex II: Detonation and deflagration; Annex III: Toxicity limits.

Optimal design of an activated sludge plant: theoretical analysis

Science.gov (United States)

Islam, M. A.; Amin, M. S. A.; Hoinkis, J.

2013-06-01

The design procedure of an activated sludge plant consisting of an activated sludge reactor and settling tank has been theoretically analyzed assuming that (1) the Monod equation completely describes the growth kinetics of microorganisms causing the degradation of biodegradable pollutants and (2) the settling characteristics are fully described by a power law. For a given reactor height, the design parameter of the reactor (reactor volume) is reduced to the reactor area. Then the sum total area of the reactor and the settling tank is expressed as a function of activated sludge concentration X and the recycled ratio α. A procedure has been developed to calculate X opt, for which the total required area of the plant is minimum for given microbiological system and recycled ratio. Mathematical relations have been derived to calculate the α-range in which X opt meets the requirements of F/ M ratio. Results of the analysis have been illustrated for varying X and α. Mathematical formulae have been proposed to recalculate the recycled ratio in the events, when the influent parameters differ from those assumed in the design.
Human factor engineering applied to nuclear power plant design

International Nuclear Information System (INIS)

Manrique, A.; Valdivia, J.C.; Jimenez, A.

2001-01-01

For the design and construction of new nuclear power plants as well as for maintenance and operation of the existing ones new man-machine interface designs and modifications are been produced. For these new designs Human Factor Engineering must be applied the same as for any other traditional engineering discipline. Advantages of implementing adequate Human Factor Engineering techniques in the design of nuclear reactors have become not only a fact recognized by the majority of engineers and operators but also an explicit requirement regulated and mandatory for the new designs of the so called advanced reactors. Additionally, the big saving achieved by a nuclear power plant having an operating methodology which significantly decreases the risk of operating errors makes it necessary and almost vital its implementation. The first step for this is preparing a plan to incorporate all the Human Factor Engineering principles and developing an integral design of the Instrumentation and Control and Man-machine interface systems. (author)
An overview of some basic design features of Koeberg Nuclear Power Station highlighting how regulatory requirements can influence design

International Nuclear Information System (INIS)

Morrison, A.R.

1985-01-01

The paper attempts to show that licensing requirements significantly influence the design of nuclear power plants. The French designed Pressurised Water Reactor system adopted by Escom at Koeberg has its origins in the General Design Criteria set out in the American Code of Federal Regulations document 10CFR50. Three of the General Design Criteria have been selected for illustrating how the requirements have influenced Koeberg in terms of design, both from a hardware and software view point. The requirements of the criteria on quality standard and records are to a certain extent reflected in the Licensing Branch Guide developed by the Atomic Energy Corporation to address quality assurance. The criterion on containment design sets requirements in respect of containment design which are incorporated in the Koeberg design. The criterion on electric power systems sets many of the basic design requirements for the electrical power supply systems inside and outside the station. The existence of the criterion led Escom to introduce changes in the transmission network to meet the requirements in respect of the independent criteria for the grid connections
Design Provisions for Station Blackout at Nuclear Power Plants

International Nuclear Information System (INIS)

Duchac, Alexander

2015-01-01

A station blackout (SBO) is generally known as 'a plant condition with complete loss of all alternating current (AC) power from off-site sources, from the main generator and from standby AC power sources important to safety to the essential and nonessential switchgear buses. Direct current (DC) power supplies and un-interruptible AC power supplies may be available as long as batteries can supply the loads. Alternate AC power supplies are available'. A draft Safety Guide DS 430 'Design of Electrical Power Systems for Nuclear Power Plants' provides recommendations regarding the implementation of Specific Safety Requirements: Design: Requirement 68 for emergency power systems. The Safety Guide outlines several design measures which are possible as a means of increasing the capability of the electrical power systems to cope with a station blackout, without providing detailed implementation guidance. A committee of international experts and advisors from numerous countries is currently working on an IAEA Technical Document (TECDOC) whose objective is to provide a common international technical basis from which the various criteria for SBO events need to be established, to support operation under design basis and design extension conditions (DEC) at nuclear power plants, to document in a comprehensive manner, all relevant aspects of SBO events at NPPs, and to outline critical issues which reflect the lessons learned from the Fukushima Dai-ichi accident. This paper discusses the commonly encountered difficulties associated with establishing the SBO criteria, shares the best practices, and current strategies used in the design and implementation of SBO provisions and outline the structure of the IAEA's SBO TECDOC under development. (author)
Experience of upgrading existing Russian designed nuclear plants

International Nuclear Information System (INIS)

Yanev, P.I.; Facer, R.I.

1993-01-01

From the reviewed experiences of upgrading existing Russian designed nuclear plants both of WWER and RBMK type the conclusions drawn are as follows. For the countries operating Russian designed plants it is necessary to adopt a pragmatic approach where all changes must be demonstrated to improve the safety of the plant and safety must be demonstrably improving. Care must be taken to avoid the pitfalls of excessive regulatory demands which are not satisfied and the development of an attitude of disregarding requirements on the basis that they are not enforced. For the lending countries and organizations, it is necessary to ensure that assistance is given to the operating organizations so that the most effective use of funds can be achieved. The experience in the West is that over-regulation and excessive expenditure do not necessarily lead to improved safety. They can lead to significant waste of resources. The use of western technology is recommended but where it is necessary and where it provides the greatest benefit
General requirements for pressure-retaining systems and components in CANDU nuclear power plants

International Nuclear Information System (INIS)

1991-11-01

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

DEFF Research Database (Denmark)

Schleisner, Liselotte; Hamacher, T.; Cabal, H.

2001-01-01

The energy and material necessary to construct a power plant and the land covered by the plant are indicators for the ‘consumption’ of environment by a certain technology. Based on current knowledge, estimations show that the material necessary to construct a fusion plant will exceed the material...... requirement of a fission plant by a factor of two. The material requirement for a fusion plant is roughly 2000 t/MW and little less than 1000 t/MW for a fission plant. The land requirement for a fusion plant is roughly 300 m2/MW and the land requirement for a fission plant is a little less than 200 m2/MW...... less ‘environment’ for the construction than renewable technologies, especially wind and solar....
Discussion on several important safety requirements for the new nuclear power plant

International Nuclear Information System (INIS)

Yan Tianwen; Li Jigen; Zhang Lin; Feng Youcai; Jia Xiang; Li Wenhong

2013-01-01

Post the Fukushima nuclear accident, the Chinese government raised higher safety goals and safety requirements for the new nuclear power plant to be constructed. The paper expounded the important indicators of safety requirements and the aspects of safety modification that had been developed for the new NPPs. It also discussed and analyzed the main fields required by the new NPPs safety requirements in the safety goals, safety evaluation of sites, defenses of internal and external events, severe accident prevention and mitigation, design of reactor core, containment system and I and C system, and optimization of engineering measure, which gave some references to the design, construction and safety modifications of new NPPs in China. (authors)
Some engineering considerations when designing centrifuge enrichment plants

International Nuclear Information System (INIS)

Edwards, T.T.

1982-01-01

A review is given of the three main areas where flexibility is needed in the design of centrifuge enrichment plants. These are: the need to cope with market requirements, the limitations imposed by currently available centrifuges and ever advancing centrifuge technology. Details of BNFL's experience with centrifuge enrichment at Capenhurst are presented. (U.K.)
The Design and Manufacturing of Essential oil Distillation Plant for ...

African Journals Online (AJOL)

Choice-Academy

The paper presents economic value of the design and manufacturing of essential oil production plant ... system with the required precision for standard quality of oil at affordable cost. Thus, the ..... still, steam injection and distribution systems,.
Westinghouse AP1000 advanced passive plant: design features and benefits

International Nuclear Information System (INIS)

Walls, S.J.; Cummins, W.E.

2003-01-01

The Westinghouse AP1000 Program is aimed at implementing the AP1000 plant to provide a further major improvement in plant economics while maintaining the passive safety advantages established by the AP600. An objective is to retain to the maximum extent possible the plant design of the AP600 so as to retain the licensing basis, cost estimate, construction schedule, modularization scheme, and the detailed design from the AP600 program. Westinghouse and the US Nuclear Regulatory Commission staff have embarked on a program to complete Design Certification for the AP1000 by 2004. A pre-certification review phase was completed in March 2002 and was successful in establishing the applicability of the AP600 test program and AP600 safety analysis codes to the AP1000 Design Certification. On March 28, 2002, Westinghouse submitted to US NRC the AP1000 Design Control Document and Probabilistic Risk Assessment, thereby initiating the formal design certification review process. The results presented in these documents verify the safety performance of the API 000 and conformance with US NRC licensing requirements. Plans are being developed for implementation of a series of AP1000 plants in the US. Key factors in this planning are the economics of AP1000, and the associated business model for licensing, constructing and operating these new plants. Similarly plans are being developed to get the AP1000 design reviewed for use in the UK. Part of this planning has been to examine the AP1000 design relative to anticipated UK safety and licensing issues. (author)
A framework for regulatory requirements and industry standards for new nuclear power plants

International Nuclear Information System (INIS)

Duran, Felicia A.; Camp, Allen L.; Apostolakis, George E.; Golay, Michael W.

2000-01-01

This paper summarizes the development of a framework for risk-based regulation and design for new nuclear power plants. Probabilistic risk assessment methods and a rationalist approach to defense in depth are used to develop a framework that can be applied to identify systematically the regulations and standards required to maintain the desired level of safety and reliability. By implementing such a framework, it is expected that the resulting body of requirements will provide a regulatory environment that will ensure protection of the public, will eliminate the burden of requirements that do not contribute significantly to safety, and thereby will improve the market competitiveness of new plants. (author)
Requirements for the support power systems of CANDU nuclear power plants

International Nuclear Information System (INIS)

1990-08-01

This Standard covers principal criteria and requirements for design, fabrication, installation, qualification, inspection, and documentation for assurance that support power will be available as required. The minimum requirements for support power are determined by the special safety systems and other safety-related systems that must function to ensure that the public health risk is acceptably low. Support power systems of a CANDU nuclear power plant include those parts of the electrical systems and instrument air systems that are necessary for the operation of safety-related systems
Improvements in nuclear plant staffing resulting from the AP600 design programme

International Nuclear Information System (INIS)

Mycoff, C.

2001-01-01

The staffing for a single-unit AP600 is estimated to require a staff for operation and maintenance about 32% smaller than current generation power plants of similar size. These staffing reductions are driven primarily by various features incorporated into the AP600 plant design. (author)
Multi-purpose hydrogen isotopes separation plant design

Energy Technology Data Exchange (ETDEWEB)

Boniface, H.A.; Gnanapragasam, N.V.; Ryland, D.K.; Suppiah, S.; Castillo, I. [Atomic Energy of Canada Limited - AECL, Chalk River, ON (Canada)

2015-03-15

There is a potential interest at AECL's Chalk River Laboratories to remove tritium from moderately tritiated light water and to reclaim tritiated, downgraded heavy water. With only a few limitations, a single CECE (Combined Electrolysis and Catalytic Exchange) process configuration can be designed to remove tritium from heavy water or light water and upgrade heavy water. Such a design would have some restrictions on the nature of the feed-stock and tritium product, but could produce essentially tritium-free light or heavy water that is chemically pure. The extracted tritium is produced as a small quantity of tritiated heavy water. The overall plant capacity is fixed by the total amount of electrolysis and volume of catalyst. In this proposal, with 60 kA of electrolysis a throughput of 15 kg*h{sup -1} light water for detritiation, about 4 kg*h{sup -1} of heavy water for detritiation and about 27 kg*h{sup -1} of 98% heavy water for upgrading can be processed. Such a plant requires about 1,000 liters of AECL isotope exchange catalyst. The general design features and details of this multi-purpose CECE process are described in this paper, based on some practical choices of design criteria. In addition, we outline the small differences that must be accommodated and some compromises that must be made to make the plant capable of such flexible operation. (authors)
Waste Receiving and Packaging, Module 2A, Supplemental Design Requirements Document

International Nuclear Information System (INIS)

Lamberd, D.L.; Boothe, G.F.; Hinkle, A.L.; Horgos, R.M.; LeClair, M.D.; Nash, C.R.; Ocampo, V.P.; Pauly, T.R.; Stroup, J.L.; Weingardt, K.M.

1994-01-01

The Supplemental Design Requirements Document (SDRD) is used to communicate plant design information from Westinghouse Hanford Company (WHC) to the US Department of Energy (DOE) and the cognizant Architect Engineer (A/E). Information in the SDRD serves two purposes: to convey design requirements that are too detailed for inclusion in a Functional Design Criteria (FDC) report; and to serve as a means of change control for design commitments in the Conceptual Design Report. The mission of WRAP 2A on the Hanford site is the treatment of contact handled low level mixed waste (MW) for final disposal. The overall systems engineering steps used to reach construction and operation of WRAP 2A are depicted in Figure 1. The WRAP 2A SDRD focuses on the requirements to address the functional analysis provided in Figure 1. This information is provided in sections 2 through 5 of this SDRD. The mission analysis and functional analysis are to be provided in a separate supporting document. The organization of sections 2 through 5 corresponds to the requirements identified in the WRAP 2A functional analysis
Preliminary Design Requirements Document for Project W-314

Energy Technology Data Exchange (ETDEWEB)

MCGREW, D.L.

2000-04-27

This document sets forth functional requirements, performance requirements, and design constraints for the tank farm systems elements identified in Section 3.1 of this document. These requirements shall be used to develop the Design Requirements Baseline for those system elements. System Overview--The tank farm system at Hanford Site currently consists of 149 single shell tanks and 28 double shell tanks with associated facilities and equipment, located in 18 separate groupings. Each grouping is known as a tank farm. They are located in the areas designated as 200 West and 200 East. Table 1-1 shows the number of tanks in each farm. The farms are connected together through a transfer system consisting of piping, diversion boxes, Double Contained Receiver Tanks (DCRT) and other miscellaneous facilities and elements. The tank farm system also connects to a series of processing plants which generate radioactive and hazardous wastes. The primary functions of the tank farm system are to store, transfer, concentrate, and characterize radioactive and hazardous waste generated at Hanford, until the waste can be safely retrieved, processed and dispositioned. The systems provided by Project W-314 support the store and transfer waste functions. The system elements to be upgraded by Project W-314 are identified in Section 3.1.
Preliminary Design Requirements Document for Project W-314

International Nuclear Information System (INIS)

MCGREW, D.L.

2000-01-01

This document sets forth functional requirements, performance requirements, and design constraints for the tank farm systems elements identified in Section 3.1 of this document. These requirements shall be used to develop the Design Requirements Baseline for those system elements. System Overview--The tank farm system at Hanford Site currently consists of 149 single shell tanks and 28 double shell tanks with associated facilities and equipment, located in 18 separate groupings. Each grouping is known as a tank farm. They are located in the areas designated as 200 West and 200 East. Table 1-1 shows the number of tanks in each farm. The farms are connected together through a transfer system consisting of piping, diversion boxes, Double Contained Receiver Tanks (DCRT) and other miscellaneous facilities and elements. The tank farm system also connects to a series of processing plants which generate radioactive and hazardous wastes. The primary functions of the tank farm system are to store, transfer, concentrate, and characterize radioactive and hazardous waste generated at Hanford, until the waste can be safely retrieved, processed and dispositioned. The systems provided by Project W-314 support the store and transfer waste functions. The system elements to be upgraded by Project W-314 are identified in Section 3.1
General design safety principles for nuclear power plants

International Nuclear Information System (INIS)

1986-01-01

This Safety Guide provides the safety principles and the approach that have been used to implement the Code in the Safety Guides. These safety principles and the approach are tied closely to the safety analyses needed to assist the design process, and are used to verify the adequacy of nuclear power plant designs. This Guide also provides a framework for the use of other design Safety Guides. However, although it explains the principles on which the other Safety Guides are based, the requirements for specific applications of these principles are mostly found in the other Guides
Protection against internal fires and explosions in the design of nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2004-01-01

Experience of the past two decades in the operation of nuclear power plants and modern analysis techniques confirm that fire may be a real threat to nuclear safety and should receive adequate attention from the beginning of the design process throughout the life of the plant. Within the framework of the NUSS programme, a Safety Guide on fire protection had therefore been developed to enlarge on the general requirements given in the Code. Since its first publication in 1979, there has been considerable development in protection technology and analysis methods and after the Chernobyl accident it was decided to revise the existing Guide. This Safety Guide supplements the requirements established in Safety of Nuclear Power Plants: Design. It supersedes Safety Series No. 50-SG-D2 (Rev. 1), Fire Protection in Nuclear Power Plants: A Safety Guide, issued in 1992.The present Safety Guide is intended to advise designers, safety assessors and regulators on the concept of fire protection in the design of nuclear power plants and on recommended ways of implementing the concept in some detail in practice

Ex-vessel core catcher design requirements and preliminary concepts evaluation

International Nuclear Information System (INIS)

Friedland, A.J.; Tilbrook, R.W.

1974-01-01

As part of the overall study of the consequences of a hypothetical failure to scram following loss of pumping power, design requirements and preliminary concepts evaluation of an ex-vessel core catcher (EVCC) were performed. EVCC is the term applied to a class of devices whose primary objective is to provide a stable subcritical and coolable configuration within containment following a postulated accident in which it is assumed that core debris has penetrated the Reactor Vessel and Guard Vessel. Under these assumed conditions a set of functional requirements were developed for an EVCC and several concepts were evaluated. The studies were specifically directed toward the FFTF design considering the restraints imposed by the physical design and construction of the FFTF plant
Guideline on radiation protection requirements for ionizing radiation shielding in nuclear power plants

International Nuclear Information System (INIS)

1988-01-01

The guideline which entered into force on 1 May 1988 stipulates the radiation protection requirements for shielding against ionizing radiation to be met in the design, construction, commissioning, operation, and decommissioning of nuclear power plants
Current summary of international extreme load design requirements for nuclear power plant facilities

International Nuclear Information System (INIS)

Stevenson, J.D.

1980-01-01

The development of extreme load design criteria both as to rate and depth within any national jurisdiction as applied to nuclear power plant design is a function of several factors. The prime factor is the number of nuclear power plant facilities which are operating, under construction or planned in a given country. The second most important factor seems to be the degree of development of a domestic independent nuclear steam system supplier, NSSS vendor. Finally, countries whose domestic NSSS firms are active in the export market appear to have more active criteria development programs or at least they appear more visible to the foreign observer. For the purposes of this paper, extreme loads are defined as those loads having probability of occurence less than 10 -1 /yr and whose occurence could result in radiological consequences in excess of those permitted by national health standards. The specific loads considered include earthquake, extreme wind (tornado), airplane crash, detonation, and high energy system rupture. The paper identifies five national centers for extreme load criteria development; Canada, Great Britian, USA, USSR, and West Germany with both France and Japan also about to appear as independent centers of criteria development. Criteria under development by each national center are discussed in detail. (orig.)
Thermal power plant design and operation

CERN Document Server

Sarkar, Dipak

2015-01-01

Thermal Power Plant: Design and Operation deals with various aspects of a thermal power plant, providing a new dimension to the subject, with focus on operating practices and troubleshooting, as well as technology and design. Its author has a 40-long association with thermal power plants in design as well as field engineering, sharing his experience with professional engineers under various training capacities, such as training programs for graduate engineers and operating personnel. Thermal Power Plant presents practical content on coal-, gas-, oil-, peat- and biomass-fueled thermal power
Expert systems for design, operation and management of industrial plant elctrical systems

Energy Technology Data Exchange (ETDEWEB)

Delfino, B.; Forzano, P.; Invernizzi, M.; Massucco, S. (Genoa Univ. (Italy) Pavia Univ. (Italy) Ansaldo Industria, Genoa (Italy))

1991-02-01

A discussion is made of modern industrial plant requirements with regard to man-machine interfacing. Indications are then given as to the optimum hardware and software for electrical plant and process control systems. Illustrative examples are provided on the use of expert systems to aid in the design of industrial plant electrical systems and to allow safe and reliable on-line control and monitoring.
Consequences for designer and manufacturer of mechanical components due to future requirements in Europe

International Nuclear Information System (INIS)

Hans-Joachim, Frank

2001-01-01

In the frame of European harmonization, a lot of changes on requirements for designer and manufacturer of mechanical components have been performed. Differed organizations are involved in preparing future requirements for nuclear application. On one side the French German cooperation on the development of EPR. At the origin of this project was the common decision in 1989 of Framatome and Siemens to cooperate through NPI, to design the Nuclear Island, which meets the future needs of utilities. EDF and a group of the main German Utilities joined this cooperation in 1991 and since then they have been totally involved to the progress of the work. In addition, all the process was backed up to the end by the strong cooperation between the French and the German. Safety Authorities, which have a long lasting cooperation to define common requirements, which have to be applied to future Nuclear Power Plants. Furthermore an organization has been set up to elaborate common codes related to the EPR design, at the level of the French design and construction rules (RCC) or the German KTA safety standards, the so-called EPR technical codes (ETC). On the other side, the European utilities co-operate on a much broader basis for the establishment of European Utilities Requirements (EUR). These requirements are prepared by a group of European utilities that represent the major European electricity generating companies that are determined to keep the nuclear option open. The technical requirements specified in the EUR document define the boundaries in which future plants need to be designed in order to be acceptable for the needs of the utilities and in order to fulfill the basic requirements of competitive power generation costs and licensability in all countries represented in the EUR group. All the new requirements have to be applied by designer and manufacturer. Siemens /SNP act as a designer of a lot of various vessels and tanks, heat exchangers and other items of process
Advanced control room design for nuclear power plants

International Nuclear Information System (INIS)

Scarola, K.

1987-01-01

The power industry has seen a continuous growth of size and complexity of nuclear power plants. Accompanying these changes have been extensive regulatory requirements resulting in significant construction, operation and maintenance costs. In response to related concerns raised by industry members, Combustion Engineering developed the NUPLEX 80 Advanced Control Room. The goal of NUPLEX 80 TM is to: reduce design and construction costs; increase plant safety and availability through improvements in the man-machine interface; and reduce maintenance costs. This paper provides an overview of the NUPLEX 80 Advanced Control Room and explains how the stated goals are achieved. (author)
Design of the reactor coolant system and associated systems in nuclear power plants. Safety guide (Spanish Edition)

International Nuclear Information System (INIS)

2010-01-01

This Safety Guide was prepared under the IAEA programme for establishing safety standards for nuclear power plants. The basic requirements for the design of safety systems for nuclear power plants are established in the Safety Requirements publication, Safety Standards Series No. NS-R-1 on Safety of Nuclear Power Plants: Design, which it supplements. This Safety Guide describes how the requirements for the design of the reactor coolant system (RCS) and associated systems in nuclear power plants should be met. This publication is a revision and combination of two previous Safety Guides, Safety Series No. 50-SG-D6 on Ultimate Heat Sink and Directly Associated Heat Transport Systems for Nuclear Power Plants (1982), and Safety Series No. 50-SG-D13 on Reactor Coolant and Associated Systems in Nuclear Power Plants (1987), which are superseded by this new Safety Guide. The revision takes account of developments in the design of the RCS and associated systems in nuclear power plants since the earlier Safety Guides were published in 1982 and 1987, respectively. The other objectives of the revision are to ensure consistency with Ref., issued in 2004, and to update the technical content. In addition, an appendix on pressurized heavy water reactors (PHWRs) has been included.
Reality testing a plant design 'virtually' anywhere

International Nuclear Information System (INIS)

Anon.

1997-01-01

The development of a new world-wide-web compatible information system known as HyperPlant will allow users to navigate real-time three-dimensional plant design and contraction software. It is anticipated that corporate Intranets will be created to facilitate computer-aided design of industrial plants such as piping routes, process schematics, fabrication drawings, and allow use of PDMS (the Plant Design Management System). HyperPlant can also assist in plant commissioning and operation as well as for planning operation and maintenance procedures. (UK)
High-temperature gas-cooled reactor steam-cycle/cogeneration lead plant. Plant Protection and Instrumentation System design description

International Nuclear Information System (INIS)

1983-01-01

The Plant Protection and Instrumentation System provides plant safety system sense and command features, actuation of plant safety system execute features, preventive features which maintain safety system integrity, and safety-related instrumentation which monitors the plant and its safety systems. The primary function of the Plant Protection and Instrumentation system is to sense plant process variables to detect abnormal plant conditions and to provide input to actuation devices directly controlling equipment required to mitigate the consequences of design basis events to protect the public health and safety. The secondary functions of the Plant Protection and Instrumentation System are to provide plant preventive features, sybsystems that monitor plant safety systems status, subsystems that monitor the plant under normal operating and accident conditions, safety-related controls which allow control of reactor shutdown and cooling from a remote shutdown area
The computer program system for structural design of nuclear power plants

International Nuclear Information System (INIS)

Aihara, S.; Atsumi, K.; Sasagawa, K.; Satoh, S.

1979-01-01

In recent days, the design method of the Nuclear Power Plant has become more complex than in the past. The Finite Element Method (FEM) applied for analysis of Nuclear Power Plants, especially requires more computer use. The recent computers have made remarkable progress, so that in design work manpower and time necessary for analysis have been reduced considerably. However, instead the arrangement of outputs have increased tremendously. Therefore, a computer program system was developed for performing all of the processes, from data making to output arrangement, and rebar evaluations. This report introduces the computer program system pertaining to the design flow of the Reactor Building. (orig.)
Regulatory issues resolved through design certification on the System 80+trademark standard plant design

International Nuclear Information System (INIS)

Ritterbusch, S.E.; Brinkman, C.B.

1996-01-01

The US Nuclear Regulatory Commission (NRC) has completed its review of the System 80+trademark Standard Plant Design, approving advanced design features and closing severe accident licensing issues. Final Design Approval was granted in July 1994. The NRC review was extensive, requiring written responses to over 4,950 questions and formal printing of over 50,000 Safety Analysis Report pages. New safety issues never before addressed in a regulatory atmosphere had to be resolved with detailed analysis and evaluation of design features. the System 80+ review demonstrated that regulatory issues can be firmly resolved only through presentation of a detailed design and completion of a comprehensive regulatory review
Xenon lighting adjusted to plant requirements

Energy Technology Data Exchange (ETDEWEB)

Koefferlein, M.; Doehring, T.; Payer, H.D.; Seidlitz, H.K. [GSF-Forschungszentrum fuer Umwelt und Gesundheit, Oberschleissheim (Germany)

1994-12-31

The high luminous flux and spectral properties of xenon lamps would provide an ideal luminary for plant lighting if not excess IR radiation poses several problems for an application: the required filter systems reduce the irradiance at spectral regions of particular importance for plant development. Most of the economical drawbacks of xenon lamps are related to the difficult handling of that excess IR energy. Furthermore, the temporal variation of the xenon output depending on the oscillations of the applied AC voltage has to be considered for the plant development. However, xenon lamps outperform other lighting systems with respect to spectral stability, immediate response, and maximum luminance. Therefore, despite considerable competition by other lighting techniques, xenon lamps provide a very useful tool for special purposes. In plant lighting however, they seem to play a less important role as other lamp and lighting developments can meet these particular requirements at lower costs.
Evaluation of masonry wall design at nuclear power plants

International Nuclear Information System (INIS)

Con, V.N.; Subramonian, N.; Chokshi, N.

1983-01-01

The structural integrity of safety-related masonry walls in operating nuclear power plants may not be maintained when subjected to certain loads and load combinations. The paper presents some findings based upon the review of the design and analysis procedures used by the licensees in the reevaluation of safety-related masonry walls. The design criteria developed by the Structural Engineering Branch (SEB) of the United States Nuclear Regulatory Commission (NRC) along with other standard codes such as the Uniform Building Code, ACI 531-79, ATC 3-06, and NCMA were used as guidance in evaluating the design criteria developed by the licensees. The paper deals with the following subject areas: loads and load combinations, allowable stresses, analytical procedures, and modification methods. The paper concludes that, in general, the masonry walls in nuclear power plants comply with the working stress design requirements. In some cases, certain nonlinear analysis methods were used. The applicability of these methods is discussed. (orig.)
Establishment of design and performance requirements using cost and systems analysis

International Nuclear Information System (INIS)

Waganer, L.M.; Carosella, L.A.; Defreece, D.A.

1977-01-01

The current uncertainty in design approach and performance requirements for a commercial fusion power plant poses a problem for the designer in configuring the plant and for the utilities in analyzing the attractiveness of a future fusion power plant. To provide direction and insight in this area, a systems analysis model was constructed at McDonnell Douglas, utilizing fusion subsystem algorithms with subsystem cost estimating relationships into a self-consistent computerized model for several fusion reactor concepts. Cost estimating data has been compiled by utilizing McDonnell Douglas' experience in fabricating large, complex metal assemblies and soliciting the accumulated store of knowledge in existing power plants and new emerging technologies such as the Clinch River Breeder Reactor. Using the computer model, sensitivities to plasma, reactor and plant parameters are a few of the options that have been evaluated to yield recommended concepts/techniques/solutions. This is a very beneficial tool in assessing the impact of the fusion reactor on the electrical power community and charting the optimum developmental approach
Evaluation of divertor conceptual designs for a fusion power plant

International Nuclear Information System (INIS)

Ferrari, M.; Giancarli, L.; Kleefeldt, K.; Nardi, C.; Roedig, M.; Reimann, J.; Salavy, J.F.

2001-01-01

In the frame of the preliminary study of plants suitable for the energy production from the fusion power, particular emphasis has been given on the divertor studies. Since a significant percentage of the power generated from the fusion process is absorbed in the divertor, the thermal efficiency of the power conversion cycle requires a high coolant outlet temperature of the divertor, leading to solutions that are different from those adopted for the present experimental fusion plants. Therefore, copper alloys having extremely high thermal conductivity, cannot be used as structural material for this kind of devices. The most suitable coolants to be used in the divertor are water, helium and liquid metals. A conceptual design study has been developed for each of these three fluids, with the aim to evaluate the maximum allowable thermal flux at the divertor target plate and the R and D requirements for each solution. While a water-cooled divertor can be designed with a limited R and D effort, the development of helium or liquid metal cooled divertors requires a more engaging R and D program
Design of the robust synchronous generator stator voltage regulator based on the interval plant model

Directory of Open Access Journals (Sweden)

Stojić Đorđe

2013-01-01

Full Text Available In this paper a novel method for the stator voltage regulator of a synchronous generator based on the interval plant mode, is presented. Namely, it is shown in the literature that, in order to design a controller for the first-order compensator, the limited number of interval plants needs to be examined. Consequently, the intervals of the plant model parameter variations used to calculate the four extreme interval plants required for the sequential PI controller design are determined. The controller is designed using frequency-domain-based techniques, while its robust performance is examined using simulation tests.
Export Control Requirements for Tritium Processing Design and R&D

Energy Technology Data Exchange (ETDEWEB)

Hollis, William Kirk [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Maynard, Sarah-Jane Wadsworth [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-04-05

This document will address requirements of export control associated with tritium plant design and processes. Los Alamos National Laboratory has been working in the area of tritium plant system design and research and development (R&D) since the early 1970’s at the Tritium Systems Test Assembly (TSTA). This work has continued to the current date with projects associated with the ITER project and other Office of Science Fusion Energy Science (OS-FES) funded programs. ITER is currently the highest funding area for the DOE OS-FES. Although export control issues have been integrated into these projects in the past a general guidance document has not been available for reference in this area. To address concerns with currently funded tritium plant programs and assist future projects for FES, this document will identify the key reference documents and specific sections within related to tritium research. Guidance as to the application of these sections will be discussed with specific detail to publications and work with foreign nationals.
Export Control Requirements for Tritium Processing Design and R&D

Energy Technology Data Exchange (ETDEWEB)

Hollis, William Kirk [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Maynard, Sarah-Jane Wadsworth [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2015-10-30

This document will address requirements of export control associated with tritium plant design and processes. Los Alamos National Laboratory has been working in the area of tritium plant system design and research and development (R&D) since the early 1970’s at the Tritium Systems Test Assembly (TSTA). This work has continued to the current date with projects associated with the ITER project and other Office of Science Fusion Energy Science (OS-FES) funded programs. ITER is currently the highest funding area for the DOE OS-FES. Although export control issues have been integrated into these projects in the past a general guidance document has not been available for reference in this area. To address concerns with currently funded tritium plant programs and assist future projects for FES, this document will identify the key reference documents and specific sections within related to tritium research. Guidance as to the application of these sections will be discussed with specific detail to publications and work with foreign nationals.
Closing the loops between plant design and operator-An automatic logging system

International Nuclear Information System (INIS)

Tally, C.

1985-01-01

The close relationship between plant owner and NSSS designer frequently ceases after the plant is through startup testing. Thus, there is no continuous feedback between the operations staff and the designer. As a result, there is no assurance that the plant is being operated within the design envelope defined by the NSSS component stress reports. The link between plant operation and the plant design basis is vital to ensure that the plant can be safely operated for its full licensed life. This link is also a key to extending the life of the plant since the fatigue history of critical components is an important element of any justification for extended component life. An allowable Operating Transient Cycles Program established by Duke Power and Babcock and Wilcox successfully closed the operator-designer loop at the Oconee Nuclear Station. This paper describes that program, some of its conclusions, and also describes the next logical step in its development...automation of the transient logging process. A transient monitoring program must satisfy many requirements ranging from sensing the onset of a transient or slow power maneuver to recording sufficient data to provide for human checking of all computerized conclusions and results. Although not yet available to the industry, this type of program will ultimately be a virtual necessity for all nuclear stations

Balance of plant design issues for small reactors in Canada

International Nuclear Information System (INIS)

Harvel, G.; Meneley, D.

2014-01-01

Internationally, several companies are exploring design and development of Small Modular Reactors (SMR) ranging in power from 10 MWe to 300 MWe. While the designs are proceeding, the main issue at hand is finding a site for deployment of the first unit. Connection to existing well established grids is currently not competitive in part due to First of a Kind (FOAK) costs. As such, many vendors are exploring unique and remote applications where FOAK costs are not as significant a concern. One of the major assumptions in the design process usually followed is that the major effort needs to concentrate on reactor core development. While the reactor core is important, costs associated with the balance of plant and operations of the unit are likely to play an important role in the final decision of purchase. In this work, a series of conceptual designs is performed for the support systems of a small modular reactor by successive teams of undergraduate students working over semester long periods during a 3 year period. The goal of this process is to determine to what extent current technology for the balance of plant supports the development of a cost effective SMR. Each system is given to a team with an open set of criteria for design. At the completion of the design exercise, an open discussion with the teams is held regarding the staffing requirements for an SMR. The results are preliminary and reflect the open nature of the exercise. That said, the results indicate that for an SMR to be truly competitive, significant innovation is required in addressing the supporting systems of the plant. (author)
Balance of plant design issues for small reactors in Canada

Energy Technology Data Exchange (ETDEWEB)

Harvel, G.; Meneley, D., E-mail: Glenn.Harvel@uoit.ca, E-mail: dan.meneley@sympatico.ca [Univ. of Ontario Inst. of Tech.y, Oshawa, ON (Canada)

2014-07-01

Internationally, several companies are exploring design and development of Small Modular Reactors (SMR) ranging in power from 10 MWe to 300 MWe. While the designs are proceeding, the main issue at hand is finding a site for deployment of the first unit. Connection to existing well established grids is currently not competitive in part due to First of a Kind (FOAK) costs. As such, many vendors are exploring unique and remote applications where FOAK costs are not as significant a concern. One of the major assumptions in the design process usually followed is that the major effort needs to concentrate on reactor core development. While the reactor core is important, costs associated with the balance of plant and operations of the unit are likely to play an important role in the final decision of purchase. In this work, a series of conceptual designs is performed for the support systems of a small modular reactor by successive teams of undergraduate students working over semester long periods during a 3 year period. The goal of this process is to determine to what extent current technology for the balance of plant supports the development of a cost effective SMR. Each system is given to a team with an open set of criteria for design. At the completion of the design exercise, an open discussion with the teams is held regarding the staffing requirements for an SMR. The results are preliminary and reflect the open nature of the exercise. That said, the results indicate that for an SMR to be truly competitive, significant innovation is required in addressing the supporting systems of the plant. (author)
Design assurance programme for modifications at the Ringhals Nuclear Power Plant

International Nuclear Information System (INIS)

Persson, K.

1985-01-01

The Engineering Department within the site organization is responsible for modifications and improvements of plant design. Internal and external specialists are also involved in the design activities. Modifications follow the procedures laid down in the Design Assurance Programme (DAP), according to the overall directions given in 10CFR50 APP B and the more detailed recommendations in American Nuclear Standard 3.2, and the United States Nuclear Regulatory Commission's Regulatory Guides 1.33 and 1.64. The site Quality Assurance Department has adapted these directions to Swedish conditions. All plant modifications are ordered from customers within the site organization. The design work is performed according to the following steps: priority of problems or requirements; overall analysis of problems or requirements; basic design (preliminary and final); detailed design (electrical, mechanical, civil); purchase of equipment; installation in plant; testing of installed equipment. The project engineer is responsible for modifications from the initiation of analysis until testing and documentation are finalized. For each step, procedures and checklists are available. Internal and external specialists as well as the Swedish Nuclear Inspectorate are involved when appropriate. Design engineers and QC/QE engineers are also involved. Special groups within the site organization handle purchase and installation. Testing the installed equipment is performed in co-operation with the operation staff. Deviations from DAP must be reported by all personnel, and corrective actions must be taken. Internal audits have to be done once a year for the entire DAP and the audit, with its findings and the corrective actions taken, must be reported to the QA group. (author)
MHD/gas turbine systems designed for low cooling water requirements

International Nuclear Information System (INIS)

Annen, K.D.; Eustis, R.H.

1983-01-01

The MHD/gas turbine combined-cycle system has been designed specifically for applications where the availability of cooling water is very limited. The base case systems which were studied consist of a coal-fired MHD plant with an air turbine bottoming plant and require no cooling water. In addition to the base case systems, systems were considered which included the addition of a vapor cycle bottoming plant to improve the thermal efficiency. These systems require a small amount of cooling water. The results show that the MHD/gas turbine systems have very good thermal and economic performances. The base case I MHD/gas turbine system (782 MW /SUB e/ ) requires no cooling water, has a heat rate which is 13% higher, and a cost of electricity which is only 7% higher than a comparable MHD/steam system (878 MW /SUB e/ ) having a cooling tower heat load of 720 MW. The case I vapor cycle bottomed systems have thermal and economic performances which approach and even exceed those of the MHD/steam system, while having substantially lower cooling water requirements. Performances of a second-generation MHD/gas turbine system and an oxygen-enriched, early commercial system are also evaluated. An analysis of nitric oxide emissions shows compliance with emission standards
Manpower simulation for the power plant design engineering

International Nuclear Information System (INIS)

Moon, B.S.; Juhn, P.E.

1982-01-01

Some observation from the examination of actual manhour curves for the power design engineering obtained from Sargent and Lundy Engineers and of a few of the model curves proposed by Bechtel, are analyzed in this paper. A model curve representing typical design engineering manhour has been determined as probability density function for the Gamma Distribution. By means of this model curve, we strategically forecast the future engineering manpower requirements to meet the Covernment's long range nuclear power plan. As a sensitivity analysis, the directions for the localization of nuclear power plant design engineering, are studied in terms of the performance factor for the experienced versus inexperienced engineers. (Author)
Evolution of general design requirements for french pressurized water reactors

International Nuclear Information System (INIS)

Gros, G.; Jalouneix, J.; Rollinger, F.

1988-10-01

The design of French pressurized water reactors is based first on deterministic principles, using the well-known defense in depth concept. This safety approach, basically reflected current American practice at that time, which consisted notably in designing engineered safeguard systems capable of limiting the consequences of accidents assumed to be credible despite the preventive measures taken. Further reflections have led to complete this approach, resulting in modifications to regulatory practice, mainly related to better practical assimilation of the problems arising during plant unit operation and reactor control after an accident and to the determination to enhance the overall consistency of the safety approach. As regards system redundancy, it should be noted that common cause failures can result in the total loss of a redundant system. System redundancy aspects will be dealt with in Chapter 2. As regards study of design basis accidents, attention was focused on the human intervention stage following automatic activation of protection and safeguard systems. This resulted, for all plant units, in the revision of operating procedures, accompanied by examination of the means required for their implementation. These subjects will be discussed in Chapter 3. Finally, as regards equipment classification, the range of equipment subjected to particular requirements, formerly limited to design basis safety classified equipment, was enlarged to include important for safety equipment. This subject will be dealt with in Chapter 5
Osiris and SOMBRERO inertial confinement fusion power plant designs. Volume 2, Designs, assessments, and comparisons, Final report

Energy Technology Data Exchange (ETDEWEB)

Meier, W.R.; Bieri, R.L.; Monsler, M.J.

1992-03-01

The primary objective of the of the IFE Reactor Design Studies was to provide the Office of Fusion Energy with an evaluation of the potential of inertial fusion for electric power production. The term reactor studies is somewhat of a misnomer since these studies included the conceptual design and analysis of all aspects of the IFE power plants: the chambers, heat transport and power conversion systems, other balance of plant facilities, target systems (including the target production, injection, and tracking systems), and the two drivers. The scope of the IFE Reactor Design Studies was quite ambitious. The majority of our effort was spent on the conceptual design of two IFE electric power plants, one using an induction linac heavy ion beam (HIB) driver and the other using a Krypton Fluoride (KrF) laser driver. After the two point designs were developed, they were assessed in terms of their (1) environmental and safety aspects; (2) reliability, availability, and maintainability; (3) technical issues and technology development requirements; and (4) economics. Finally, we compared the design features and the results of the assessments for the two designs.
US GCFR demonstration plant design

International Nuclear Information System (INIS)

Hunt, P.S.; Snyder, H.J.

1980-05-01

A general description of the US GCFR demonstration plant conceptual design is given to provide a context for more detailed papers to follow. The parameters selected for use in the design are presented and the basis for parameter selection is discussed. Nuclear steam supply system (NSSS) and balance of plant (BOP) component arrangements and systems are briefly discussed
Design measures to facilitate implementation of safeguards at future water cooled nuclear power plants

International Nuclear Information System (INIS)

1999-01-01

The report is intended to present guidelines to the State authorities, designers and prospective purchasers of future water cooled power reactors which, if taken into account, will minimize the impact of IAEA safeguards on plant operation and ensure efficient and effective acquisition of safeguards data to the mutual benefit of the Member State, the plant operator and the IAEA. These guidelines incorporate the IAEA's experience in establishing and carrying out safeguards at currently operating nuclear power plants, the ongoing development of safeguards techniques and feedback of experience from plant operators and designers on the impact of IAEA safeguards on plant operation. The following main subjects are included: The IAEA's safeguards function for current and future nuclear power plants; summary of the political and legal foundations of the IAEA's safeguards system; the technical objective of safeguards and the supply and use of required design information; safeguards approaches for nuclear power plants; design implications of experience in safeguarding nuclear power plants and guidelines for future water cooled reactors to facilitate the implementation of safeguards
Guideline for design and construction radiation monitoring equipments for Tokai reprocessing plant

International Nuclear Information System (INIS)

Miyabe, Kenjiro; Ninomiya, Kazushige; Jin, Kazumi; Morifuji, Masayuki; Nemoto, Kazuhiko; Sato, Akira; Kawai, Keiichi

1999-12-01

Various kind of radiation monitoring equipment are used in radiation controlled area at each facility of Tokai reprocessing plant. These equipments have been designed and constructed based on the users requirements, and permitted by governmental regulation office. And, design has been carried out in consideration of the adoption of the new technology and our operational experience. Then, it has been used effectively for the radiation control of the facilities. This report summarizes the technical requirements that should be taken into consideration in the design and installation of radiation monitoring equipments. These requirements are fundamentally applicable when the equipments of the new facilities will be designed or the present instruments will be replaced. (author)
Design of plant safety model in plant enterprise engineering environment

International Nuclear Information System (INIS)

Gabbar, Hossam A.; Suzuki, Kazuhiko; Shimada, Yukiyasu

2001-01-01

Plant enterprise engineering environment (PEEE) is an approach aiming to manage the plant through its lifecycle. In such environment, safety is considered as the common objective for all activities throughout the plant lifecycle. One approach to achieve plant safety is to embed safety aspects within each function and activity within such environment. One ideal way to enable safety aspects within each automated function is through modeling. This paper proposes a theoretical approach to design plant safety model as integrated with the plant lifecycle model within such environment. Object-oriented modeling approach is used to construct the plant safety model using OO CASE tool on the basis of unified modeling language (UML). Multiple views are defined for plant objects to express static, dynamic, and functional semantics of these objects. Process safety aspects are mapped to each model element and inherited from design to operation stage, as it is naturally embedded within plant's objects. By developing and realizing the plant safety model, safer plant operation can be achieved and plant safety can be assured
The implications of plant design on the life-time costs for nuclear fuel cycle facilities

International Nuclear Information System (INIS)

Macphee, D.S.; Hexter, B.C.; Young, M.P.; Wilson, B.J.

1997-01-01

Utilising the experience gained during many years of design and project management of nuclear plant, BNFL is now approaching the final stages of the construction and commissioning of the Sellafield MOX Plant (SMP) in the UK. The paper uses the SMP project to highlight the benefits of these experiences, in particular addressing the implications of the approach to plant design on life time costs. In addition to providing BNFL with a state of the art, commercial scale MOX fuel fabrication facility, the construction of this 120 tHM/yr facility, which is currently in the advanced stages of commissioning, represents a significant demonstration of the design and project management skills of BNFL Engineering Ltd. As well as meeting the main process requirements, the plant design incorporates the highest standards of safety, together with input from the future plant operators and potential customers. As befits a commercial scale plutonium handling facility, SMP also incorporates material accountancy and security provisions that will meet all international requirements. Design, construction and commissioning of this complex and highly automated plant, has benefited from a totally integrated approach to design and documentation that considers not only project implementation but also overall lifetime costs. In addition, project management techniques, developed over many years of major project construction at Sellafield, have been utilised in order to ensure successful project implementation against a background of significant technical challenge and 'fast track' timescales. (author)
Preconceptual design studies and cost data of depleted uranium hexafluoride conversion plants

International Nuclear Information System (INIS)

Jones, E

1999-01-01

One of the more important legacies left with the Department of Energy (DOE) after the privatization of the United States Enrichment Corporation is the large inventory of depleted uranium hexafluoride (DUF6). The DOE Office of Nuclear Energy, Science and Technology (NE) is responsible for the long-term management of some 700,000 metric tons of DUF6 stored at the sites of the two gaseous diffusion plants located at Paducah, Kentucky and Portsmouth, Ohio, and at the East Tennessee Technology Park in Oak Ridge, Tennessee. The DUF6 management program resides in NE's Office of Depleted Uranium Hexafluoride Management. The current DUF6 program has largely focused on the ongoing maintenance of the cylinders containing DUF6. However, the long-term management and eventual disposition of DUF6 is the subject of a Programmatic Environmental Impact Statement (PEIS) and Public Law 105-204. The first step for future use or disposition is to convert the material, which requires construction and long-term operation of one or more conversion plants. To help inform the DUF6 program's planning activities, it was necessary to perform design and cost studies of likely DUF6 conversion plants at the preconceptual level, beyond the PEIS considerations but not as detailed as required for conceptual designs of actual plants. This report contains the final results from such a preconceptual design study project. In this fast track, three month effort, Lawrence Livermore National Laboratory and Bechtel National Incorporated developed and evaluated seven different preconceptual design cases for a single plant. The preconceptual design, schedules, costs, and issues associated with specific DUF6 conversion approaches, operating periods, and ownership options were evaluated based on criteria established by DOE. The single-plant conversion options studied were similar to the dry-conversion process alternatives from the PEIS. For each of the seven cases considered, this report contains information on
Plant design and beam utilization

International Nuclear Information System (INIS)

Svendsen, E.B.

1983-01-01

Plant design and beam utilization are two things closely tied together: without a proper plant design, one can never get good beam utilization. When a company decides to build an irradiation facility, there are some major decisions to be made right in the beginning. These decisions can be most important for the long-term success or failure of the irradiation facility, because the company normally will have to live with these decisions during the whole life-time of the irradiation equipment. To start with the decision has to be made whether to select a cobalt-60 irradiation plant or an accelerator irradiation plant. This decision can only be reached after a careful study of the products and the 'weight' and the material of the products the company wants to irradiate. As an old accelerator-man, I tend to personally favor accelerators, although I am very impressed by the newer cobalt-60 pallet irradiation plants from A.E.C.L. I believe that they have a great future in the emerging field of food irradiation. As I have primarily been involved with accelerators during the last 14 years, this paper is only dealing with different design approaches and utilizations of accelerator-plants. (author)
Development of design methodology for communication network in nuclear power plants

International Nuclear Information System (INIS)

Kim, Dong Hoon; Seong, Seung Hwan; Jang Gwi Sook; Koo, In Soo; Lee Soon Sung.

1996-06-01

This report describe the design methodology of communication network (CN) in nuclear power plants (NPPs). The construction procedure for the NPP CN consists of 4 phases, in study and review phase, design concepts and goals are established through technical review, collection of background information and feasibility study. In design phase, all of design activities such as extraction of requirements, communication modelling, overall and detail architecture design are performed. Implementation and test phase includes the manufacturing, installation and testing of hardware and software. In operation phase, CN construction is finalized through the evaluation and correction during operation. The requirements of CN consist of general requirements such as function, structure, reliability, standardization and detail requirements related with protocol, media, error, performance and etc. CN design also should follow the safety-related requirements such as isolation, redundancy, reliability and verify these requirements. For the selection of each technical element form commercial CN, the evaluation and selection elements are extracted from reliability, performance, operating factors and the required-level which classified into essential, primary, preference, recommendation should be assigned to each element. This report will be used as a technical reference for the CN implementation in NPP. (author). 3 tabs., 5 figs., 25 refs
A new approach to the design of the large turbofan power plant

Energy Technology Data Exchange (ETDEWEB)

Wilde, G L [Economobile Projects Ltd., Belper (United Kingdom)

1995-06-01

The lower direct operating costs of the Big Twin subsonic transports encourage the building of ever larger turbofan engines installed on the wings. The steadily improving reliability of the turbofan and the good safety statistics of twin-engined aircraft over many years encourages this trend. Fuel economy is still the dominant factor in determining the design layout of turbofan engines. It requires the combination of the highest possible thermal efficiency of the gas generator core of the engine with optimum propulsion efficiency of the power plant as a whole in cruise flight, allowing for engine nacelle drag and nacelle to wing interference drag. The paper presents two possible turbofan design layouts intended to overcome the limitation of current turbofan power plant designs. The aim is to design a power plant with the highest thrust per unit frontal area combined with the highest air miles per gallon in cruise flight. (author)
Toward the ultimate goal of tritium self-sufficiency: Technical issues and requirements imposed on ARIES advanced power plants

International Nuclear Information System (INIS)

El-Guebaly, Laila A.; Malang, Siegfried

2009-01-01

Due to the lack of external tritium sources, all fusion power plants must demonstrate a closed tritium fuel cycle. The tritium breeding ratio (TBR) must exceed unity by a certain margin. The key question is: how large is this margin and how high should the calculated TBR be? The TBR requirement is design and breeder-dependent and evolves with time. At present, the ARIES requirement is 1.1 for the calculated overall TBR of LiPb systems. The Net TBR during plant operation could be around 1.01. The difference accounts for deficiencies in the design elements (nuclear data evaluation, neutronics code validation, and 3D modeling tools). Such a low Net TBR of 1.01 is potentially achievable in advanced designs employing advanced physics and technology. A dedicated R and D effort will reduce the difference between the calculated TBR and Net TBR. A generic breeding issue encountered in all fusion designs is whether any fusion design will over-breed or under-breed during plant operation. To achieve the required Net TBR with sufficient precision, an online control of tritium breeding is highly recommended for all fusion designs. This can easily be achieved for liquid breeders through online adjustment of Li enrichment.
Safety related requirements on future nuclear power plants

International Nuclear Information System (INIS)

Niehaus, F.

1991-01-01

Nuclear power has the potential to significantly contribute to the future energy supply. However, this requires continuous improvements in nuclear safety. Technological advancements and implementation of safety culture will achieve a safety level for future reactors of the present generation of a probability of core-melt of less than 10 -5 per year, and less than 10 -6 per year for large releases of radioactive materials. There are older reactors which do not comply with present safety thinking. The paper reviews findings of a recent design review of WWER 440/230 plants. Advanced evolutionary designs might be capable of reducing the probability of significant off-site releases to less than 10 -7 per year. For such reactors there are inherent limitations to increase safety further due to the human element, complexity of design and capability of the containment function. Therefore, revolutionary designs are being explored with the aim of eliminating the potential for off-site releases. In this context it seems to be advisable to explore concepts where the ultimate safety barrier is the fuel itself. (orig.) [de
Design of reactor containment systems for nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2008-01-01

This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. It is a revision of the Safety Guide on Design of the Reactor Containment Systems in Nuclear Power Plants (Safety Series No. 50-Sg-D1) issued in 1985 and supplements the Safety Requirements publication on Safety of Nuclear Power Plants: Design. The present Safety Guide was prepared on the basis of a systematic review of the relevant publications, including the Safety of Nuclear Power Plants: Design, the Safety fundamentals publication on The Safety of Nuclear Installations, Safety Guides, INSAG Reports, a Technical Report and other publications covering the safety of nuclear power plants. 1.2. The confinement of radioactive material in a nuclear plant, including the control of discharges and the minimization of releases, is a fundamental safety function to be ensured in normal operational modes, for anticipated operational occurrences, in design basis accidents and, to the extent practicable, in selected beyond design basis accidents. In accordance with the concept of defence in depth, this fundamental safety function is achieved by means of several barriers and levels of defence. In most designs, the third and fourth levels of defence are achieved mainly by means of a strong structure enveloping the nuclear reactor. This structure is called the 'containment structure' or simply the 'containment'. This definition also applies to double wall containments. 1.3. The containment structure also protects the reactor against external events and provides radiation shielding in operational states and accident conditions. The containment structure and its associated systems with the functions of isolation, energy management, and control of radionuclides and combustible gases are referred to as the containment systems
Design of reactor containment systems for nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2004-01-01

This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. It is a revision of the Safety Guide on Design of the Reactor Containment Systems in Nuclear Power Plants (Safety Series No. 50-Sg-D1) issued in 1985 and supplements the Safety Requirements publication on Safety of Nuclear Power Plants: Design. The present Safety Guide was prepared on the basis of a systematic review of the relevant publications, including the Safety of Nuclear Power Plants: Design, the Safety fundamentals publication on The Safety of Nuclear Installations, Safety Guides, INSAG Reports, a Technical Report and other publications covering the safety of nuclear power plants. 1.2. The confinement of radioactive material in a nuclear plant, including the control of discharges and the minimization of releases, is a fundamental safety function to be ensured in normal operational modes, for anticipated operational occurrences, in design basis accidents and, to the extent practicable, in selected beyond design basis accidents. In accordance with the concept of defence in depth, this fundamental safety function is achieved by means of several barriers and levels of defence. In most designs, the third and fourth levels of defence are achieved mainly by means of a strong structure enveloping the nuclear reactor. This structure is called the 'containment structure' or simply the 'containment'. This definition also applies to double wall containments. 1.3. The containment structure also protects the reactor against external events and provides radiation shielding in operational states and accident conditions. The containment structure and its associated systems with the functions of isolation, energy management, and control of radionuclides and combustible gases are referred to as the containment systems

NSS design and plant construction interfaces

International Nuclear Information System (INIS)

Stewart, J.J.; Cobb, W.A.

1976-01-01

Interface management between NSS design, balance-of-plant design, and plant construction may have a significant effect on schedule stretchout and total plant costs. The paper discusses the importance of the NSS supplier's interface management role, the favorable and unfavorable influencing factors, and examples of interface areas in which experience has demonstrated that problems may arise. Where appropriate, actions are defined to avoid the problems or mitigate the consequences
Design information verification for spent fuel conditioning plants and for geological repositories

International Nuclear Information System (INIS)

Myatt, J.; Ward, M.D.

1995-01-01

The disposal of spent fuel is a major option for the back-end of the nuclear fuel cycle. It will require the construction, operation and eventual closure of conditioning plants and geological repositories. Consequently, a safeguards approach including Design Information Verification (DIV) must be developed for these facilities. DIV Is the examination of a completed facility to verify that it has been built to the design declared by the operator. Although DIV takes place chiefly before a plant begins routine operation, there is a continuing interest in ensuring that the plant remains as declared. That is, that the continuity of knowledge of design information is maintained during the operational phase of the plant and also post closure if necessary. A major problem with DIV of a repository is that there will be continuous structural changes during its operational life requiring advanced or special techniques for reverification. Some of these are briefly reviewed. Furthermore, since a disposal facility is expected to be operational for several decades, new mining technology may also have an impact on the DIV methods employed. Another factor in the safeguards supervision of a repository is that when the fuel has been backfilled and/or scaled in place a reassay will be a very costly exercise. The role of DIV in such novel circumstances must, therefore, be fully considered
Ergonomic requirements on computer-based information- and handling engineering in nuclear power plants

International Nuclear Information System (INIS)

Fassmann, W.

2002-01-01

This project provides regulatory authorities with a set of criteria for evaluating hybrid man-machine interfaces in nuclear power plant control rooms from a human factors point of view. Such standards are necessary for two reasons: (1) More and more computerised information and control systems have been and will be introduced in nuclear power plant control rooms. One possible result of this trend will be the creation of hybrid man machine interfaces which will provide both conventional and computer-based display and control devices. (2) Available rules and regulations do not contain detailed requirements on how to integrate both types of interface in such a way that plant operation by means of hybrid interfaces will be performed at least as reliably and safely as by means of conventional ones. To fill this gap, criteria and methods were developed which support practical checks of requirements to be applied to hybrid control rooms. This approach is based on state of the art methods and criteria in ergonomics. It makes it possible to analyse and to describe personnel's actions in a consistent and structured way in order to provide the information which is necessary for evaluating human reliability of task performance. Reliability can be evaluated with respect to - accuracy of required information on displays, - networking of tasks, - possibilities of interrupting and cancelling measures which have already been initiated, - possibility to carry out required manuel actions, - level of mental work-strain, - workload level, - probability of erroneous actions. This method is part of a catalogue of recommendations for evaluating hybrid nuclear power plant control rooms. The catalogue also contains recommendations for the design of computerised parts of the man-machine-interface. Application of these design recommendations will help create favourable conditions for an acceptable level of work-strain and for reliable task performance. (orig.) [de
Design requirements for new nuclear reactor facilities in Canada (focus on important improvements from RD-337 version 1 of 2008)

International Nuclear Information System (INIS)

Shim, S.; Harwood, C.; Ohn, M-Y; Liu, Y.C.; Young, T.

2014-01-01

The Canadian Nuclear Safety Commission (CNSC) has established the regulatory framework that includes the documentation of the requirements and guidance for each of CNSC's 14 Safety and Control Areas, one important area being the design of nuclear power plants (NPPs). For the design area, the CNSC published RD-337 version 1 Design of New Nuclear Power Plants in 2008. As such regulatory documents are reviewed on a regular basis, this document was recently updated as RD-337 version 2, and its guidance document GD-337, Guidance for the Design of New Nuclear Power Plants was developed to provide guidance on how to meet the requirements. REGDOC-2.5.2 Design of Reactor Facilities: Nuclear Power Plants that combines RD-337 version 2 and GD-337 version 1 was presented to the Commission on March 27, 2014 after two rounds of consultation with stakeholders, and was subsequently published in May 28, 2014. Although REGDOC-2.5.2 maintains the structure and the contents nearly the same as RD-337 version 1, it introduces several important improvements to: Include GD-337 guidance for further clarity to applicants, licensees and vendors on how to meet the requirements. This guidance provides the review criteria considered in CNSC staff's review in a transparent way; Ensure alignment with international standards including recent IAEA SSR 2/1, Safety of Nuclear Power Plants: Design; Implement CNSC Fukushima Task Force Report findings that pertain to the design of reactor facilities for severe accidents; and, Make necessary improvements such as addition of requirements for cyber security. This paper describes the overall regulatory framework related to CNSC's design requirements and guidance for NPPs, and focus on the important improvements included in REGDOC-2.5.2 and their reasoning. (author)
Design requirements for new nuclear reactor facilities in Canada (focus on important improvements from RD-337 version 1 of 2008)

Energy Technology Data Exchange (ETDEWEB)

Shim, S.; Harwood, C.; Ohn, M-Y; Liu, Y.C.; Young, T. [Canadian Nuclear Safety Commission, Ottawa, ON (Canada)

2014-07-01

The Canadian Nuclear Safety Commission (CNSC) has established the regulatory framework that includes the documentation of the requirements and guidance for each of CNSC's 14 Safety and Control Areas, one important area being the design of nuclear power plants (NPPs). For the design area, the CNSC published RD-337 version 1 Design of New Nuclear Power Plants in 2008. As such regulatory documents are reviewed on a regular basis, this document was recently updated as RD-337 version 2, and its guidance document GD-337, Guidance for the Design of New Nuclear Power Plants was developed to provide guidance on how to meet the requirements. REGDOC-2.5.2 Design of Reactor Facilities: Nuclear Power Plants that combines RD-337 version 2 and GD-337 version 1 was presented to the Commission on March 27, 2014 after two rounds of consultation with stakeholders, and was subsequently published in May 28, 2014. Although REGDOC-2.5.2 maintains the structure and the contents nearly the same as RD-337 version 1, it introduces several important improvements to: Include GD-337 guidance for further clarity to applicants, licensees and vendors on how to meet the requirements. This guidance provides the review criteria considered in CNSC staff's review in a transparent way; Ensure alignment with international standards including recent IAEA SSR 2/1, Safety of Nuclear Power Plants: Design; Implement CNSC Fukushima Task Force Report findings that pertain to the design of reactor facilities for severe accidents; and, Make necessary improvements such as addition of requirements for cyber security. This paper describes the overall regulatory framework related to CNSC's design requirements and guidance for NPPs, and focus on the important improvements included in REGDOC-2.5.2 and their reasoning. (author)
Information management systems improve advanced plant design

International Nuclear Information System (INIS)

Turk, R.S.; Serafin, S.A.; Leckley, J.B.

1994-01-01

Computer-aided engineering tools are proving invaluable in both the design and operation of nuclear power plants. ABB Combustion Engineering's Advanced Light Water Reactor (ALWR) features a computerized Information Management System (IMS) as an integral part of the design. The System 80+IMS represents the most powerful information management tool for Nuclear Power Plants commercially available today. Developed by Duke Power Company specifically for use by nuclear power plant owner operators, the IMS consists of appropriate hardware and software to manage and control information flow for all plant related work or tasks in a systematic, consistent, coordinated and informative manner. A significant feature of this IMS is that it is primarily based on plant data. The principal design tool, PASCE (Plant Application and Systems from Combustion Engineering), is comprised of intelligent databases that describe the design and from which accurate plant drawings are created. Additionally the IMS includes, at its hub, a relational database management system and an associated document management system. The data-based approach and applications associated with the IMS were developed, and have proven highly effective, for plant modifications, configuration management, and operations and maintenance applications at Duke Power Company's operating nuclear plants. This paper presents its major features and benefits. 4 refs
Solar Pilot Plant, Phase I. Preliminary design report. Volume II. System description and system analysis. CDRL item 2

Energy Technology Data Exchange (ETDEWEB)

None

1977-05-01

Honeywell conducted a parametric analysis of the 10-MW(e) solar pilot plant requirements and expected performance and established an optimum system design. The main analytical simulation tools were the optical (ray trace) and the dynamic simulation models. These are described in detail in Books 2 and 3 of this volume under separate cover. In making design decisions, available performance and cost data were used to provide a design reflecting the overall requirements and economics of a commercial-scale plant. This volume contains a description of this analysis/design process and resultant system/subsystem design and performance.
Design requirements for a metal-smelting facility

International Nuclear Information System (INIS)

Williams, L.C.; Mack, J.E.

1982-01-01

Functional requirements for the smelting of metal scrap contaminated with low-enriched uranium in a Metal Smelting Faclity (MSF) have been determined. The process will be designed to smelt ferrous metal scrap that has accumulated at the Oak Ridge Gaseous Diffusion Plant (ORGDP) into one-ton ingots at a rate of 40 ingots per day (10,000 tons/year). Total scrap inventories at the ORGDP are currently estimated at 28,000 tons. The diffusion plant scrap is primarily contaminated with 100 to 200 ppm U at an enrichment of 0.5 to 1.5% 235 U. The scrap is considered special nuclear material (SNM) and cannot be handled by commercial smelters without specific licensing. Slagging will be performed to remove contaminants from the metal and concentrate them in the slag. Process systems will include scrap handling, size reduction, preheating and charging, melting and slagging, ingot casting and storage, and fume exhaust. The MSF has been proposed for FY 1984 line item funding
Design of a requirements system for decommissioning of a nuclear power plant based on systems engineering

Energy Technology Data Exchange (ETDEWEB)

Park, Hee Seong; Park, Seung Kook; Jin, Hyung Gon; Song, Chan Ho; Choi, Jong won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

The nuclear industry has required an advanced system that can manage decommissioning information ever since the Korean government decide to decommission the Gori No.1 nuclear power plant. The D and D division at KAERI has been developing a system that can secure the reliability and sustainability of the decommissioning project based on the engineering system of the KRR-2 (Korean Research Reactor-2). To establish a decommissioning information system, a WBS that needs to be managed for the decommissioning of an NPP has been extracted, and requirements management research composed of system engineering technology has progressed. This paper propose a new type of system based on systems engineering technology. Even though a decommissioning engineering system was developed through the KRR-2, we are now developing an advanced decommissioning information system because it is not easy to apply this system to a commercial nuclear power plant. An NPP decommissioning is a project requiring a high degree of safety and economic feasibility. Therefore, we have to use a systematic project management at the initial phase of the decommissioning. An advanced system can manage the decommissioning information from preparation to remediation by applying a previous system to the systems engineering technology that has been widely used in large-scale government projects. The first phase of the system has progressed the requirements needed for a decommissioning project for a full life cycle. The defined requirements will be used in various types of documents during the decommissioning preparation phase.
Computer aided design and management of cables in a nuclear power plant

International Nuclear Information System (INIS)

Pralus, B.

1978-01-01

Specific problems encountered at both design offices and on-site on management of cables for large nuclear power plants are analyzed. These problems, for a large pert planning, come from safety regulations and quality assurance requirements as well. Emphasis is put on the effect of the QA exigences applicable to the organization and execution of cabling management for nuclear power plants. (J.E. de C.)
Next Generation Nuclear Plant System Requirements Manual

International Nuclear Information System (INIS)

Not Listed

2008-01-01

System Requirements Manual for the NGNP Project. The Energy Policy Act of 2005 (H.R. 6; EPAct), which was signed into law by President George W. Bush in August 2005, required the Secretary of the U.S. Department of Energy (DOE) to establish a project to be known as the Next Generation Nuclear Plant (NGNP) Project. According to the EPAct, the NGNP Project shall consist of the research, development, design, construction, and operation of a prototype plant (to be referred to herein as the NGNP) that (1) includes a nuclear reactor based on the research and development (R and D) activities supported by the Generation IV Nuclear Energy Systems initiative, and (2) shall be used to generate electricity, to produce hydrogen, or to both generate electricity and produce hydrogen. The NGNP Project supports both the national need to develop safe, clean, economical nuclear energy and the Nuclear Hydrogen Initiative (NHI), which has the goal of establishing greenhouse-gas-free technologies for the production of hydrogen. The DOE has selected the helium-cooled High Temperature Gas-Cooled Reactor (HTGR) as the reactor concept to be used for the NGNP because it is the only near-term Generation IV concept that has the capability to provide process heat at high-enough temperatures for highly efficient production of hydrogen. The EPAct also names the Idaho National Laboratory (INL), the DOE's lead national laboratory for nuclear energy research, as the site for the prototype NGNP
Modelled basic parameters for semi-industrial irradiation plant design

International Nuclear Information System (INIS)

Mangussi, J.

2009-01-01

The basic parameters of an irradiation plant design are the total activity, the product uniformity ratio and the efficiency process. The target density, the minimum dose required and the throughput depends on the use to which the irradiator will be put at. In this work, a model for calculating the specific dose rate at several depths in an infinite homogeneous medium produced by a slab source irradiator is presented. The product minimum dose rate for a set of target thickness is obtained. The design method steps are detailed and an illustrative example is presented. (author)
Correct safety requirements during the life cycle of heating plants; Korrekta saekerhetskrav under vaermeanlaeggningars livscykel

Energy Technology Data Exchange (ETDEWEB)

Tegehall, Jan; Hedberg, Johan [Swedish National Testing and Research Inst., Boraas (Sweden)

2006-10-15

The safety of old steam boilers or hot water generators is in principle based on electromechanical components which are generally easy to understand. The use of safety-PLC is a new and flexible way to design a safe system. A programmable system offers more degrees of freedom and consequently new problems may arise. As a result, new standards which use the Safety Integrity Level (SIL) concept for the level of safety have been elaborated. The goal is to define a way of working to handle requirements on safety in control systems of heat and power plants. SIL-requirements are relatively new within the domain and there is a need for guidance to be able to follow the requirements. The target of this report is the people who work with safety questions during new construction, reconstruction, or modification of furnace plants. In the work, the Pressure Equipment Directive, 97/23/EC, as well as standards which use the SIL concept have been studied. Additionally, standards for water-tube boilers have been studied. The focus has been on the safety systems (safety functions) which are used in water-tube boilers for heat and power plants; other systems, which are parts of these boilers, have not been considered. Guidance has been given for the aforementioned standards as well as safety requirements specification and risk analysis. An old hot water generator and a relatively new steam boiler have been used as case studies. The design principles and safety functions of the furnaces have been described. During the risk analysis important hazards were identified. A method for performing a risk analysis has been described and the appropriate content of a safety requirements specification has been defined. If a heat or power plant is constructed, modified, or reconstructed, a safety life cycle shall be followed. The purpose of the safety life cycle is to plan, describe, document, perform, check, test, and validate that everything is correctly done. The components of the safety
Seismic design of a uranium conversion plant building

International Nuclear Information System (INIS)

Peixoto, O.J.M.; Botelho, C.L.A.; Braganca, A. Jr.; C. Santos, S.H. de.

1992-01-01

The design of facilities with small radioactive inventory has been traditionally performed following the usual criteria for industrial buildings. In the last few years, more stringent criteria have been adopted in new nuclear facilities in order to achieve higher standards for environmental protection. In uranium conversion plants, the UF 6 (uranium hexafluoride) production step is the part of the process with the highest potential for radioactivity release to the environment because of the operations performed in the UF 6 desublimers and cylinder filling areas as well as UF 6 distillation facilities, when they are also required in the process. This paper presents the design guidelines and some details of the seismic resistance design of a UF 6 production building to be constructed in Brazil
Plant design and layout of the different buildings with respect to safety, operational and maintenance requirements

International Nuclear Information System (INIS)

Liebich, H.

1981-01-01

The descriptions and pictures in this lecture show that the arrangement of the buildings and the location of components and systems are based on proven ideas with the aim to fulfil safety, operational and maintenance requirements also from the point of view of plant layout. (orig.)
Design of Radioactive Waste Management Systems at Nuclear Power Plants

International Nuclear Information System (INIS)

1986-01-01

This Safety Guide is addressed to the administrative and technical authorities and specialists dealing with the design, construction and operation of nuclear power plants, and in particular waste management facilities at nuclear power plants. This Guide has been prepared as part of the IAEA Waste Handling, Treatment and Storage programme. It is a follow-up document to the Code of Practice on Management of Radioactive Wastes from Nuclear Power Plants published in 1985 in the IAEA Safety Standards, Safety Series No. 69, in which basic principles for management of radioactive wastes at nuclear power plants are set out. The IAEA has established wide ranging programmes to provide Member States with guidance on different aspects of safety and technology related to thermal neutron power reactors and associated nuclear fuel cycle operations, including those for management of radioactive wastes. There are many IAEA publications related to various technical and safety aspects of different nuclear energy applications. All these publications are issued by the Agency for the use of Member States in connection with their own nuclear technological safety requirements. They are based on national experience contributed by experts from different countries and relate to common features in approaches to the problems discussed. However, the final decision and legal responsibility in any regulatory procedure always rest with the Member State. This particular Guide aims to provide general and detailed principles for the design of waste management facilities at nuclear power plants. It emphasizes what and how specific safety requirements for the management of radioactive wastes from nuclear power plants can be met in the design and construction stage. The safety requirements for operation of such facilities will be considered in the Agency's next Safety Series publication, Safety Guide 50-SG-011, Operational Management for Radioactive Effluents and Wastes Arising in Nuclear Power Plants
Design methodology for the physical protection upgrade rule requirements for fixed sites. Technical report

International Nuclear Information System (INIS)

Evans, L.J. Jr.; Allen, T.

1980-06-01

This Design Methodology document aids the licensee in understanding how the fixed site requirements of the Physical Protection Upgrade Rule affect the design of physical protection systems for fuel processing plants, fuel manufacturing plants, or other fixed site special nuclear material operations involving possession or use of formula quantities of strategic special nuclear material. The document consists of three major elements: Logic Trees, Safeguards Jobs and Component Matrices, and Effectiveness Test Questionnaires. The work is based upon a previous study conducted by Sandia Laboratories for the Nuclear Regulatory Commission
The Design and Manufacturing of Essential oil Distillation Plant for ...

African Journals Online (AJOL)

Choice-Academy

industry in the country do not have the capacity to manufacture the complete distillation plant system with the required precision for standard quality of oil at affordable cost. Thus, the design and the experiment in the use of a prototype small size distillation unit showed that the technology is appropriate for essential oil ...
Safety requirements for a nuclear power plant electric power system

Energy Technology Data Exchange (ETDEWEB)

Fouad, L F; Shinaishin, M A

1988-06-15

This work aims at identifying the safety requirements for the electric power system in a typical nuclear power plant, in view of the UNSRC and the IAEA. Description of a typical system is provided, followed by a presentation of the scope of the information required for safety evaluation of the system design and performance. The acceptance and design criteria that must be met as being specified by both regulatory systems, are compared. Means of implementation of such criteria as being described in the USNRC regulatory guides and branch technical positions on one hand and in the IAEA safety guides on the other hand are investigated. It is concluded that the IAEA regulations address the problems that may be faced with in countries having varying grid sizes ranging from large stable to small potentially unstable ones; and that they put emphasis on the onsite standby power supply. Also, in this respect the Americans identify the grid as the preferred power supply to the plant auxiliaries, while the IAEA leaves the possibility that the preferred power supply could be either the grid or the unit main generator depending on the reliability of each. Therefore, it is found that it is particularly necessary in this area of electric power supplies to deal with the IAEA and the American sets of regulations as if each complements and not supplements the other. (author)
Science requirements and the design of cabled ocean observatories

Directory of Open Access Journals (Sweden)

H. Mikada

2006-06-01

Full Text Available The ocean sciences are beginning a new phase in which scientists will enter the ocean environment and adaptively observe the Earth-Ocean system through remote control of sensors and sensor platforms. This new ocean science paradigm will be implemented using innovative facilities called ocean observatories which provide unprecedented levels of power and communication to access and manipulate real-time sensor networks deployed within many different environments in the ocean basins. Most of the principal design drivers for ocean observatories differ from those for commercial submarine telecommunications systems. First, ocean observatories require data to be input and output at one or more seafloor nodes rather than at a few land terminuses. Second, ocean observatories must distribute a lot of power to the seafloor at variable and fluctuating rates. Third, the seafloor infrastructure for an ocean observatory inherently requires that the wet plant be expandable and reconfigurable. Finally, because the wet communications and power infrastructure is comparatively complex, ocean observatory infrastructure must be designed for low life cycle cost rather than zero maintenance. The origin of these differences may be understood by taking a systems engineering approach to ocean observatory design through examining the requirements derived from science and then going through the process of iterative refinement to yield conceptual and physical designs. This is illustrated using the NEPTUNE regional cabled observatory power and data communications sub-systems.

The 18 basic requirement of quality assurance for American design NPP

International Nuclear Information System (INIS)

Baliza, Ana Rosa

2013-01-01

On April 17th, 1969, the Atomic Energy Commission (AEC) published in the U.S. Federal Register (FR), Volume 34, Number 73, a proposed amendment to 10CFR50 to insert Appendix B - 'Quality assurance criteria for Nuclear Power Plant'. This Appendix was officially approved on June 27th, 1970 and published in the FR, volume 35, number 125. Appendix B is the Quality Assurance document for U.S. nuclear facilities. This document establishes eighteen basic requirements (BR) to design, construction, manufacture and operation of structures, systems and components (SSC) related to safety. The 18 BR describe 'what' shall be done, but not 'how' to do. In order to standardize the actions of nuclear facilities during 10 CFR 50 App B implementation, the industry has developed some documents, the main ones are: ASME NQA-1 (Quality Assurance Requirements for Nuclear Facility Applications) and the series ANSI N 45.2 (Quality Assurance Program Requirements for Nuclear Facilities). Both documents are approved by the NRC (Nuclear Regulatory Commission). The NRC is the licensing body of U.S. nuclear facilities. In Brazil, the licensing body is CNEN (Comissao Nacional de Energia Nuclear). This paper describes the 18 BR for American Designed Nuclear Power Plant (NPP), applicable to Angra-1 NPP. (author)
Improved design for vibration-proof platinum RTD in special position of nuclear power plant

International Nuclear Information System (INIS)

Liu Zhuo; Ma Jinna; Wu Bin

2014-01-01

In accordance with the actual situation for the vibration of violence in a special position of nuclear power plant, an improved design for platinum RTD was proposed. The structure design is verified to meet the measure requirement in the special position. (authors)
Tornado-resistance design for the nuclear safety structure of Qinshan Nuclear Power Plant

International Nuclear Information System (INIS)

Xia Zufeng.

1987-01-01

The primary design consideration of anti-tornado of the nuclear safety structure of Qinshan Nuclear Power Plant is briefly presented. It mainly includes estimating the probability of tornado arising in the site, ascertaining the design requirments of the anti-tornado structures and deciding the tornado load acted on the structures
Licensing evaluation of CANDU-PHW nuclear power plants relative to U.S. regulatory requirements

International Nuclear Information System (INIS)

Erp, J.B. van

1978-01-01

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

International Nuclear Information System (INIS)

Chang, N.E.

1981-01-01

An overview of reliability analysis techniques is presented as applied to power plant design. The key terms, power plant performance, reliability, availability and maintainability are defined. Reliability modeling, methods of analysis and component reliability data are briefly reviewed. Application of reliability analysis techniques from a design engineering approach to improving power plant productivity is discussed. (author)
Influence of student-designed experiments with fast plants on their understanding of plants and of scientific inquiry

Science.gov (United States)

Akey, Ann Kosek

2000-10-01

This dissertation investigates the influence of student designed experiments with Fast Plants in an undergraduate agroecology course on the students' conceptual understanding of plant life cycles and on their procedural understanding of scientific experimentation. It also considers students' perspectives on the value of these experiences. Data sources included semi-structured interviews with students and the instructor, a written task, course evaluations, and observations of class meetings. Students came into the course having strong practical experience with plants from their agricultural backgrounds. Students did not always connect aspects of plant biology that they studied in class, particularly respiration and photosynthesis, to plant growth requirements. The instructor was able to bridge the gap between some practical knowledge and textbook knowledge with experiences other than the Fast Plant project. Most students held an incomplete picture of plant reproduction that was complicated by differences between agricultural and scientific vocabulary. There is need for teaching approaches that help students tie together their knowledge of plants into a cohesive framework. Experiences that help students draw on their background knowledge related to plants, and which give students the opportunity to examine and discuss their ideas, may help students make more meaningful connections. The Fast Plant project, a positive experience for most students, was seen by these undergraduate students as being more helpful in learning about scientific experimentation than about plants. The process of designing and carrying out their own experiments gave students insight into experimentation, provoked their curiosity, and resulted in a sense of ownership and accomplishment.
Designing new nuclear chemical processing plants for safeguards accountability

International Nuclear Information System (INIS)

Sprouse, K.M.

1987-01-01

New nuclear chemical processing plants will be required to develop material accountability control limits from measurement error propagation analysis rather than historical inventory difference data as performed in the past. In order for measurement error propagation methods to be viable alternatives, process designers must ensure that two nondimensional accountability parameters are maintained below 0.1. These parameters are ratios between the material holdup increase and the variance in inventory difference measurement uncertainty. Measurement uncertainty data for use in error propagation analysis is generally available in the open literature or readily derived from instrument calibration data. However, nuclear material holdup data has not been adequately developed for use in the material accountability design process. Long duration development testing on isolated unit operations is required to generate this necessary information
Preliminary design of the Carrisa Plains solar central receiver power plant. Volume III, Book 1. Design description

Energy Technology Data Exchange (ETDEWEB)

1983-12-31

The design of the 30 MWe central receiver solar power plant to be located at Carrisa Plains, San Luis Obispo County, California, is summarized. The plant uses a vertical flat-panel (billboard solar receiver located at the top of a tower to collect solar energy redirected by approximately 1900 heliostats located to the north of the tower. The solar energy is used to heat liquid sodium pumped from ground level from 610 to 1050/sup 0/F. The power conversion system is a non-reheat system, cost-effective at this size level, and designed for high-efficiency performance in an application requiring daily startup. Successful completion of this project will lead to power generation starting in 1986. This report discusses in detail the design of the collector system, heat transport system, thermal storage subsystem, heat transport loop, steam generation subsystem, electrical, instrumentation, and control systems, power conversion system, master control system, and balance of plant. The performance, facility cost estimate and economic analysis, and development plan are also discussed.
Guidebook on design, construction and operation of pilot plants for uranium ore processing

International Nuclear Information System (INIS)

1990-01-01

The design, construction and operation of a pilot plant are often important stages in the development of a project for the production of uranium concentrates. Since building and operating a pilot plant is very costly and may not always be required, it is important that such a plant be built only after several prerequisites have been met. The main purpose of this guidebook is to discuss the objectives of a pilot plant and its proper role in the overall project. Given the wide range of conditions under which a pilot plant may be designed and operated, it is not possible to provide specific details. Instead, this book discusses the rationale for a pilot plant and provides guidelines with suggested solutions for a variety of problems that may be encountered. This guidebook is part of a series of Technical Reports on uranium ore processing being prepared by the IAEA's Division of Nuclear Fuel Cycle and Waste Management. 42 refs, 7 figs, 3 tabs
VGB-requirements regarding technical data for power plants

International Nuclear Information System (INIS)

Richnow, Joerg

2009-01-01

Much of the technical plant data resulting from the planning, construction and start-up of power plants is needed for subsequent management and maintenance. Because of this, VGB has taken the initiative and has defined standard minimum requirements from power plant operators for technical plant data. They relate to the details and structure of this data, the definition of material classes and characteristics for the main power plant components and IT implementation for delivery of the technical plant data. (orig.)
Basic national requirements for safe design, construction and operation

International Nuclear Information System (INIS)

Franzen, L.F.

1980-01-01

Nuclear power plants have to be save. Vendors and utilities operating such plants, are convinced that their plants meet this requirement. Who, however, is establishing the safety requirements to be met by those manufacturing and operating nuclear power plants. What are the mechanisms to control whether the features provided assure the required safety level. Who controls whether the required and planned safety features are really provided. Who is eventually responsible for assuring safety after commissioning of a nuclear power plant. These fundamental questions being raised in many discussions on safety and environmental protection are dealt with in the following sections: (1) Fundamental safety requirements on nuclear power plants, in which such items as risk, legal bases and licensing procedure are discussed, (2) Surveillance during construction, in which safety analysis report, siting, safety evaluation, document examination, quality assurance, and commissioning testing are dealt with, (3) Operating tests and conditions in which recurrent inspections, environmental protection during operation, investigation of abnormal occurences and backfitting requirements as reviewed, and (4) Safety philosophy and safety policy to conclude this presentation. The German approach to nuclear safety serves as an example for an effective way of assuring safe nuclear power. (orig.)
Tool-based requirement traceability between requirement and design artifacts

CERN Document Server

Turban, Bernhard

2013-01-01

Processes for developing safety-critical systems impose special demands on ensuring requirements traceability. Achieving valuable traceability information, however, is especially difficult concerning the transition from requirements to design. Bernhard Turban analyzes systems and software engineering theories cross-cutting the issue (embedded systems development, systems engineering, software engineering, requirements engineering and management, design theory and processes for safety-critical systems). As a solution, the author proposes a new tool approach to support designers in their thinkin
Nuclear power plant C and I design verification by simulation

International Nuclear Information System (INIS)

Storm, Joachim; Yu, Kim; Lee, D.Y

2003-01-01

An important part of the Advanced Boiling Water Reactor (ABWR) in the Taiwan NPP Lungmen Units no.1 and no.2 is the Full Scope Simulator (FSS). The simulator was to be built according to design data and therefore, apart from the training aspect, a major part of the development is to apply a simulation based test bed for the verification, validation and improvement of plant design in the control and instrumentation (C and I) areas of unit control room equipment, operator Man Machine Interface (MMI), process computer functions and plant procedures. Furthermore the Full Scope Simulator will be used after that to allow proper training of the plant operators two years before Unit no.1 fuel load. The article describes scope, methods and results of the advanced verification and validation process and highlights the advantages of test bed simulation for real power plant design and implementation. Subsequent application of advanced simulation software tools like instrumentation and control translators, graphical model builders, process models, graphical on-line test tools and screen based or projected soft panels, allowed a team to fulfil the task of C and I verification in time before the implementation of the Distributed Control and Information System (DCIS) started. An additional area of activity was the Human Factors Engineering (HFE) for the operator MMI. Due to the fact that the ABWR design incorporates a display-based operation with most of the plant components, a dedicated verification and validation process is required by NUREG-0711. In order to support this activity an engineering test system had been installed for all the necessary HFE investigations. All detected improvements had been properly documented and used to update the plant design documentation by a defined process. The Full Scope Simulator (FSS) with hard panels and stimulated digital control and information system are in the final acceptance test process with the end customer, Taiwan Power Company
Design process of the nanofluid injection mechanism in nuclear power plants

Directory of Open Access Journals (Sweden)

Bang In Choel

2011-01-01

Full Text Available Abstract Nanofluids, which are engineered suspensions of nanoparticles in a solvent such as water, have been found to show enhanced coolant properties such as higher critical heat flux and surface wettability at modest concentrations, which is a useful characteristic in nuclear power plants (NPPs. This study attempted to provide an example of engineering applications in NPPs using nanofluid technology. From these motivations, the conceptual designs of the emergency core cooling systems (ECCSs assisted by nanofluid injection mechanism were proposed after following a design framework to develop complex engineering systems. We focused on the analysis of functional requirements for integrating the conventional ECCSs and nanofluid injection mechanism without loss of performance and reliability. Three candidates of nanofluid-engineered ECCS proposed in previous researches were investigated by applying axiomatic design (AD in the manner of reverse engineering and it enabled to identify the compatibility of functional requirements and potential design vulnerabilities. The methods to enhance such vulnerabilities were referred from TRIZ and concretized for the ECCS of the Korean nuclear power plant. The results show a method to decouple the ECCS designs with the installation of a separate nanofluids injection tank adjacent to the safety injection tanks such that a low pH environment for nanofluids can be maintained at atmospheric pressure which is favorable for their injection in passive manner.
Challenges associated with the design of underground grinding plant at McArthur River project

International Nuclear Information System (INIS)

Jamrozek, J.S.

2000-01-01

McArthur River is an unique high grade uranium underground mine. Ore grinding and thickening are part of the underground operation. The grinding circuit is designed to operate in conditions different from conventional plant environments. Design of the grinding plant was a collective effort of a multi-disciplinary engineering team closely cooperating with project operating personnel. The equipment had to be selected to reflect widely varying ore properties. A user-friendly plant layout provides access to equipment inspections, services, and the delivery of necessary components. The size of the grinding chamber was limited in order to keep the rock stress levels within allowable values. All underground equipment brought to the construction site was restricted in size and weight. Plant construction faced limited storage space underground, tight erection sequencing, and schedule. Plant ventilation is a critical design feature. It efficiently removes radioactive dust from work areas, eliminates stagnant air pockets, and separates clean air from contaminated air areas. Radiation shielding on the equipment is designed to correspond with operational and maintenance functions. Plant operation is remotely controlled and requires little attendance. Video cameras are used on critical equipment and in controlled access areas. An extensive program of preventive and predictive maintenance allows highly reliable plant operation. (author)
Safety design of Qinshan Nuclear Power Plant

International Nuclear Information System (INIS)

Ouyang Yu; Zhang Lian; Du Shenghua; Zhao Jiayu

1984-01-01

Safety issues have been greatly emphasized through the design of the Qinshan Nuclear Power Plant. Reasonable safety margine has been taken into account in the plant design parameters, the design incorporated various safeguard systems, such as engineering safety feature systems, safety protection systems and the features to resist natural catastrophes, e. g. earthquake, hurricanes, tide and so on. Preliminary safety analysis and environmental effect assessment have been done and anti-accident provisions and emergency policy were carefully considered. Qinshan Nuclear Power Plant safety related systems are designed in accordance with the common international standards established in the late 70's, as well as the existing engineering standard of China
Natural gas infrastructure requirements for merchant plant

International Nuclear Information System (INIS)

Sukaly, B.

1998-01-01

Merchant power plants are complicated with diverse risks. Of course where there are risks there are opportunities for reward. Creating an effective merchant plant requires a strong organization that is committed to marketing, trading and risk management. The organization must have the infrastructure to capitalize on the opportunities a merchant plant provides. The market dynamics are ever changing and move at incredible speeds--what was a moneymaking deal yesterday is no longer valid today. The merchant plant owner is the expert in setting up the actual infrastructure for trading the various commodities, including forward pricing, cash and physical trades, transportation and operation for maximizing the plant's potential. Optionally, the plant's risk profile and a risk management program are the key factors in determining the sucres of the merchant plant project
Safeguards by Design at the Encapsulation Plant in Finland

International Nuclear Information System (INIS)

Ingegneri, M.; Baird, K.; Park, W.-S.; Coyne, J.M.; Enkhjin, L.; Chew, L.S.; Plenteda, R.; Sprinkle, J.; Yudin, Y.; Ciuculescu, C.; Koutsoyannopoulos, C.; Murtezi, M.; Schwalbach, P.; Vaccaro, S.; Pekkarinen, J.; Thomas, M.; Zein, A.; Honkamaa, T.; Hamalainen, M.; Martikka, E.; Moring, M.; Okko, O.

2015-01-01

Finland has launched a spent fuel disposition project to encapsulate all of its spent fuel assemblies and confine the disposal canisters in a deep geological repository. The construction of the underground premises started several years ago with the drilling, blasting and reinforcement of tunnels and shafts to ensure the safe deep underground construction and disposal techniques in the repository, while the design of the encapsulation plant (EP) enters the licencing phase preliminary to its construction. The spent fuel assemblies, which have been safeguarded for decades at the nuclear power plants, are going to be transported to the EP, loaded into copper canisters and stored in underground tunnels where they become inaccessible after backfilling. Safeguards measures are needed to ensure that final spent fuel verification is performed before its encapsulation and that no nuclear material is diverted during the process. This is an opportunity for the inspectorates to have the infrastructure necessary for the safeguards equipment incorporated in the design of the encapsulation plant before licencing for construction occurs. The peculiarity of this project is that it is going to run for more than a century. Therefore, significant changes are to be expected in the technical capabilities available for implementing safeguards (e.g., verification techniques and instruments), as well as in the process itself, e.g., redesign for the encapsulation of future fuel types. For these reasons a high degree of flexibility is required in order to be able to shift to different solutions at a later stage while minimizing the interference with the licencing process and facility operations. This paper describes the process leading to the definition of the technical requirements by IAEA and Euratom to be incorporated in the facility's design. (author)
9 CFR 590.24 - Egg products plants requiring continuous inspection.

Science.gov (United States)

2010-01-01

... 9 Animals and Animal Products 2 2010-01-01 2010-01-01 false Egg products plants requiring..., DEPARTMENT OF AGRICULTURE EGG PRODUCTS INSPECTION INSPECTION OF EGGS AND EGG PRODUCTS (EGG PRODUCTS INSPECTION ACT) Scope of Inspection § 590.24 Egg products plants requiring continuous inspection. No plant in...
Design and Application of Cables and Overhead Lines in Wind Power Plants

Energy Technology Data Exchange (ETDEWEB)

Behnke, M. R. [IEEE PES Wind Plant Collector System Design Working Group; Bellei, T.A. [IEEE PES Wind Plant Collector System Design Working Group; Bloethe, W.G. [IEEE PES Wind Plant Collector System Design Working Group; Bradt, M. [IEEE PES Wind Plant Collector System Design Working Group; Brooks, C. [IEEE PES Wind Plant Collector System Design Working Group; Camm, E H [IEEE PES Wind Plant Collector System Design Working Group; Dilling, W. [IEEE PES Wind Plant Collector System Design Working Group; Goltz, B. [IEEE PES Wind Plant Collector System Design Working Group; Hermanson, J. [IEEE PES Wind Plant Collector System Design Working Group; Li, J. [IEEE PES Wind Plant Collector System Design Working Group; Loy, P. [IEEE PES Wind Plant Collector System Design Working Group; McLean, K. [IEEE PES Wind Plant Collector System Design Working Group; Niemira, J. [IEEE PES Wind Plant Collector System Design Working Group; Nuckles, K. [IEEE PES Wind Plant Collector System Design Working Group; Patino, J. [IEEE PES Wind Plant Collector System Design Working Group; Reza, M [IEEE PES Wind Plant Collector System Design Working Group; Richardson, B. [IEEE PES Wind Plant Collector System Design Working Group; Samaan, N. [IEEE PES Wind Plant Collector System Design Working Group; Schoene, Jens [IEEE PES Wind Plant Collector System Design Working Group; Smith, Travis M [ORNL; Snyder, Isabelle B [ORNL; Starke, Michael R [ORNL; Tesch, M. [IEEE PES Wind Plant Collector System Design Working Group; Walling, R. [IEEE PES Wind Plant Collector System Design Working Group; Zahalka, G. [IEEE PES Wind Plant Collector System Design Working Group

2010-01-01

This paper presents a summary of the most impor- tant considerations for wind power plant collection system un- derground and overhead cable designs. Various considerations, including conductor selection, soil thermal properties, installa- tion methods, splicing, concentric grounding, and NESC/NEC requirements are discussed.

Examples for cost reduction in the design of a WWER-1000 nuclear power plant

International Nuclear Information System (INIS)

Kukkola, T.

1991-01-01

In a design project during recent years, a version for Finnish conditions has been and is being developed based on the Soviet WWER-1000 PWR plant with four horizontal steam generators. The plant will have a double containment. The inner containment will be a dry full pressure prestressed concrete containment with liner and the secondary containment will be made of ordinary concrete. Four train safety approach is adopted. It is supposed that the plant is to be designed according to the present Finnish safety requirements, e.g. severe reactor accidents are considered. When striving at an economic plant no compromises are made as far as safety is concerned. This paper describes possible cost reduction by redesigning the main technical equipment. (author). 1 ref
Receiver subsystem analysis report (RADL Item 4-1). The 10-MWe solar thermal central-receiver pilot plant: Solar-facilities design integration

Science.gov (United States)

1982-04-01

The results of thermal hydraulic, design for the stress analyses which are required to demonstrate that the receiver design for the Barstow Solar Pilot Plant satisfies the general design and performance requirements during the plant's design life are presented. Recommendations are made for receiver operation. The analyses are limited to receiver subsystem major structural parts (primary tower, receiver unit core support structure), pressure parts (absorber panels, feedwater, condensate and steam piping/components, flash tank, and steam mainfold) and shielding.
Design verification for large reprocessing plants (Proposed procedures)

International Nuclear Information System (INIS)

Rolandi, G.

1988-07-01

In the 1990s, four large commercial reprocessing plants will progressively come into operation: If an effective and efficient safeguards system is to be applied to these large and complex plants, several important factors have to be considered. One of these factors, addressed in the present report, concerns plant design verification. Design verification provides an overall assurance on plant measurement data. To this end design verification, although limited to the safeguards aspects of the plant, must be a systematic activity, which starts during the design phase, continues during the construction phase and is particularly performed during the various steps of the plant's commissioning phase. The detailed procedures for design information verification on commercial reprocessing plants must be defined within the frame of the general provisions set forth in INFCIRC/153 for any type of safeguards related activities and specifically for design verification. The present report is intended as a preliminary contribution on a purely technical level, and focusses on the problems within the Agency. For the purpose of the present study the most complex case was assumed: i.e. a safeguards system based on conventional materials accountancy, accompanied both by special input and output verification and by some form of near-real-time accountancy involving in-process inventory taking, based on authenticated operator's measurement data. C/S measures are also foreseen, where necessary to supplement the accountancy data. A complete ''design verification'' strategy comprehends: informing the Agency of any changes in the plant system which are defined as ''safeguards relevant''; ''reverifying by the Agency upon receiving notice from the Operator on any changes, on ''design information''. 13 refs
New technologies for lower-cost design and construction of new nuclear power plants. Annex 20

International Nuclear Information System (INIS)

Ritterbusch, S.E.; Bryan, R.E.; Harmon, D.L.

2002-01-01

Electric Power Research Institute studies indicate that in order to be competitive with gas-fired electric power plant capital costs, new nuclear plant capital cost in the USA must be decreased by at least 35% to 40% relative to costs of some Advanced Light Water Reactors designed in the early 1990s. To address this need, the U. S. Department of Energy is sponsoring three separate projects under its Nuclear Energy Research Initiative. These projects are the Risk-Informed Assessment of Regulatory and Design Requirements for Future Nuclear Power Plants, the Smart Equipment Nuclear Power Plant Program, and the Design, Procure, Construct, Install and Test Program. The goal of the Design-Construction program is reduction of the complete nuclear plant design-procure-construct-install-test cycle schedule and cost. A 3D plant model was combined with a construction schedule to produce a 4D visualization of plant construction, which was then used to analyze plant construction methods. Insights include the need for concurrent engineering, a plant-wide central database, and use of the World-Wide WEB. The goal of Smart Equipment program is to design, develop, and evaluate the methods for implementing smart equipment and predictive maintenance technology. 'Smart' equipment means components and systems that are instrumented and monitored to detect incipient failures in order to improve their reliability. The resulting smart equipment methods will be combined with a more risk-informed regulatory approach to allow plant designers to (1) simplify designs without compromising overall reliability and safety and (2) maintain more reliable plants at lower cost. Initial results show that rotating equipment such as charging pumps would benefit most from smart instrumentation and that the technique of Bayesian Belief Networks would be most appropriate for providing input to a health monitoring system. (author)
Divertor conceptual designs for a fusion power plant

International Nuclear Information System (INIS)

Norajitra, P.; Ihli, T.; Janeschitz, G.; Abdel-Khalik, S.; Mazul, I.; Malang, S.

2007-01-01

The development of a divertor concept for post-ITER fusion power plants is deemed to be an urgent task to meet the EU Fast Track scenario. Developing a divertor is particularly challenging due to the wide range of requirements to be met including the high incident peak heat flux, the blanket design with which the divertor has to be integrated, sputtering erosion of the plasma-facing material caused by the incident a particles, radiation effects on the properties of structural materials, and efficient recovery and conversion of the divertor thermal power (∝15% of the total fusion thermal power) by maximizing the coolant operating temperature while minimizing the pumping power. In the course of the EU PPCS, three near-term (A, B and AB) and two advanced power plant models (C, D) were investigated. Model A utilizes a water-cooled lead-lithium (WCLL) blanket and a water-cooled divertor with a peak heat flux of 15 MW/m 2 . Model B uses a He-cooled ceramics/beryllium pebble bed (HCPB) blanket and a He-cooled divertor concept (10 MW/m 2 ). Model AB uses a He-cooled lithium-lead (HCLL) blanket and a He-cooled divertor concept (10 MW/m 2 ). Model C is based on a dual-coolant (DC) blanket (lead/lithium self-cooled bulk and He-cooled structures) and a He-cooled divertor (10 MW/m 2 ). Model D employs a self-cooled lead/lithium (SCLL) blanket and lead-lithiumcooled divertor (5 MW/m 2 ). The values in parenthesis correspond to the maximum peak heat fluxes required. It can be noted that the helium-cooled divertor is used in most of the EU plant models; it has also been proposed for the US ARIES-CS reactor study. Since 2002, it has been investigated extensively in Europe under the PPCS with the goal of reaching a maximum heat flux of at least 10 MW/m2. Work has covered many areas including conceptual design, analysis, material and fabrication issues, and experiments. Generally, the helium-cooled divertor is considered to be a suitable solution for fusion power plants, as it
Hot cell design in the vitrification plant China

International Nuclear Information System (INIS)

Jiang Yubo; Wang Guangkai; Zhang Wei; Liang Runan; Dou Yuan

2015-01-01

In the area of reprocessing and radioactive waste management, gloveboxes and cells are a kind of non-standard equipments providing an isolated room to operate radioactive material inside, while the operator outside with essential biological shield and protection. The hot cell is a typical one, which could handle high radioactive material with various operating means and tight enclosure. The dissertation is based on Vitrification Plant China, a cooperation project between China and Germany. For the sino-western difference in design philosophy, it was presented how to draft an acceptable design proposal of applicable huge hot cells by analysing the design requirements, such as radioprotection, observation, illumination, remote handling, transportation, maintenance and decontamination. The construction feasibility of hot cells was also approved. Thanks to 3D software Autodesk Inventor, digital hot cell was built to integrate all the interfaces inside, which validated the design by checking the mechanical interference. (author)
Towards prioritizing flexibility in the design and construction of concentrating solar power plants

DEFF Research Database (Denmark)

Topel, Monika; Lundqvist, Mårten; Haglind, Fredrik

2017-01-01

In the operation and maintenance of concentrating solar power plants, high operational flexibility is required in order to withstand the variability from the inherent solar fluctuations. However, during the development phases of a solar thermal plant, this important objective is overlooked...... as a relevant factor for cost reduction in the long term. This paper will show the value of including flexibility aspects in the design of a concentrating solar power plant by breaking down their potential favorable impact on the levelized cost of electricity (LCOE) calculations. For this, three scenarios...... to include flexibility as a design objective are analyzed and their potential impact on the LCOE is quantified. The scenarios were modeled and analyzed using a techno-economic model of a direct steam generation solar tower power plant. Sensitivity studies were carried out for each scenario, in which...
Leaf-age and soil-plant relationships: key factors for reporting trace-elements hyperaccumulation by plants and design applications.

Science.gov (United States)

Losfeld, Guillaume; L'Huillier, Laurent; Fogliani, Bruno; Mc Coy, Stéphane; Grison, Claude; Jaffré, Tanguy

2015-04-01

Relationships between the trace-elements (TE) content of plants and associated soil have been widely investigated especially to understand the ecology of TE hyperaccumulating species to develop applications using TE phytoextraction. Many studies have focused on the possibility of quantifying the soil TE fraction available to plants, and used bioconcentration (BC) as a measure of the plants ability to absorb TE. However, BC only offers a static view of the dynamic phenomenon of TE accumulation. Accumulation kinetics are required to fully account for TE distributions in plants. They are also crucial to design applications where maximum TE concentrations in plant leaves are needed. This paper provides a review of studies of BC (i.e. soil-plant relationships) and leaf-age in relation to TE hyperaccumulation. The paper focuses of Ni and Mn accumulators and hyperaccumulators from New Caledonia who were previously overlooked until recent Ecocatalysis applications emerged for such species. Updated data on Mn hyperaccumulators and accumulators from New Caledonia are also presented and advocate further investigation of the hyperaccumulation of this element. Results show that leaf-age should be considered in the design of sample collection and allowed the reclassification of Grevillea meisneri known previously as a Mn accumulator to a Mn hyperaccumulator.
Safety and security aspects in design of digital safety I and C in nuclear power plants

International Nuclear Information System (INIS)

Ding, Yongjian; Waedt, Karl

2016-01-01

The paper describes a safety objective oriented systematic design approach of digital (computerized) safety I and C in modern nuclear power plants which considers the plant safety requirements as well as cybersecurity needs. The defence in depth philosophy is applied by using different defence lines in the I and C architecture and protection zones in the plant IT environment.
Safety and security aspects in design of digital safety I and C in nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Ding, Yongjian [University of Applied Sciences Magdeburg-Stendal, Magdeburg (Germany). Inst. of Electrical Engineering; Waedt, Karl [Areva GmbH, Erlangen (Germany). PEAS-G

2016-05-15

The paper describes a safety objective oriented systematic design approach of digital (computerized) safety I and C in modern nuclear power plants which considers the plant safety requirements as well as cybersecurity needs. The defence in depth philosophy is applied by using different defence lines in the I and C architecture and protection zones in the plant IT environment.
General design criteria for diesel-generator sets for nuclear power plants

International Nuclear Information System (INIS)

Rangarao, G.

1975-01-01

The design criteria for diesel-generators for nuclear power plants are examined. Applicable standards, loading, design performance, and characteristics to be considered in the selection of diesel-generator set and its auxiliary system are discussed. Also, engineered safety features loads together with loss of power safe shutdown loads and their starting sequence, analysis of voltage and frequency response and the diesel-generator ability to start various load blocks successfully to meet the reactor emergency core cooling requirements are discussed
MHI - Westinghouse joint FBR tank plant design

International Nuclear Information System (INIS)

Arnold, W.H.; Vijuk, R.M.; Aoki, I.; Messhil, T.

1988-01-01

Mitsubishi Heavy Industries and Westinghouse Advanced Energy Systems Division have combined their experience and capabilities to design a tank type fast breeder reactor plant. This tank type reactor has been refined and improved during the last three years to better compete in cost, safety, and operation with alternative power plants. This Mitsubishi/Westinghouse joint design offers economic advantages due to the use of steel structures, modular construction, nitrogen cells for the intermediate loops, reactor cavity air cooling and the use of the guard vessel as the containment vessel. Inherent characteristics in the reactor design provide protection to the public and the plant investment
Design of community biogas plant for cooking and lighting from cowdung (a proposal case study)

International Nuclear Information System (INIS)

Mekki, Ibtisam I.

1999-01-01

A calculation design for community biogas plant was proposed for a village in Sudan consisting 100 families, each consisting of 6 persons (adult). Two children are equivalent to one person. Based on the requirement for cooking and lighting, the community biogas plant is deigned for production of total biogas of 360 m-3 gas per day. This volume of the gas will be produced from total dung per day of 6000kg. This amount of dung will expect to obtained from 600 cows. The actual digester volume equal to 60.5m-3. The layout of the plant was designed of two identical plants, each consisted of 3 digester, i.e. 6digesters will be needed.(Author)
Nuclear plant requirements during power system restoration

International Nuclear Information System (INIS)

Adamski, G.; Jenkins, R.; Gill, P.

1995-01-01

This paper is one of a series presented on behalf of the System Operation Subcommittee with the intent of focusing industry attention on power system restoration issues. This paper discusses a number of nuclear power plant requirements that require special attention during power system restoration
Biosafety Procedure for Safe Handling of Genetically Modified Plant Materials in Bio Design Facility

International Nuclear Information System (INIS)

Zaiton Ahmad; Shuhaimi Shamsudin; Mohamed Najli Mohamed Yasin; Affrida Abu Hassan; Mohd Zaid Hassan; Rusli Ibrahim

2015-01-01

Bio Design Facility is the specifically designed glass house for propagation, screening and analysis of high quality plant varieties developed through biotechnology or a combination of nuclear technology and biotechnology. High quality plant varieties especially genetically modified plants (GMO) require a special glass house facility for propagation and screening to isolate them from cross-pollinating with wild type varieties in surrounding ecosystem, and for carrying out evaluation of possible risks of the plants to human, animal and environment before they are proven safe for field trials or commercial release. This facility which was developed under the Ninth Malaysia Plan is classified as the Plant Containment Level 2 and is compliance with the bio safety regulations and guidance for the safe release of GMO according to Malaysian Bio safety Act 2007. Bio Design Facility is fully operational since 2010 and in 2012, it has also been certified as the glass house for post-entry quarantine by The Department of Agriculture. This paper summarizes the bio safety procedure for a safe, controlled and contained growing and evaluation of GMO in Bio Design Facility. This procedure covers the physical (containment and equipment's) and operational (including responsibility, code of practice, growing, decontamination and disposal of plant materials, emergency and contingency plan) aspects of the facility. (author)
Design-Only Conceptual Design Report: Plutonium Immobilization Plant

International Nuclear Information System (INIS)

DiSabatino, A.; Loftus, D.

1999-01-01

This design-only conceptual design report was prepared to support a funding request by the Department of Energy Office of Fissile Materials Disposition for engineering and design of the Plutonium Immobilization Plant, which will be used to immobilize up to 50 tonnes of surplus plutonium. The siting for the Plutonium Immobilization Plant will be determined pursuant to the site-specific Surplus Plutonium Disposition Environmental Impact Statement in a Plutonium Deposition Record of Decision in early 1999. This document reflects a new facility using the preferred technology (ceramic immobilization using the can-in-canister approach) and the preferred site (at Savannah River). The Plutonium Immobilization Plant accepts plutonium from pit conversion and from non-pit sources and, through a ceramic immobilization process, converts the plutonium into mineral-like forms that are subsequently encapsulated within a large canister of high-level waste glass. The final immobilized product must make the plutonium as inherently unattractive and inaccessible for use in nuclear weapons as the plutonium in spent fuel from commercial reactors and must be suitable for geologic disposal. Plutonium immobilization at the Savannah River Site uses: (1) A new building, the Plutonium Immobilization Plant, which will convert non-pit surplus plutonium to an oxide form suitable for the immobilization process, immobilize plutonium in a titanate-based ceramic form, place cans of the plutonium-ceramic forms into magazines, and load the magazines into a canister; (2) The existing Defense Waste Processing Facility for the pouring of high-level waste glass into the canisters; and (3) The Actinide Packaging and Storage Facility to receive and store feed materials. The Plutonium Immobilization Plant uses existing Savannah River Site infra-structure for analytical laboratory services, waste handling, fire protection, training, and other support utilities and services. The Plutonium Immobilization Plant
Shielding design for better plant availability

International Nuclear Information System (INIS)

Biro, G.G.

1975-01-01

Design methods are described for providing a shield system for nuclear power plants that will facilitate maintenance and inspection, increase overall plant availability, and ensure that man-rem exposures are as low as practicable
Towards intelligent automation of power plant design and operations: The role of interactive simulations and distributed expert systems

International Nuclear Information System (INIS)

Otaduy, P.J.

1992-01-01

The design process of a power plant can be viewed as machine- chromosome engineering: When the final layout is implemented, the lifetime operating characteristics, constraints, strengths, and weaknesses of the resulting power-plant-specimen are durably determined. Hence, the safety, operability, maneuverability, availability, maintenance requirements, and costs of a power plant are directly related to the goodness of its electromechanical-genes. This paper addresses the desirability of incorporating distributed computing, distributed object management, and multimedia technologies to power plant engineering, in particular, to design and operations. The promise these technologies have for enhancing the quality and amount of engineering knowledge available, concurrently, online, to plant designers, maintenance crews, and operators is put into perspective. The role that advanced interactive simulations and expert systems will play in the intelligent automation of power plant design and operations is discussed
GTHTR300—A nuclear power plant design with 50% generating efficiency

International Nuclear Information System (INIS)

Sato, Hiroyuki; Yan, Xing L.; Tachibana, Yukio; Kunitomi, Kazuhiko

2014-01-01

Highlights: • GTHTR300 reported 10 years ago is updated. • Cycle and reactor core designs as well as turbine blade material are improved. • The study showed that GTHTR300 is able to yield a net plant efficiency of 50.4%. - Abstract: GTHTR300 is a gas turbine high temperature reactor power generation plant design. The baseline design reported by Japan Atomic Energy Agency a decade ago attained 45.6% net efficiency. Technological improvements have since been made that make further increase in efficiency practical: first, the cycle parameters are upgraded by utilizing the newly acquired design data including those from component tests. Next, the core design is optimized to raise the reactor outlet coolant temperature from the baseline of 850 °C to the level of 950 °C demonstrated on the long-term test reactor operation. Both core physics and thermal hydraulics are investigated to demonstrate the corresponding temperature rise is within the design limit so that the existing fuel design can continue to apply. Finally, an advanced type of turbine blade material that has only recently entered in commercial service in aircraft engine is found to be useable for this design to realize a turbine inlet temperature of 950 °C without requiring blade cooling. As detailed in this paper, these design improvements result in a nearly 5% gain in overall plant efficiency and enable the GTHTR300 to break the 50% efficiency barrier of nuclear plant while using only the existing technologies. This result is expected to contribute to the early market deployment of high temperature gas-cooled reactor
GTHTR300—A nuclear power plant design with 50% generating efficiency

Energy Technology Data Exchange (ETDEWEB)

Sato, Hiroyuki, E-mail: sato.hiroyuki09@jaea.go.jp; Yan, Xing L.; Tachibana, Yukio; Kunitomi, Kazuhiko

2014-08-15

Highlights: • GTHTR300 reported 10 years ago is updated. • Cycle and reactor core designs as well as turbine blade material are improved. • The study showed that GTHTR300 is able to yield a net plant efficiency of 50.4%. - Abstract: GTHTR300 is a gas turbine high temperature reactor power generation plant design. The baseline design reported by Japan Atomic Energy Agency a decade ago attained 45.6% net efficiency. Technological improvements have since been made that make further increase in efficiency practical: first, the cycle parameters are upgraded by utilizing the newly acquired design data including those from component tests. Next, the core design is optimized to raise the reactor outlet coolant temperature from the baseline of 850 °C to the level of 950 °C demonstrated on the long-term test reactor operation. Both core physics and thermal hydraulics are investigated to demonstrate the corresponding temperature rise is within the design limit so that the existing fuel design can continue to apply. Finally, an advanced type of turbine blade material that has only recently entered in commercial service in aircraft engine is found to be useable for this design to realize a turbine inlet temperature of 950 °C without requiring blade cooling. As detailed in this paper, these design improvements result in a nearly 5% gain in overall plant efficiency and enable the GTHTR300 to break the 50% efficiency barrier of nuclear plant while using only the existing technologies. This result is expected to contribute to the early market deployment of high temperature gas-cooled reactor.

Comparison of the General Electric BWR/6 standard plant design to the IAEA NUSS codes and guides

International Nuclear Information System (INIS)

D'Ardenne, W.H.; Sherwood, G.G.

1985-01-01

The General Electric BWR/6 Mark III standard plant design meets or exceeds current requirements of published International Atomic Energy Agency (IAEA) Nuclear Safety Standards (NUSS) codes and guides. This conclusion is based on a review of the NUSS codes and guides by General Electric and by the co-ordinated US review of the NUSS codes and guides during their development. General Electric compared the published IAEA NUSS codes and guides with the General Electric design. The applicability of each code and guide to the BWR/6 Mark III standard plant design was determined. Each code or guide was reviewed by a General Electric engineer knowledgeable about the structures, systems and components addressed and the technical area covered by that code or guide. The results of this review show that the BWR/6 Mark III standard plant design meets or exceeds the applicable requirements of the published IAEA NUSS codes and guides. The co-ordinated US review of the IAEA NUSS codes and guides corroborates the General Electric review. In the co-ordinated US review, the USNRC and US industry organizations (including General Electric) review the NUSS codes and guides during their development. This review ensures that the NUSS codes and guides are consistent with the current US government regulations, guidance and regulatory practices, US voluntary industry codes and standards, and accepted US industry design, construction and operational practices. If any inconsistencies are identified, comments are submitted to the IAEA by the USNRC. All US concerns submitted to the IAEA have been resolved. General Electric design reviews and the Final Design Approval (FDA) issued by the USNRC have verified that the General Electric BWR/6 Mark III standard plant design meets or exceeds the current US requirements, guidance and practices. Since these requirements, guidance and practices meet or exceed those of the NUSS codes and guides, so does the General Electric design. (author)
LBB considerations for a new plant design

Energy Technology Data Exchange (ETDEWEB)

Swamy, S.A.; Mandava, P.R.; Bhowmick, D.C.; Prager, D.E. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

1997-04-01

The leak-before-break (LBB) methodology is accepted as a technically justifiable approach for eliminating postulation of Double-Ended Guillotine Breaks (DEGB) in high energy piping systems. This is the result of extensive research, development, and rigorous evaluations by the NRC and the commercial nuclear power industry since the early 1970s. The DEGB postulation is responsible for the many hundreds of pipe whip restraints and jet shields found in commercial nuclear plants. These restraints and jet shields not only cost many millions of dollars, but also cause plant congestion leading to reduced reliability in inservice inspection and increased man-rem exposure. While use of leak-before-break technology saved hundreds of millions of dollars in backfit costs to many operating Westinghouse plants, value-impacts resulting from the application of this technology for future plants are greater on a per plant basis. These benefits will be highlighted in this paper. The LBB technology has been applied extensively to high energy piping systems in operating plants. However, there are differences between the application of LBB technology to an operating plant and to a new plant design. In this paper an approach is proposed which is suitable for application of LBB to a new plant design such as the Westinghouse AP600. The approach is based on generating Bounding Analyses Curves (BAC) for the candidate piping systems. The general methodology and criteria used for developing the BACs are based on modified GDC-4 and Standard Review Plan (SRP) 3.6.3. The BAC allows advance evaluation of the piping system from the LBB standpoint thereby assuring LBB conformance for the piping system. The piping designer can use the results of the BACs to determine acceptability of design loads and make modifications (in terms of piping layout and support configurations) as necessary at the design stage to assure LBB for the, piping systems under consideration.
Sizes of secondary plant components for modularized IRIS balance of plant design

International Nuclear Information System (INIS)

Williamson, Martin; Townsend, Lawrence

2003-01-01

Herein we report on a conceptual design for a balance of plant (BOP) layout to coordinate with IRIS-like plants. The report consists of results of calculations that sizes of various BOP components. These calculations include the thermodynamic analyses and general sizing of the components in order to determine plant capability and plant layout for studies on modularity and transportability. Mathematical modeling of the BOP system involves a modified ORCENT2 code as well as standard heat transfer methods. Using typical values for PWR type plants, a general BOP design, and IRIS steam generator values, an ORCENT2 heat balance is carried out for the secondary side of the plant. Using the ORCENT2 output, standard heat transfer methods are then used to calculate system performance and component sizes. (author)
MHI-Westinghouse joint FBR tank plant design

International Nuclear Information System (INIS)

Arnold, W.H.; Vijuk, R.M.; Aoki, I.; Meshii, T.

1987-01-01

Mitsubishi Heavy Industries and Westinghouse Advanced Energy Systems Division have combined their experience and capabilities to design a tank type fast breeder reactor plant. This tank type reactor has been refined and improved during the last three years to better compete in cost, satety, and operation with alternative power plants. This Mitsubishi/Westinghouse joint design offers economic advantages due to the use of steel structures, modular construction, nitrogen cells for the intermediate loops, reactor cavity air cooling and the use of the guard vessel as the containment vessel. Inherent characteristics in the reactor design provide protection to the public and the plant investment. (author)
Novel partial-subsidence tower-type boiler design in an ultra-supercritical power plant

International Nuclear Information System (INIS)

Xu, Gang; Xu, Cheng; Yang, Yongping; Fang, Yaxiong; Zhou, Luyao; Zhang, Kai

2014-01-01

Highlights: • The two-pass type and tower-type boilers were compared. • A novel partial-subsidence tower-type boiler design was proposed. • Thermodynamic and economic analyses were quantitatively conducted. • The application of the partial-subsidence boiler to a 700 °C stage unit was further analyzed. - Abstract: An increasing number of tower-type boilers have been applied to ultra-supercritical power plants because of the simple design of the membrane walls and the smooth increase in temperature of such boilers. Nevertheless, the significant height and long steam pipelines of this boiler type will expand the power plant investment cost and increase steam-side pressure losses, especially for higher parameters units requiring high costs of nickel-based alloy materials. Thus, a novel partial-subsidence tower-type boiler design was proposed. In this boiler type, nearly 1/2–2/3 of the boiler height was embedded underground to reduce the height of the boiler and the length of the steam pipelines significantly. Thermodynamic and economic analyses were conducted on a state-of-the-art 1000 MW ultra-supercritical power plant and a prospective 700 °C-stage double reheat power plant. Results showed that the proposed tower-type boiler design could result in a 0.1% point increase in net efficiency and a $0.56/MW h reduction in the cost of electricity in a 1000 MW power plant. This economic benefit was enhanced for power plants with higher steam parameters and larger capacity. The concept of the proposed boiler design may provide a promising method for tower-type boiler applications, especially in new-generation double reheat plants with higher parameters
Seismic design of nuclear power plants - an assessment

International Nuclear Information System (INIS)

Howard, G.E.; Ibanez, P.; Smith, C.B.

1976-01-01

This paper presents a review and evaluation of the design standards and the analytical and experimental methods used in the seismic design of nuclear power plants with emphasis on United States practice. Three major areas were investigated: (a) soils, siting, and seismic ground motion specification; (b) soil-structure interaction; and (c) the response of major nuclear power plant structures and components. The purpose of this review and evaluation program was to prepare an independent assessment of the state-of-the-art of the seismic design of nuclear power plants and to identify seismic analysis and design research areas meriting support by the various organizations comprising the 'nuclear power industry'. Criteria used for evaluating the relative importance of alternative research areas included the potential research impact on nuclear power plant siting, design, construction, cost, safety, licensing, and regulation. (Auth.)
Investigation of human system interface design in nuclear power plant

International Nuclear Information System (INIS)

Feng Yan; Zhang Yunbo; Wang Zhongqiu

2012-01-01

The paper introduces the importance of HFE in designing nuclear power plant, and introduces briefly the content and scope of HFE, discusses human system interface design of new built nuclear power plants. This paper also describes human system interface design of foreign nuclear power plant, and describes in detail human system interface design of domestic nuclear power plant. (authors)
International requirements for life extension of nuclear power plants; Internationale Anforderungen zur Lebensdauerverlaengerung von Kernkraftwerken

Energy Technology Data Exchange (ETDEWEB)

Wernicke, Robert [TUeV NORD SysTec GmbH und Co. KG, Abt. Festigkeit und Konstruktion, Hamburg (Germany)

2009-11-15

Lifetime extension or long-term operation of nuclear facilities are topics of great international significance against the backdrop of a fleet of nuclear power plants of which many have reached 2/3 of their planned life. The article deals with the conditions for, and the specific requirements of, seeking long-term operation of nuclear power plants as established internationally and on the basis of IAEA collections. Technically, long-term operation is possible for many of the nuclear power plants in the world because, normally, they were built on the basis of conservative rules and regulations and, as a consequence, incorporate significant additional safety. Application of requirements to specific plants implies assessments of technical safety which show that conservative design philosophies created reserves and, as a consequence, there is an adequate level of safety also in long-term plant operation. For this purpose, the technical specifications must be revised, necessary additions made, and (international) operating experience taken into account and management of aging established. Two examples are presented to show how the approach to long-term plant operation is put into practice on a national level. (orig.)
Advanced Light Water Reactor Plants System 80+trademark Design Certification Program

International Nuclear Information System (INIS)

1993-01-01

The purpose of this report is to provide a status of the progress that was made towards Design Certification of System 80+trademark during the US government's 1993 fiscal year. The System 80+ Advanced Light Water Reactor (ALWR) is a 3931 MW t (1350 MWe) Pressurized Water Reactor (PWR). The design consists of an essentially complete plant. It is based on evolutionary improvements to the Standardized System 80 nuclear steam supply system in operation at Palo Verde Units 1, 2, and 3, and the Duke Power Company P-81 balance-of-plant (BOP) that was designed and partially constructed at the Cherokee plant site. The System 80/P-81 original design has been substantially enhanced to increase conformance with the EPRI ALWR Utility Requirements Document (URD). Some design enhancements incorporated in the System 80+ design are included in the four units currently under construction in the Republic of Korea. These units form the basis of the Korean standardization program. The full System 80+ standard design has been offered to the Republic of China, in response to their recent bid specification. The ABB-CE Standard Safety Analysis Report (CESSAR-DC) was submitted to the NRC and a Draft Safety Evaluation Report was issued by the NRC in October 1992. CESSAR-DC contains the technical basis for compliance with the EPRI URD for simplified emergency planning. The Nuclear Steam Supply System (NSSS) is the standard ABB-Combustion Engineering two-loop arrangement with two steam generators, two hot legs and four cold legs each with a reactor coolant pump. The System 80+ standard plant includes a sperical steel containment vessel which is enclosed in a concrete shield building, thus providing the safety advantages of a dual containment
Phase I: the pipeline-gas demonstration plant. Demonstration plant engineering and design. Volume 18. Plant Section 2700 - Waste Water Treatment

Energy Technology Data Exchange (ETDEWEB)

None

1981-05-01

Contract No. EF-77-C-01-2542 between Conoco Inc. and the US Department of Energy provides for the design, construction, and operation of a demonstration plant capable of processing bituminous caking coals into clean pipeline quality gas. The project is currently in the design phase (Phase I). This phase is scheduled to be completed in June 1981. One of the major efforts of Phase I is the process and project engineering design of the Demonstration Plant. The design has been completed and is being reported in 24 volumes. This is Volume 18 which reports the design of Plant Section 2700 - Waste Water Treatment. The objective of the Waste Water Treatment system is to collect and treat all plant liquid effluent streams. The system is designed to permit recycle and reuse of the treated waste water. Plant Section 2700 is composed of primary, secondary, and tertiary waste water treatment methods plus an evaporation system which eliminates liquid discharge from the plant. The Waste Water Treatment Section is designed to produce 130 pounds per hour of sludge that is buried in a landfill on the plant site. The evaporated water is condensed and provides a portion of the make-up water to Plant Section 2400 - Cooling Water.
Safety-related concrete structure design and construction of Rokkasho Reprocessing Plant

International Nuclear Information System (INIS)

Morishita, Hideki; Munakata, Yoshinari; Togashi, Akihito

2003-01-01

The Rokkasho Reprocessing Plant of the Japan Nuclear Fuel Co. Ltd., is a facility to reprocess remained uranium without firing and newly formed plutonium contained in spent fuels used at the nuclear power stations, to produce fuels to be repeatedly used. Constructions in this facility has some characteristics shown as follows: 1) radiation shielding and seismic isolated functions like those at the nuclear power plants, 2) reduction of wall thickness based on partially using heavy concrete at walls required for radiation shielding, 3) protective design against fly-coming matters such as aircrafts, 4) construction period reduction based on winter construction and large scale block engineering. Here were described characteristics of designs on radiation shielding, seismic isolated and fly-coming matters protection construction engineering and quality control on concrete. (G.K.)
Designs of new plants of high capacity

International Nuclear Information System (INIS)

Borges R, Diego A

1999-01-01

The Caracas electricity in their desire to lend the best service to the community is doing the necessary projects of generation expansion to fulfill and requirements of demand of next decade in to the metropolitan area. The projects of the new plants of Recifes and El Sitio, have been conceptualized in way of executing the engineering and patterns construction of highest quality and once setting in service, to reach the highest indexes of operative. To reach these goals it is planned to use the most advanced technological designs that are in the market at the moment to world in generation and transmission of power
Conceptual design of a lunar oxygen pilot plant Lunar Base Systems Study (LBSS) task 4.2

Science.gov (United States)

1988-01-01

The primary objective was to develop conceptual designs of two pilot plants to produce oxygen from lunar materials. A lunar pilot plant will be used to generate engineering data necessary to support an optimum design of a larger scale production plant. Lunar oxygen would be of primary value as spacecraft propellant oxidizer. In addition, lunar oxygen would be useful for servicing nonregenerative fuel cell power systems, providing requirements for life support, and to make up oxygen losses from leakage and airlock cycling. Thirteen different lunar oxygen production methods are described. Hydrogen reduction of ilmenite and extraction of solar-wind hydrogen from bulk lunar soil were selected for conceptual design studies. Trades and sensitivity analyses were performed with these models.
Osiris and SOMBRERO inertial fusion power plant designs - summary, conclusions, and recommendations

International Nuclear Information System (INIS)

Meier, Wayne R.

1994-01-01

An 18 month study to evaluate the potential of inertial fusion energy (IFE) for electric power production has been completed. The primary objective of the study was to provide the US Department of Energy with an evaluation of the potential of inertial fusion for electric power production. The study included the conceptual design of two inertial fusion power plants. Osiris uses an induction linac heavy ion beam driver, and SOMBRERO uses a krypton fluoride laser driver. Conceptual designs were completed for the reactors, power conversion and plant facilities, and drivers. Environmental and safety aspects, technical issues, technology development needs, and economics of the final point designs were assessed and compared. This paper summarizes the results and conclusions of the conceptual designs and results of the assessment studies. We conclude that IFE has the potential of producing technically credible designs with environmental, safety, and economics characteristics that are just as attractive as magnetic fusion. Realizing this potential will require additional research and development on target physics, chamber design, target production and injection systems, and drivers. ((orig.))
Site layout and balance of plant design for an accelerator-driven materials processing complex

Energy Technology Data Exchange (ETDEWEB)

Cunliffe, J.; Taussig, R.; Ghose, S. [Bechtel Corporation, San Francisco, CA (United States)] [and others

1995-10-01

High energy proton beam accelerators are under consideration for use in radioisotope production, surplus weapons material destruction, radioactive waste transmutation, and thorium-based energy conversion cycles. While there are unique aspects to each of these applications that must be accommodated in the design of the associated facility, all share a set of fundamental characteristics that in large measure dictate the site layout features and many balance-of-plant (BOP) design requirements found to be common to all. This paper defines these key design determinants and goes on to discuss the manner in which they have been accommodated in the pre-conceptual design for a particular materials production application. An estimate of the costs associated with this BOP design is also presented with the aim of guiding future evaluations where the basic plant designs are similar to that of this specific case.
Supplmental design requirements document enhanced radioactive and mixed waste storage: Phase 5, Project W-113

International Nuclear Information System (INIS)

Ocampo, V.P.

1994-11-01

This Supplemental Design Requirements Document (SDRD) is used to communicate Project W-113 specific plant design information from Westinghouse Hanford Company (WHC) to the United States Department of Energy (DOE) and the cognizant Architect Engineer (A/E). The SDRD is prepared after the completion of the project Conceptual Design report (CDR) and prior to the initiation of definitive design. Information in the SDRD serves two purposes: to convey design requirements that are too detailed for inclusion in the Functional Design Criteria (FDC) report and to serve as a means of change control for design commitments in the Title I and Title II design. The Solid Waste Retrieval Project (W-113) SDRD has been restructured from the equipment based outline used in previous SDRDs to a functional systems outline. This was done to facilitate identification of deficiencies in the information provided in the initial draft SDRD and aid design confirmation. The format and content of this SDRD adhere as closely as practicable to the requirements of WHC-CM-6-1, Standard Engineering Practices for Functional Design Criteria
Structural modules in AP1000 plant design

International Nuclear Information System (INIS)

Prasad, N.; Tunon-Sanjur, L.

2007-01-01

Structural modules are extensively used in AP1000 plant design. The shop manufacturing of modules components improves the quality and reliability of plant structures. The application of modules has a positive impact on construction schedules, and results in substantial savings in the construction cost. This paper describes various types of structural modules used for AP1000 plant structures. CA structural wall modules are steel plate modules with concrete placed, on or within the module, after module installation. The layout and design of the largest CA wall modules, CA01 and CA20, is described in detail. General discussion of structural floor modules, such as the composite and finned floors, is also included. Steel form CB modules (liners) consist of plate reinforced with angle stiffeners and tee sections. The angles and the tee sections are on the concrete side of the plate. Design of CB20 has been included as an example of CB type modules. Design codes and structural concepts related to module designs are discussed. (authors)
Operating experience and systems analysis at Trillo NPP: A program intended for systematic review of plant safety systems to assess design basis requirements compliance

International Nuclear Information System (INIS)

Vega, R. de la

1996-01-01

The program was defined to apply to all plant safety systems and/or systems included in plant Technical Specifications. The goal of the program was to ensure, by systematic design, construction, and commissioning review, the adequacy of safety systems, structures and components to fulfill their safety functions. Also, as a result of the program, it was established that a complete, unambiguous, systematic, design basis definition shall take place. And finally, a complete documental review of the plant design shall result from the program execution
Configuration and engineering design of the ARIES-RS tokamak power plant

International Nuclear Information System (INIS)

Tillack, M.S.; Malang, S.; Waganer, L.; Wang, X.R.; Sze, D.K.; El-Guebaly, L.; Wong, C.P.C.; Crowell, J.A.; Mau, T.K.; Bromberg, L.

1997-01-01

ARIES-RS is a conceptual design study which has examined the potential of an advanced tokamak-based power plant to compete with future energy sources and play a significant role in the future energy market. The design is a 1000 MWe, DT-burning fusion power plant based on the reversed-shear tokamak mode of plasma operation, and using moderately advanced engineering concepts such as lithium-cooled vanadium-alloy plasma-facing components. A steady-state reversed shear tokamak currently appears to offer the best combination of good economic performance and physics credibility for a tokamak-based power plant. The ARIES-RS engineering design process emphasized the attainment of the top-level mission requirements developed in the early part of the study in a collaborative effort between the ARIES Team and representatives from U.S. electric utilities and industry. Major efforts were devoted to develop a credible configuration that allows rapid removal of full sectors followed by disassembly in the hot cells during plant operation. This was adopted as the only practical means to meet availability goals. Use of an electrically insulating coating for the self-cooled blanket and divertor provides a wide design window and simplified design. Optimization of the shield, which is one of the larger cost items, significantly reduced the power core cost by using ferritic steel where the power density and radiation levels are low. An additional saving is made by radial segmentation of the blanket, such that large segments can be reused. The overall tokamak configuration is described here, together with each of the major fusion power core components: the first-wall, blanket and shield; divertor; heating, current drive and fueling systems; and magnet systems. (orig.)
Conceptual design of a laser fusion power plant

International Nuclear Information System (INIS)

Maniscalco, J.A.; Meier, W.R.; Monsler, M.J.

1977-01-01

A conceptual design of a laser fusion power plant is extensively discussed. Recent advances in high gain targets are exploited in the design. A smaller blanket structure is made possible by use of a thick falling region of liquid lithium for a first wall. Major design features of the plant, reactor, and laser systems are described. A parametric analysis of performance and cost vs. design parameters is presented to show feasible design points. A more definitive follow-on conceptual design study is planned

Risk informed life cycle plant design

International Nuclear Information System (INIS)

Hill, Ralph S. III; Nutt, Mark M.

2003-01-01

Many facility life cycle activities including design, construction, fabrication, inspection and maintenance are evolving from a deterministic to a risk-informed basis. The risk informed approach uses probabilistic methods to evaluate the contribution of individual system components to total system performance. Total system performance considers both safety and cost considerations including system failure, reliability, and availability. By necessity, a risk-informed approach considers both the component's life cycle and the life cycle of the system. In the nuclear industry, risk-informed approaches, namely probabilistic risk assessment (PRA) or probabilistic safety assessment (PSA), have become a standard tool used to evaluate the safety of nuclear power plants. Recent studies pertaining to advanced reactor development have indicated that these new power plants must provide enhanced safety over existing nuclear facilities and be cost-competitive with other energy sources. Risk-informed approaches, beyond traditional PRA, offer the opportunity to optimize design while considering the total life cycle of the plant in order to realize these goals. The use of risk-informed design approaches in the nuclear industry is only beginning, with recent promulgation of risk-informed regulations and proposals for risk-informed codes. This paper briefly summarizes the current state of affairs regarding the use of risk-informed approaches in design. Key points to fully realize the benefit of applying a risk-informed approach to nuclear power plant design are then presented. These points are equally applicable to non-nuclear facilities where optimization for cost competitiveness and/or safety is desired. (author)
Design of a fault diagnosis system for next generation nuclear power plants

International Nuclear Information System (INIS)

Zhao, K.; Upadhyaya, B.R.; Wood, R.T.

2004-01-01

A new design approach for fault diagnosis is developed for next generation nuclear power plants. In the nuclear reactor design phase, data reconciliation is used as an efficient tool to determine the measurement requirements to achieve the specified goal of fault diagnosis. In the reactor operation phase, the plant measurements are collected to estimate uncertain model parameters so that a high fidelity model can be obtained for fault diagnosis. The proposed algorithm of fault detection and isolation is able to combine the strength of first principle model based fault diagnosis and the historical data based fault diagnosis. Principal component analysis on the reconciled data is used to develop a statistical model for fault detection. The updating of the principal component model based on the most recent reconciled data is a locally linearized model around the current plant measurements, so that it is applicable to any generic nonlinear systems. The sensor fault diagnosis and process fault diagnosis are decoupled through considering the process fault diagnosis as a parameter estimation problem. The developed approach has been applied to the IRIS helical coil steam generator system to monitor the operational performance of individual steam generators. This approach is general enough to design fault diagnosis systems for the next generation nuclear power plants. (authors)
Nuclear power plant equipment design and construction rules

International Nuclear Information System (INIS)

Boiron, P.

1983-03-01

Presentation of the AFCEN (French association for nuclear power plant equipment design and construction rules) working, of its edition activity and of somes of its edited documents such as RCC-C (design and construction rules for PWR power plant fuel assemblies) and RCC-E (design and construction rules for nuclear facility electrical equipments) [fr
Compiling Utility Requirements For New Nuclear Power Plant Project

International Nuclear Information System (INIS)

Patrakka, Eero

2002-01-01

Teollisuuden Voima Oy (TVO) submitted in November 2000 to the Finnish Government an application for a Decision-in-Principle concerning the construction of a new nuclear power plant in Finland. The actual investment decision can be made first after a positive decision has been made by the Government and the Parliament. Parallel to the licensing process, technical preparedness has been upheld so that the procurement process can be commenced without delay, when needed. This includes the definition of requirements for the plant and preliminary preparation of bid inquiry specifications. The core of the technical requirements corresponds to the specifications presented in the European Utility Requirement (EUR) document, compiled by major European electricity producers. Quite naturally, an amount of modifications to the EUR document are needed that take into account the country- and site-specific conditions as well as the experiences gained in the operation of the existing NPP units. Along with the EUR-related requirements concerning the nuclear island and power generation plant, requirements are specified for scope of supply as well as for a variety of issues related to project implementation. (author)
49 CFR 229.206 - Design requirements.

Science.gov (United States)

2010-10-01

...-climber, emergency egress, emergency interior lighting, and interior configuration design requirements set... 49 Transportation 4 2010-10-01 2010-10-01 false Design requirements. 229.206 Section 229.206..., DEPARTMENT OF TRANSPORTATION RAILROAD LOCOMOTIVE SAFETY STANDARDS Locomotive Crashworthiness Design...
Feedback of operation and maintenance experience into evolutionary plant design (HWRs)

International Nuclear Information System (INIS)

Hedges, K.R.; Sanatkumar, A.; Kwon, Oh-Cheol

1999-01-01

The process of feeding back operation and maintenance information into the CANDU plant design process has been in constant evolution since the beginning of the CANDU program. The commissioning and operation experience from the first commercial reactors at Pickering A and Bruce A was used extensively in the design of the first generation CANDU 6 Plants. These units have been in operation for 15 years, producing electricity at an average lifetime capacity factor of about 85%. In further advancing the CANDU 6 and 9 design, greater emphasis is placed on enhancements that can reduce operational costs and further improve plant performance by reducing the planned outage time. The plant design has been improved to facilitate maintenance scheduling, equipment isolation, maintenance and post maintenance testing. Individual tasks have been analyzed as well as the interaction between tasks during outages to reduce the down time required and simplify the execution of the work. This results in shorter outages, reduced radioactive dose and reduced costs. The Utilities have continued to play an important role in CANDU 6 Evolution. Specifically; the Korea Utility KEPCO has one of the original four CANDU 6 Plants and three of the most modem. Their feedback to the designers has been very helpful. One of the most important feedback processes is through the CANDU Owners Group, which provides information exchange between members. In India eight PHWRs of 220 MWe capacity are in operation. Four reactors, also of 220 MWe capacity are in advanced stages of construction. Site construction work of two units of 500 MWe PHWRs at Tarapur will be taken up shortly. Over the years, during construction and operation of these power stations, a large amount of experience has been accumulated. Operation and maintenance experience is shared with operating stations by intensive participation of design engineers in Station Operation Review meetings, trouble shooting and root cause analysis of problems
Plant cell walls throughout evolution: towards a molecular understanding of their design principles.

Science.gov (United States)

Sarkar, Purbasha; Bosneaga, Elena; Auer, Manfred

2009-01-01

Throughout their life, plants typically remain in one location utilizing sunlight for the synthesis of carbohydrates, which serve as their sole source of energy as well as building blocks of a protective extracellular matrix, called the cell wall. During the course of evolution, plants have repeatedly adapted to their respective niche, which is reflected in the changes of their body plan and the specific design of cell walls. Cell walls not only changed throughout evolution but also are constantly remodelled and reconstructed during the development of an individual plant, and in response to environmental stress or pathogen attacks. Carbohydrate-rich cell walls display complex designs, which together with the presence of phenolic polymers constitutes a barrier for microbes, fungi, and animals. Throughout evolution microbes have co-evolved strategies for efficient breakdown of cell walls. Our current understanding of cell walls and their evolutionary changes are limited as our knowledge is mainly derived from biochemical and genetic studies, complemented by a few targeted yet very informative imaging studies. Comprehensive plant cell wall models will aid in the re-design of plant cell walls for the purpose of commercially viable lignocellulosic biofuel production as well as for the timber, textile, and paper industries. Such knowledge will also be of great interest in the context of agriculture and to plant biologists in general. It is expected that detailed plant cell wall models will require integrated correlative multimodal, multiscale imaging and modelling approaches, which are currently underway.
Plant cell walls throughout evolution: towards a molecular understanding of their design principles

Energy Technology Data Exchange (ETDEWEB)

Sarkar, Purbasha; Bosneaga, Elena; Auer, Manfred

2009-02-16

Throughout their life, plants typically remain in one location utilizing sunlight for the synthesis of carbohydrates, which serve as their sole source of energy as well as building blocks of a protective extracellular matrix, called the cell wall. During the course of evolution, plants have repeatedly adapted to their respective niche,which is reflected in the changes of their body plan and the specific design of cell walls. Cell walls not only changed throughout evolution but also are constantly remodelled and reconstructed during the development of an individual plant, and in response to environmental stress or pathogen attacks. Carbohydrate-rich cell walls display complex designs, which together with the presence of phenolic polymers constitutes a barrier for microbes, fungi, and animals. Throughout evolution microbes have co-evolved strategies for efficient breakdown of cell walls. Our current understanding of cell walls and their evolutionary changes are limited as our knowledge is mainly derived from biochemical and genetic studies, complemented by a few targeted yet very informative imaging studies. Comprehensive plant cell wall models will aid in the re-design of plant cell walls for the purpose of commercially viable lignocellulosic biofuel production as well as for the timber, textile, and paper industries. Such knowledge will also be of great interest in the context of agriculture and to plant biologists in general. It is expected that detailed plant cell wall models will require integrated correlative multimodal, multiscale imaging and modelling approaches, which are currently underway.
Design criteria for the new waste calcining facility at the Idaho Chemical Processing Plant

International Nuclear Information System (INIS)

Anderson, F.H.; Bingham, G.E.; Buckham, J.A.; Dickey, B.R.; Slansky, C.M.; Wheeler, B.R.

1976-01-01

The New Waste Calcining Facility (NWCF) at the Idaho Chemical Processing Plant (ICPP) is being built to replace the existing fluidized-bed, high-level waste calcining facility (WCF). Performance of the WCF is reviewed, equipment failures in WCF operation are examined, and pilot-plant studies on calciner improvements are given in relation to NWCF design. Design features of the NWCF are given with emphasis on process and equipment improvements. A major feature of the NWCF is the use of remote maintenance facilities for equipment with high maintenance requirements, thereby reducing personnel exposures during maintenance and reducing downtime resulting from plant decontamination. The NWCF will have a design net processing rate of 11.36 m 3 of high-level waste per day, and will incorporate in-bed combustion of kerosene for heating the fluidized bed calciner. The off-gas cleaning system will be similar to that for the WCF
Upgrading instrumentation and control in nuclear power plants. Design criteria

International Nuclear Information System (INIS)

Rodriguez Rodriguez, M.C.; Alvarez Menendez, A.

1997-01-01

The use of programmed digital technology in Protection, Control, Monitoring and Information Systems in new generation nuclear power plants, or the use of this technology to replace or upgrade existing systems based on wired analog instrumentation and electromechanical relays, has led to new international standards which establish new design requirements or adapt existing requirements to this technology. Additionally, both regulatory organisations and the industry are discussing the reliability of this technology, regarding common mode failures that may occur in redundant protection channels, due to the use of equipment and software with the same characteristics. The first part of this paper addresses the most important aspects of new international standards regarding classification criteria for I and C systems, equipment and functions, depending on their importance to safety and the design criteria applicable to each category. Special attention is drawn to requirements concerning software quality assurance and the design of new control rooms. The paper then goes on to discuss the different technical solutions being implemented, using equipment and software diversification, in order to prevent the possibility of common mode failures affecting the protection function. (Author)
Nuclear Safeguards Infrastructure Required for the Next Generation Nuclear Plant (NGNP)

Energy Technology Data Exchange (ETDEWEB)

Dr. Mark Schanfein; Philip Casey Durst

2012-07-01

The Next Generation Nuclear Plant (NGNP) is a Very High Temperature Gas-Cooled Reactor (VHTR) to be constructed near Idaho Falls, Idaho The NGNP is intrinsically safer than current reactors and is planned for startup ca. 2021 Safety is more prominent in the minds of the Public and Governing Officials following the nuclear reactor meltdown accidents in Fukushima, Japan The authors propose that the NGNP should be designed with International (IAEA) Safeguards in mind to support export to Non-Nuclear-Weapons States There are two variants of the NGNP design; one using integral Prismatic-shaped fuel assemblies in a fixed core; and one using recirculating fuel balls (or Pebbles) The following presents the infrastructure required to safeguard the NGNP This infrastructure is required to safeguard the Prismatic and Pebble-fueled NGNP (and other HTGR/VHTR) The infrastructure is based on current Safeguards Requirements and Practices implemented by the International Atomic Energy Agency (IAEA) for similar reactors The authors of this presentation have worked for decades in the area of International Nuclear Safeguards and are recognized experts in this field Presentation for INMM conference in July 2012.
Design requirement on HYPER blanket fuel assembly

International Nuclear Information System (INIS)

Hwang, Woan; Lee, B. O.; Nam, C.; Ryu, W. S.; Lee, B. S.; Park, W. S.

2000-07-01

This document describes design requirements which are needed for designing the blanket assembly of the HYPER as design guidance. The blanket assembly of the HYPER consists of blanket fuel rods, mounting rail, spacer, upper nozzle with handling socket, bottom nozzle with mounting rail and skeleton structure. The blanket fuel rod consists of top end plug, bottom end plug with key way, blanket fuel slug, and cladding. In the assembly, the rods are in a triangular pitch array. This report contains functional requirements, performance and operational requirements, interfacing systems requirements, core restraint and interface requirements, design limits and strength requirements, system configuration and essential feature requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements for the blanket fuel assembly of the HYPER
Design of a multivariable controller for a CANDU 600 MWe nuclear power plant using the INA method

International Nuclear Information System (INIS)

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

1984-04-01

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

International Nuclear Information System (INIS)

Asmuss, G.

1980-01-01

The establishment and implementation for a practicable quality assurance program for a nuclear power plant demands a detailed background in the field of engineering, manufacturing, organization and quality assurance. It will be demonstrated with examples to define and control the achievement of quality related activities during the phases of design, procurement, manufactoring, commissioning and operation. In general the quality assurance program applies to all items, processes and services important to safety of nuclear power plant. The classification for safety related and non-safety related items and services demonstrate the levels of quality assurance requirements. The lecture gives an introduction of QA Program preparation under the following topics: -Basic criteria and international requirements - Interaction of QA activities - Modular and product oriented QA programs - Structuring of organization for the QA program - Identification of the main quality assurance functions and required actions - Quality Assurance Program documentation - Documentation of planning of activities - Control of program documents - Definitions. (orig./RW)
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)
Exergy analysis of an IGCC design configuration for Plant Wansley

International Nuclear Information System (INIS)

Tsatsaronis, G.; Tawfik, T.; Lin, L.; Gallaspy, D.T.

1989-01-01

An integrated gasification-combined-cycle power plant design was developed for Georgia Power Company's Plant Wansley. This paper discusses the plant configuration and presents the most important results obtained from a detailed exergy analysis of the plant design. This analysis will be completed in a subsequent paper through an exergoeconomic analysis to identify design improvements for reducing the electricity cost
Improving human reliability through better nuclear power plant system design. Progress report

International Nuclear Information System (INIS)

Golay, M.W.

1995-01-01

The project on open-quotes Development of a Theory of the Dependence of Human Reliability upon System Designs as a Means of Improving Nuclear Power Plant Performanceclose quotes has been undertaken in order to address the important problem of human error in advanced nuclear power plant designs. Most of the creativity in formulating such concepts has focused upon improving the mechanical reliability of safety related plant systems. However, the lack of a mature theory has retarded similar progress in reducing the likely frequencies of human errors. The main design mechanism used to address this class of concerns has been to reduce or eliminate the human role in plant operations and accident response. The plan of work being pursued in this project is to perform a set of experiments involving human subject who are required to operate, diagnose and respond to changes in computer-simulated systems, relevant to those encountered in nuclear power plants. In the tests the systems are made to differ in complexity in a systematic manner. The computer program used to present the problems to be solved also records the response of the operator as it unfolds. Ultimately this computer is also to be used in compiling the results of the project. The work of this project is focused upon nuclear power plant applications. However, the persuasiveness of human errors in using all sorts of electromechanical machines gives it a much greater potential importance. Because of this we are attempting to pursue our work in a fashion permitting broad generalizations
The licensing process of the design modifications of Cernavoda 2 NPP resulting from the operating experience of CANDU plants

International Nuclear Information System (INIS)

Goicea, L.

2005-01-01

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

International Nuclear Information System (INIS)

King, A.C.Y.; Chen, C.

1977-01-01

The time history method has been one of the analytical tools applied in the seismic resistant design of nuclear power plants. The time histories used are required to be consistent with the specified design Spectra. Since the spectra of recorded strong motion earthquake or conventionally generated artificial time history have local peaks and valleys, iteration procedures must be applied to generate the artificial time history with desired spectra. The paper describes a detailed method for generating a time history which is consistent with a specified design spectra. There are several advantages of this method described herein. First of all, frequency content of the time history is well under control. Secondly, if one wishes to generate the three components of an earthquake at one site, the inherent nature of this method will make the correlations among these three components to simulate closely the actual recorded time histories. Thirdly, a single time history can be generated to match a spectra for different damping values. (auth.)
Design requirement on KALIMER control rod assembly duct

International Nuclear Information System (INIS)

Hwang, W.; Kang, H. Y.; Nam, C.; Kim, J. O.; Kim, Y. J.

1998-03-01

This document establishes the design guidelines which are needs for designing the control rod assembly duct of the KALIMER as design requirements. it describes control rod assembly duct of the KALIMER and its requirements that includes functional requirements, performance requirements, interfacing systems, design limits and strength requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements. The control rod system consists of three parts, which are drive mechanism, drive-line, and absorber bundle. This report deals with the absorber bundle and its outer duct only because the others are beyond the scope of fuel system design. The guidelines for design requirements intend to be used for an improved design of the control rod assembly duct of the KALIMER. (author). 19 refs

Design requirement on KALIMER control rod assembly duct

Energy Technology Data Exchange (ETDEWEB)

Hwang, W.; Kang, H. Y.; Nam, C.; Kim, J. O.; Kim, Y. J

1998-03-01

This document establishes the design guidelines which are needs for designing the control rod assembly duct of the KALIMER as design requirements. it describes control rod assembly duct of the KALIMER and its requirements that includes functional requirements, performance requirements, interfacing systems, design limits and strength requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements. The control rod system consists of three parts, which are drive mechanism, drive-line, and absorber bundle. This report deals with the absorber bundle and its outer duct only because the others are beyond the scope of fuel system design. The guidelines for design requirements intend to be used for an improved design of the control rod assembly duct of the KALIMER. (author). 19 refs.
Design of a lunar oxygen production plant

Science.gov (United States)

Radhakrishnan, Ramalingam

1990-01-01

To achieve permanent human presence and activity on the moon, oxygen is required for both life support and propulsion. Lunar oxygen production using resources existing on the moon will reduce or eliminate the need to transport liquid oxygen from earth. In addition, the co-products of oxygen production will provide metals, structural ceramics, and other volatile compounds. This will enable development of even greater self-sufficiency as the lunar outpost evolves. Ilmenite is the most abundant metal-oxide mineral in the lunar regolith. A process involving the reaction of ilmenite with hydrogen at 1000 C to produce water, followed by the electrolysis of this water to provide oxygen and recycle the hydrogen has been explored. The objective of this 1990 Summer Faculty Project was to design a lunar oxygen-production plant to provide 5 metric tons of liquid oxygen per year from lunar soil. The results of this study describe the size and mass of the equipment, the power needs, feedstock quantity and the engineering details of the plant.
Safety of Nuclear Power Plants: Commissioning and Operation. Specific Safety Requirements

International Nuclear Information System (INIS)

2016-01-01

This publication describes the requirements to be met to ensure the safe operation of nuclear power plants. It takes into account developments in areas such as long term operation of nuclear power plants, plant ageing, periodic safety review, probabilistic safety analysis and risk informed decision making processes. In addition, the requirements are governed by, and must apply, the safety objective and safety principles that are established in the IAEA Safety Standards Series No. SF-1, Fundamental Safety Principles. A review of Safety Requirements publications was commenced in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan. The review revealed no significant areas of weakness and resulted in just a small set of amendments to strengthen the requirements and facilitate their implementation, which are contained in the present publication
Osiris and SOMBRERO inertial confinement fusion power plant designs

International Nuclear Information System (INIS)

Meier, W.R.; Bieri, R.L.; Monsler, M.J.

1992-03-01

Conceptual designs and assessments have been completed for two inertial fusion energy (IFE) electric power plants. The detailed designs and results of the assessment studies are presented in this report. Osiris is a heavy-ion-beam (HIB) driven power plant and SOMBRERO is a Krypton-Fluoride (KrF) laser-driven power plant. Both plants are sized for a net electric power of 1000 MWe
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
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.
A formal design verification and validation on the human factors of a computerized information system in nuclear power plants

International Nuclear Information System (INIS)

Lee, Yong Hee; Park, Jae Chang; Cheon, Se Woo; Jung, Kwang Tae; Baek, Seung Min; Han, Seung; Park, Hee Suk; Son, Ki Chang; Kim, Jung Man; Jung Yung Woo

1999-11-01

This report describe a technical transfer under the title of ''A formal design verification and validation on the human factors of a computerized information system in nuclear power plants''. Human factors requirements for the information system designs are extracted from various regulatory and industrial standards and guidelines, and interpreted into a more specific procedures and checklists for verifying the satisfaction of those requirements. A formalized implementation plan is established for human factors verification and validation of a computerized information system in nuclear power plants. Additionally, a Computer support system, named as DIMS-web (design Issue Management System), is developed based upon web internet environment so as to enhance the implementation of the human factors activities. DIMS-Web has three maine functions: supporting requirements review, tracking design issues, and management if issues screening evaluation. DIMS-Web shows its benefits in practice through a trial application to the design review of CFMS for YGN nuclear unit 5 and 6. (author)
A formal design verification and validation on the human factors of a computerized information system in nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Lee, Yong Hee; Park, Jae Chang; Cheon, Se Woo; Jung, Kwang Tae; Baek, Seung Min; Han, Seung; Park, Hee Suk; Son, Ki Chang; Kim, Jung Man; Jung Yung Woo

1999-11-01

This report describe a technical transfer under the title of ''A formal design verification and validation on the human factors of a computerized information system in nuclear power plants''. Human factors requirements for the information system designs are extracted from various regulatory and industrial standards and guidelines, and interpreted into a more specific procedures and checklists for verifying the satisfaction of those requirements. A formalized implementation plan is established for human factors verification and validation of a computerized information system in nuclear power plants. Additionally, a Computer support system, named as DIMS-web (design Issue Management System), is developed based upon web internet environment so as to enhance the implementation of the human factors activities. DIMS-Web has three maine functions: supporting requirements review, tracking design issues, and management if issues screening evaluation. DIMS-Web shows its benefits in practice through a trial application to the design review of CFMS for YGN nuclear unit 5 and 6. (author)
Nonfunctional requirements in systems analysis and design

CERN Document Server

Adams, Kevin MacG

2015-01-01

This book will help readers gain a solid understanding of non-functional requirements inherent in systems design endeavors. It contains essential information for those who design, use, and maintain complex engineered systems, including experienced designers, teachers of design, system stakeholders, and practicing engineers. Coverage approaches non-functional requirements in a novel way by presenting a framework of four systems concerns into which the 27 major non-functional requirements fall: sustainment, design, adaptation, and viability. Within this model, the text proceeds to define each non-functional requirement, to specify how each is treated as an element of the system design process, and to develop an associated metric for their evaluation. Systems are designed to meet specific functional needs. Because non-functional requirements are not directly related to tasks that satisfy these proposed needs, designers and stakeholders often fail to recognize the importance of such attributes as availability, su...
Improvement of availability of PWR nuclear plants through the reduction of the time required for refueling/maintenance outages

International Nuclear Information System (INIS)

Mayers, J.B.; Soth, L.G.

1978-04-01

The objective of the project, conducted by Commonwealth Research Corporation and Westinghouse Electric Corporation, is to identify improvements in procedures and equipment which will reduce the time required for refueling/maintenance outages at PWR nuclear power plants. The outage of Commonwealth Edison Zion Station Unit 1 in March through May of 1976 was evaluated to identify those items which caused delays and those work activities that offer the potential for significant improvements that could reduce the overall duration of the outage and achieve an improvement in the plant's availability for power production. Modifications in procedures have been developed and were evaluated during one or more outages in 1977. Conceptual designs have been developed for equipment modifications to the refueling system that could reduce the time required for the refueling portion of the outage. The purpose of the interim report is to describe those conceptual designs and to assess their impact upon future outages. Recommendations are included for the implementation of these equipment improvements in a continuation of this program as a demonstration of plant availability benefits that can be realized in PWR nuclear plants already in operation or under construction
Structure design of water discharge surge tank of nuclear power plant

International Nuclear Information System (INIS)

Wang Fang; Hou Shuqiang

2015-01-01

Drainage is an important function of water discharge surge tank in nuclear power plant. There is little wall and beam inside the water discharge surge tank due to the requirement of major work, which is different from the general structure. Taking water discharge surge tank of nuclear power plant for example, concerned problems are expatiated in the structure scheme of water discharge surge tank, and important structural components are analyzed. Structural analysis model is established by ANSYS finite element analysis. A comprehensive and numerical analysis is performed for different combinations of structural model, and the internal force of structure is extracted. Finally, suggestions for design of similar structure are proposed. (authors)
Improving mine-mill water network design by reducing water and energy requirements

Energy Technology Data Exchange (ETDEWEB)

Gunson, A.J.; Klein, B.; Veiga, M. [British Columbia Univ., Vancouver, BC (Canada). Norman B. Keevil Inst. of Mining Engineering

2010-07-01

Mining is an energy-intensive industry, and most processing mills use wet processes to separate minerals from ore. This paper discussed water reduction, reuse and recycling options for a mining and mill operation network. A mine water network design was then proposed in order to identify and reduce water and system energy requirements. This included (1) a description of site water balance, (2) a description of potential water sources, (3) a description of water consumers, (4) the construction of energy requirement matrices, and (5) the use of linear programming to reduce energy requirements. The design was used to determine a site water balance as well as to specify major water consumers during mining and mill processes. Potential water supply combinations, water metering technologies, and recycling options were evaluated in order to identify the most efficient energy and water use combinations. The method was used to highlight potential energy savings from the integration of heating and cooling systems with plant water systems. 43 refs., 4 tabs., 3 figs.
The impact of instrumentation and control requirements on the design changes of the Westinghouse ''NSSS'' of Almaraz, Lemoniz and Asco

International Nuclear Information System (INIS)

Gerini, P.M.; Naredo, F.P.; Williams, D.W.

1978-01-01

For the nuclear power plants Almaraz, Lemoniz and Asco the NSSS set is supplied by Westinghouse. Purchasing contracts were signed in 1971 and projects design took into account the compliance with the regulatory requirements for licensing, issued and standing that time. Since 1971 licensing regulations have been subjected to a deep revision due to the issue of new standards and guides and revision of other affecting altogether the engineering design of nuclear power plants. This situation was reasonably reflected on several consecutive design revisions for the case of the Almaraz, Lemoniz and Asco Nuclear plants. This impact, from the viewpoint of the instrumentation and control context, and referred to the NSSS is analyzed in the report. In particular, attention is paid to the safeguards actuation logic, testing capability and physical separation criteria as contemplated into the regulatory requirements starting from 1971.(J.E.de C)
The Swedish Utilities joint approach to form common basis for design requirements for the future

International Nuclear Information System (INIS)

Hansson, B.

1998-01-01

The Owners of the Swedish Nuclear Power Plants have decided to form a document that should state the design principals and requirement for cost-effective and continuous development of the reactor safety in the future. The development of this document will be a part of the modernization and development of the Swedish Nuclear Power Plants. The basis for this document is an evaluation of Swedish and International standards and regulations as IAEA/INSAG, US-regulations, EUR etc. (author)
OSIRIS and SOMBRERO Inertial Fusion Power Plant Designs, Volume 2: Designs, Assessments, and Comparisons

Energy Technology Data Exchange (ETDEWEB)

Meier, W. R.; Bieri, R. L.; Monsler, M. J.; Hendricks, C. D.; Laybourne, P.; Shillito, K. R.

1992-03-01

This is a comprehensive design study of two Inertial Fusion Energy (IFE) electric power plants. Conceptual designs are presented for a fusion reactor (called Osiris) using an induction-linac heavy-ion beam driver, and another (called SOMBRERO) using a KrF laser driver. The designs covered all aspects of IFE power plants, including the chambers, heat transport and power conversion systems, balance-of-plant facilities, target fabrication, target injection and tracking, as well as the heavy-ion and KrF drivers. The point designs were assessed and compared in terms of their environmental & safety aspects, reliability and availability, economics, and technology development needs.
Systems required during and after an earthquake. Summary report. WWER-1000 nuclear power plants

International Nuclear Information System (INIS)

Monette, P.

1995-01-01

The scope of this document is to list the mechanical, instrumentation and electrical components required during and after earthquake, in order to achieve and maintain safe shutdown conditions of a WWER-1000 type nuclear power plant. The main objective pursued in establishing the systems and equipment list is to provide guidance for the design and implementation of the backfits which are necessary to increase seismic resistance of the components required after earthquake. The presented list is established on generic basis, i.e. it is applicable to any specific WWER-1000
Applying formal method to design of nuclear power plant embedded protection system

International Nuclear Information System (INIS)

Kim, Jin Hyun; Kim, Il Gon; Sung, Chang Hoon; Choi, Jin Young; Lee, Na Young

2001-01-01

Nuclear power embedded protection systems is a typical safety-critical system, which detects its failure and shutdowns its operation of nuclear reactor. These systems are very dangerous so that it absolutely requires safety and reliability. Therefore nuclear power embedded protection system should fulfill verification and validation completely from the design stage. To develop embedded system, various V and V method have been provided and especially its design using Formal Method is studied in other advanced country. In this paper, we introduce design method of nuclear power embedded protection systems using various Formal-Method in various respect following nuclear power plant software development guideline
Design methods for high temperature power plant structures

International Nuclear Information System (INIS)

Townley, C.H.A.

1984-01-01

The subject is discussed under the headings: introduction (scope of paper - reviews of design methods and design criteria currently in use for both nuclear and fossil fuelled power plant; examples chosen are (a) BS 1113, representative of design codes employed for power station boiler plant; (b) ASME Code Case N47, which is being developed for high temperature nuclear reactors, especially the liquid metal fast breeder reactor); design codes for power station boilers; Code Case N47 (design in the absence of thermal shock and thermal fatigue; design against cyclic loading at high temperature; further research in support of high temperature design methods and criteria for LMFBRs); concluding remarks. (U.K.)
Configuration management of plant modifications for nuclear power plants

International Nuclear Information System (INIS)

Ritsch, W.J.

1987-01-01

Due to the increasing complexity of nuclear power plant operation, regulatory pressure, and the large numbers of people required to operate and support the stations, the control of plant modifications at these plants needs to be expanded and improved. The aerospace and defense industries, as well as the owners or operators of large energy projects have established configuration management programs (CMPs) to control plant design changes. These programs are composed of well-defined functions for identifying, evaluating, recording, tracking, issuing, and documenting the established baseline conditions, as well as required changes to these baseline conditions. The purpose of this paper is to describe a recommended CMP for plant modifications consisting of a computerized data base installed on the utility's computer to provide a central storage of plant design and operations data necessary to control the following activities as they are affected by plant design changes: training; record management; operations; maintenance; health physics; planning/scheduling; procurement/inventory control; outage management (including modifications); and emergency response
Targeting a heterologous protein to multiple plant organelles via rationally designed 5? mRNA tags

NARCIS (Netherlands)

Voges, M.J.; Silver, P.A.; Way, J.C.; Mattozzi, M.D.

2013-01-01

Background Plant bioengineers require simple genetic devices for predictable localization of heterologous proteins to multiple subcellular compartments. Results We designed novel hybrid signal sequences for multiple-compartment localization and characterize their function when fused to GFP in

HTGR gas turbine power plant preliminary design

International Nuclear Information System (INIS)

Koutz, S.L.; Krase, J.M.; Meyer, L.

1973-01-01

The preliminary reference design of the HTGR gas turbine power plant is presented. Economic and practical problems and incentives related to the development and introduction of this type of power plant are evaluated. The plant features and major components are described, and a discussion of its performance, economics, development, safety, control, and maintenance is presented. 4 references
Rationalization of design and construction of buildings for nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Satoh, Shunsaku; Mitsumatsu, Kazuo

1987-02-01

This article presents various rationalization methods introduced in the past few years for design and construction of BWR nuclear power plant buildings. When the site for a nuclear power plant has been decided, investigation is made on various aspects of possible earthquakes, based on which anti-earthquake design for the plant site is established. The next step is to examine the displacements and stresses that may occur to various parts of the bulding from a postulated earthquake. This is normally called the earthquake response analysis and consists of calculating the behaviors of the buildings using large computers. A seismic controlled structure system has recently proposed, aiming to reduce the displacements and stresses of the building itself by controlling the flexibility of the installed seismic apparatus against the input of external loads. Lately, high strength concrete and high strength reinforcing steel bars (rebars) are being considered for practical application. If advanced computers and related accessories are utilized to the maximum, it will lead not only to efficiency in the design work but to the possibility of optimized design. For rational construction, a combined scaffolding and temporary support has been devised to reduce the time and volume of required temporary work. What have been developed for rationalization of construction work also include robots for heavy weight rebar fabrication, horizontal reed blind type rebars, portable concrete distributor, all weather environment facilities, and construction materials conveyance system. (Nogami, K.).
Rationalization of design and construction of buildings for nuclear power plants

International Nuclear Information System (INIS)

Satoh, Shunsaku; Mitsumatsu, Kazuo

1987-01-01

This article presents various rationalization methods introduced in the past few years for design and construction of BWR nuclear power plant buildings. When the site for a nuclear power plant has been decided, investigation is made on various aspects of possible earthquakes, based on which anti-earthquake design for the plant site is established. The next step is to examine the displacements and stresses that may occur to various parts of the bulding from a postulated earthquake. This is normally called the earthquake response analysis and consists of calculating the behaviors of the buildings using large computers. A seismic controlled structure system has recently proposed, aiming to reduce the displacements and stresses of the building itself by controlling the flexibility of the installed seismic apparatus against the input of external loads. Lately, high strength concrete and high strength reinforcing steel bars (rebars) are being considered for practical application. If advanced computers and related accessories are utilized to the maximum, it will lead not only to efficiency in the design work but to the possibility of optimized design. For rational construction, a combined scaffolding and temporary support has been devised to reduce the time and volume of required temporary work. What have been developed for rationalization of construction work also include robots for heavy weight rebar fabrication, horizontal reed blind type rebars, portable concrete distributor, all weather environment facilities, and construction materials conveyance system. (Nogami, K.)
Design Of Feedforward Controllers For Multivariable Plants

Science.gov (United States)

Seraji, Homayoun

1989-01-01

Controllers based on simple low-order transfer functions. Mathematical criteria derived for design of feedforward controllers for class of multiple-input/multiple-output linear plants. Represented by simple low-order transfer functions, obtained without reconstruction of states of commands and disturbances. Enables plant to track command while remaining unresponsive to disturbance in steady state. Feedback controller added independently to stabilize plant or to make control system less susceptible to variations in parameters of plant.
Overall plant design of PWRs

International Nuclear Information System (INIS)

Meyer, P.J.

1980-01-01

In the present lecture the main components and safety related systems are described, to get a general overview about the safety measures in a PWR. The idea to introduce safety systems is to protect the nuclear reactor core against the so-called design accidents and to prevent the release of activity to the environment. Furthermore the operation personnel has to be protected against radioactive contamination. All redundant and diversified safety measures used in a nuclear power station ensure reliable and safe operation of the plant in all modes of operation. To minimize the operational risk to an extended minimum besides active safety systems a lot of passive safety barriers are foreseen. With the design and construction, tests and quality assurance measures are performed to assure a safe plant operation. (orig.)
7 CFR 330.200 - Movement of plant pests regulated; permits required.

Science.gov (United States)

2010-01-01

... 7 Agriculture 5 2010-01-01 2010-01-01 false Movement of plant pests regulated; permits required... AND PLANT HEALTH INSPECTION SERVICE, DEPARTMENT OF AGRICULTURE FEDERAL PLANT PEST REGULATIONS; GENERAL; PLANT PESTS; SOIL, STONE, AND QUARRY PRODUCTS; GARBAGE Movement of Plant Pests § 330.200 Movement of...
Methodology for modular nuclear plant design and construction

International Nuclear Information System (INIS)

Lapp, C.W.; Golay, M.

1992-01-01

During the past decade, the rising cost of nuclear power plant construction has caused the cancellation of many projects and has forced some utilities into bankruptcy. Many factors have contributed to capital cost increases, including regulatory changes, the absence of standard designs, and low worker productivity. Low worker productivity can be attributed to the conventional building process, which is not conductive to productive labor. This study presents innovative ways to reduce the capital cost of nuclear plants through more efficient construction processes designed to increase worker productivity. A major portion of the plant capital cost is the interest paid during construction on borrowed capital. Modular fabrication could potentially reduce interest payments by compressing the construction schedule of nuclear facilities. Additional cost savings expected from modular designs arise from improved quality, productivity, and schedule control in fabrication of plant elements within a factory environment
SEPI an expert system for plant design

International Nuclear Information System (INIS)

Carotenuto, M.; Corleto, P.; Landeyro, P.

1988-01-01

The availability and suitability of technological information is of great importance in every kind of design task, especially when safety and reliability considerations are involved. In this paper an ''expert system for plant design'' (SEPI), is presented, together with its first application to nuclear back-end plants. This system is available on ENEA computer network. It is thought to be used both to collect know-how developed in the field and to assist unskilled designers during selection, evaluation and dimensioning tasks. It attemps to reproduce the normal way of ''reasoning'' and acting, and provides some graphic facilities
Design basis programs and improvements in plant operation

International Nuclear Information System (INIS)

Metcalf, M.F.

1991-01-01

Public Service Electric and Gas (PSE and G) Company operates three commercial nuclear power plants in southern New Jersey. The three plants are of different designs and vintages (two pressurized water reactors licensed in 1976 and 1980 and one boiling water reactor licensed in 1986). As the industry recognized the need to develop design basis programs, PSE and G also realized the need after a voluntary 52-day shutdown of one unit because of electrical design basis problems. In its drive to be a premier electric utility, PSE and G has been aggressively active in developing design basis documents (DBDs) with supporting projects and refined uses to obtain the expected value and see the return on investment. Progress on Salem is nearly 75% complete, while Hope Creek is 20% complete. To data, PSE and G has experienced success in the use of DBDs in areas such as development of plant modifications, development of the reliability-centered maintenance program, procedure upgrades, improved document retrieval, resolution of regulatory issues, and training. The paper examines the design basis development process, supporting projects, and expected improvements in plant operations as a result of these efforts
Licensing requirements for initial commissioning programs in Spain: Application to different PWR designs

International Nuclear Information System (INIS)

Munuera, A.; Conde, J.M.; Martinez, J.

1991-01-01

This paper describes the overall licensing process in Spain, focusing on the initial commissioning requirements. The significance of this part of the regulatory work is evident both from the licensing and the licensee's points of view. Licensing in Spain is ruled by different laws which determine the general requirements and fix the licensing frame. Being a nuclear technology importer country, the base of the regulatory work lies on the rules and regulations of the country of origin of the planet, with the addition of case specific requirements. The application of this methodology to plants designed in different countries produces licensing processes which are similar to the overall, but very different in its development. It also means a special technical effort on the part of the regulatory body to cope with the problems arising from the use of different technologies and safety standards. The start-up programs from fuel loading to full power of a Westinghouse plant (Vandellos 2) and a Siemens-KWU plant (Trillo 1) are compared from the technical point of view, enhancing the differences that can be relevant for the regulatory work. The difficulties arising from the application of both the German and US concepts are discussed. (orig.)
Plant modeling as a key tool for nuclear I and C design and V and V

International Nuclear Information System (INIS)

Krasnov, V.; Sokolov, O.; Symkin, B.

2006-01-01

This paper summarizes an intensive experience of LvivORGRES in the design and implementation of the digital control systems at VVER-1000 and VVER-440 nuclear power plants in Ukraine and Bulgaria. This experience is applicable to the digital I and C upgrade projects for other types of reactor equipment as well as to the design and testing of new I and C systems for new constructions. LvivORGRES was recently involved in several modernization projects as a functional designer and, also, provided technical support and supervision during the factory and site acceptance testing. It is widely accepted and proved by the industry's practice that a level and quality of system validation at all design and implementation phases are key to the successful future operation of I and C systems. The plant control systems have some additional validation requirements in comparing with the information and monitoring systems. According to the Ukrainian nuclear regulation standards, the scope of the control system projects should include the close loop stability analysis at all unit modes of operation. Besides the control system algorithms verification and validation, it was necessary to determine the tuning parameters for the system and use them initially during the system commissioning. LvivORGRES has developed the Adaptive Plant Modeling process that was used as a key tool in all design stages of control system upgrade projects: Software engineering tests, Integrated system validation tests, Factory acceptance tests. The Plant Model was developed on a modular basis which allowed the testing of all primary and secondary side regulators for all unit modes of operation including transients and unit start-up and shutdown. The Plant Model has been adapted to each project's requirements. The use of the plant simulation provided technical bases for important project decisions and documents including among others: system test strategy, initial tuning parameters, training plan, etc. The Plant
Requirements for Space Settlement Design

Science.gov (United States)

Gale, Anita E.; Edwards, Richard P.

2004-02-01

When large space settlements are finally built, inevitably the customers who pay for them will start the process by specifying requirements with a Request for Proposal (RFP). Although we are decades away from seeing the first of these documents, some of their contents can be anticipated now, and provide insight into the variety of elements that must be researched and developed before space settlements can happen. Space Settlement Design Competitions for High School students present design challenges in the form of RFPs, which predict basic requirements for space settlement attributes in the future, including structural features, infrastructure, living conveniences, computers, business areas, and safety. These requirements are generically summarized, and unique requirements are noted for specific space settlement locations and applications.
Safety of Nuclear Power Plants: Commissioning and Operation. Specific Safety Requirements

International Nuclear Information System (INIS)

2017-01-01

This publication is a revision of IAEA Safety Standards Series No. NS-R-2, Safety of Nuclear Power Plants: Operation, and has been extended to cover the commissioning stage. It describes the requirements to be met to ensure the safe commissioning, operation, and transition from operation to decommissioning of nuclear power plants. Over recent years there have been developments in areas such as long term operation of nuclear power plants, plant ageing, periodic safety review, probabilistic safety analysis review and risk informed decision making processes. It became necessary to revise the IAEA’s Safety Requirements in these areas and to correct and/or improve the publication on the basis of feedback from its application by both the IAEA and its Member States. In addition, the requirements are governed by, and must apply, the safety objective and safety principles that are established in the IAEA Safety Standards Series No. SF-1, Fundamental Safety Principles. A review of Safety Requirements publications, initiated in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan, revealed no significant areas of weakness but resulted in a small set of amendments to strengthen the requirements and facilitate their implementation. These are contained in the present publication.
Critique of Hanford Waste Vitrification Plant off-gas sampling requirements

International Nuclear Information System (INIS)

Goles, R.W.

1996-03-01

Off-gas sampling and monitoring activities needed to support operations safety, process control, waste form qualification, and environmental protection requirements of the Hanford Waste Vitrification Plant (HWVP) have been evaluated. The locations of necessary sampling sites have been identified on the basis of plant requirements, and the applicability of Defense Waste Processing Facility (DWPF) reference sampling equipment to these HWVP requirements has been assessed for all sampling sites. Equipment deficiencies, if present, have been described and the bases for modifications and/or alternative approaches have been developed
A basic design of alarm system for the future nuclear power plants in Korea

International Nuclear Information System (INIS)

Lee, Cheol-Kwon; Hur, Seop; Shin, Jae-Hwal; Koo, In-Soo; Park, Jong-Kyun

1997-01-01

The design of an advanced alarm system is under way to apply to the new MMIS for the future nuclear power plants in Korea. Based on the alarm system design bases we established the design requirements and are now refining them with the results of evaluation through the prototype. To realize the advanced system new algorithms for alarm processing and display are implemented and various new devices are examined. The evaluation for the design is performed in accordance with the verification and validation plans and through the prototype. (author). 7 refs, 2 figs
Lightning protection design of nuclear power plants. KTA safety code, version 6/99

International Nuclear Information System (INIS)

1999-06-01

This KTA safety code does not cover calculation methods for determination of lightning-induced voltage inputs to control room systems within the reactor building, as the literature presents a variety of applicable methods, which however cannot be directly applied to any power plant, due to the great diversity of geometries of the electrical systems and control room systems in nuclear power plants. Compliance with the design requirements of this safety code for shielding of buildings, and installation and shielding of cables, can be considered to offer the appropriate protection. (orig./CB) [de
Heat exchanger design for desalination plants

International Nuclear Information System (INIS)

1979-03-01

The Office of Saline Water (OSW) accomplished a very large amount of significant work related to the design and performance of large heat exchanger bundles and enhanced heat transfer surfaces. This work was undertaken to provide basic technical and economic data for the design of distillation plants for the desalination of seawater, and should be of value to other industrial applications as well. The OSW work covers almost every aspect of heat exchanger design, and ranges academic research to data gathering on commercial desalting plants. Exchanger design configurations include multistage flash plant condensers, vertical tube falling film and upflow evaporators, and horizontal tube spray film evaporators. Unfortunately, the data is scattered through a large number of reports of which many are concerned primarily with factors other than heat transfer, and the quality of reporting and the quality of the data are far from consistent. This report catalogues and organizes the heat exchanger data developed by the OSW. Some analysis as to the validity of the data is made and ranges of performance that can be expected are given. Emphasis is placed on the vertical tube, falling film evaporators. A thorough analysis of the large literature file that was surveyed was not possible. No analysis was made of the quality of original data, but apparent data discrepancies are pointed out where such discrepancies happen to be found
Implementing plant design changes economically

International Nuclear Information System (INIS)

Seamans, L.D.

1994-01-01

Consumers Power Company, as is the case with most utilities, is challenged by the necessity to control non-modification design changes. At the Consumers' Palisades Nuclear Plant, the challenge was met by the establishment of a project team that developed an innovative comprehensive design control process called Functional Equivalent Substitution (FES). This paper describes: (1) the project team composition; (2) the process development methodology; and (3) the implementation results
Advanced conceptual design report: T Plant secondary containment and leak detection upgrades. Project W-259

International Nuclear Information System (INIS)

Hookfin, J.D.

1995-01-01

The T Plant facilities in the 200-West Area of the Hanford site were constructed in the early 1940s to produce nuclear materials in support of national defense activities. T Plant includes the 271-T facility, the 221-T facility, and several support facilities (eg, 2706-T), utilities, and tanks/piping systems. T Plant has been recommended as the primary interim decontamination facility for the Hanford site. Project W-259 will provide capital upgrades to the T Plant facilities to comply with Federal and State of Washington environmental regulations for secondary containment and leak detection. This document provides an advanced conceptual design concept that complies with functional requirements for the T Plant Secondary Containment and Leak Detection upgrades
Management of design support for nuclear plant modifications

International Nuclear Information System (INIS)

Doyle, F.W.

1991-01-01

The paper will present an overview of the Ontario Hydro organization and processes for providing design support to the operating nuclear power plants. Examples of design support for Pickering GS will be highlighted. The process is described from identification of projects through the design, procurement, construction, commissioning and in-service phases. The practices for managing engineering deliverables are discussed in the context of how these integrate into the overall change control process. The interaction of Engineering with Operations, Construction, Supply and the regulatory bodies is discussed both for major retro-fit programs and for ongoing design support to the nuclear power plants. Recent experiences during the 1990 Pickering Station Outage and during the Unit 3 fuel channel replacement program are highlighted and an integrated 5 year plan for upgrading the safety related systems for the Pickering Nuclear Power Plant is presented. (author)

Developing a computer-based environment for the design of nuclear power plants: a perspective and philosophy

International Nuclear Information System (INIS)

Brey, H.; Kisner, R.A.

1985-08-01

This report surveys the usefulness and general design requirements for a large-scale database and database manager for design and analysis of nuclear power plant control systems. The control engineer's and systems integrator's need for timely, accurate, and searchable information for advanced control system design increases with competitive economics and stringent reliability demands. A philosophy is conveyed for the implementation of an integrated, comprehensive database system
Design of the vitrification plant for HLLW generated from the Tokai Reprocessing Plant

International Nuclear Information System (INIS)

Vematsu, K.

1986-01-01

Power Reactor and Nuclear Fuel Development Corporation (PNC) is now designing a vitrification plant. This plant is for the solidification of high-level liquid waste (HLLW) which is generated from the Tokai Reprocessing Plant, and for the demonstration of the vitrification technology. The detailed design of the plant which started in 1982 was completed in 1984. At present the design improvement is being made for the reduction of construction cost and for the licensing which is going to be applied in 1986. The construction will be started in autumn 1987. The plant has a large shielded cell with low flow ventilation, and employs rack-mounted module system and high performance two-armed servomanipulator system to accomplish the fully remote operations and maintenance. The vitrification of HLLW is based on the liquid-fed Joule-heated ceramic melter process. The processing capacity is equivalent to the reprocessing of 0.7 ton of heavy metals per day. The glass production rate is about 9 kg/h, and about 300 kg of glass is poured periodically from the bottom of the melter into a canister. Produced glass is stored under the forced air cooling condition
RO-75, Reverse Osmosis Plant Design Optimization and Cost Optimization

International Nuclear Information System (INIS)

Glueckstern, P.; Reed, S.A.; Wilson, J.V.

1999-01-01

1 - Description of problem or function: RO75 is a program for the optimization of the design and economics of one- or two-stage seawater reverse osmosis plants. 2 - Method of solution: RO75 evaluates the performance of the applied membrane module (productivity and salt rejection) at assumed operating conditions. These conditions include the site parameters - seawater salinity and temperature, the membrane module operating parameters - pressure and product recovery, and the membrane module predicted long-term performance parameters - lifetime and long flux decline. RO75 calculates the number of first and second stage (if applied) membrane modules needed to obtain the required product capacity and quality and evaluates the required pumping units and the power recovery turbine (if applied). 3 - Restrictions on the complexity of the problem: The program does not optimize or design the membrane properties and the internal structure and flow characteristics of the membrane modules; it assumes operating characteristics defined by the membrane manufacturers
IFE Power Plant design principles. Drivers. Solid state laser drivers

International Nuclear Information System (INIS)

Nakai, S.; Andre, M.; Krupke, W.F.; Mak, A.A.; Soures, J.M.; Yamanaka, M.

1995-01-01

The present status of solid state laser drivers for an inertial confinement thermonuclear fusion power plant is discussed. In particular, the feasibility of laser diode pumped solid state laser drivers from both the technical and economic points of view is briefly reviewed. Conceptual design studies showed that they can, in principle, satisfy the design requirements. However, development of new solid state materials with long fluorescence lifetimes and good thermal characteristics is a key issue for laser diode pumped solid state lasers. With the advent of laser diode pumping many materials which were abandoned in the past can presently be reconsidered as viable candidates. It is also concluded that it is important to examine the technical requirements for solid state lasers in relation to target performance criteria. The progress of laser diode pumped lasers in industrial applications should also be closely watched to provide additional information on the economic feasibility of this type of driver. 15 refs, 9 figs, 2 tabs
Preliminary design of the Carrisa Plains solar central receiver power plant. Volume II. Plant specifications

Energy Technology Data Exchange (ETDEWEB)

Price, R. E.

1983-12-31

The specifications and design criteria for all plant systems and subsystems used in developing the preliminary design of Carrisa Plains 30-MWe Solar Plant are contained in this volume. The specifications have been organized according to plant systems and levels. The levels are arranged in tiers. Starting at the top tier and proceeding down, the specification levels are the plant, system, subsystem, components, and fabrication. A tab number, listed in the index, has been assigned each document to facilitate document location.
Basic Study on Data-Centric design information integration system framework development for adapting Nuclear Power Plant construction in Korea

Energy Technology Data Exchange (ETDEWEB)

Lim, Byung Ki [KHNP, Gyeongju (Korea, Republic of)

2016-05-15

This study established the concept of data-centric design, which is the latest design technique, by analyzing the existing literature so that the data-centric design would be applied to the nuclear power plant projects in Korea and analyzed the status of data-centric design application by the advanced companies and the domestic design companies participating in the nuclear power plant projects. By analyzing the function of the 3D CAD commercial system and all design drawings used in the nuclear power plant projects in Korea, a data-centric design integrated system model has been developed. This study established the concept of data-centric design technology, analyzed the functions of the plant architect engineering (A/E) software being globally used in the plant field and the design process status of nuclear power plant projects in Korea. A design information integration system building model, which is capable of data-centric design, in the place of the existing document-centric system design such as P and ID and SLD, has been suggested through the investigation on the data-centric design cases of the advanced companies. The major functions of the suggested model required for the application to the domestic industry were drawn. The suggested framework builds the field design, which was performed in the 3D system of the constructor, as an owner's field design system, which can manage all design drawings generated from the field design and the related information in integrated way. An as-built full model integrated of plant architect engineering, supplier design and field design is built. It is handed over to the operation team at the O and M stage and utilized in the maintenance and repair. As a power plant full model of future construction project has been enabled, an improved design process has been suggested, in which only the design change information during the plant architect engineering (A/E) and the design change information during the field design
Preliminary design of the Carrisa Plains solar central receiver power plant. Volume I. Executive summary

Energy Technology Data Exchange (ETDEWEB)

1983-12-31

The design of the 30 MWe central receiver solar power plant to be located at Carrisa Plains, San Luis Obispo County, California, is summarized. The plant uses a vertical flat-panel (billboard) solar receiver located at the top of a tower to collect solar energy redirected by approximately 1900 heliostats located to the north of the tower. The solar energy is used to heat liquid sodium pumped from ground level from 610 to 1050/sup 0/F. The power conversion system is a non-reheat system, cost-effective at this size level, and designed for high-efficiency performance in an application requiring daily startup. Successful completion of this project will lead to power generation starting in 1986. This report also discusses plant performance, operations and maintenance, development, and facility cost estimate and economic analysis.
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
Design windows of laser fusion power plants and conceptual design of laser-diode pumped slab laser

International Nuclear Information System (INIS)

Kozaki, Y.; Eguchi, T.; Izawa, Y.

1999-01-01

An analysis of the design space available to laser fusion power plants has been carried out, in terms of design key parameters such as target gain, laser energy and laser repetition rate, the number of fusion react ion chambers, and plant size. The design windows of economically attractive laser fusion plants is identified with the constraints of key design parameters and the cost conditions. Especially, for achieving high repetition rate lasers, we have proposed and designed a diode-pumped solid-state laser driver which consists of water-cooled zig-zag path slab amplifiers. (author)
High degree utilization of computers for design of nuclear power plants

International Nuclear Information System (INIS)

Masui, Takao; Sawada, Takashi

1992-01-01

Nuclear power plants are the huge technology in which various technologies are compounded, and the high safety is demanded. Therefore, in the design of nuclear power plants, it is necessary to carry out the design by sufficiently grasping the behavior of the plants, and to confirm the safety by carrying out the accurate design evaluation supposing the various operational conditions, and as the indispensable tool for these analysis and evaluation, the most advanced computers in that age have been utilized. As to the utilization for the design, there are the fields of design, analysis and evaluation and another fields of the application to the support of design. Also in the field of the application to operation control, computers are utilized. The utilization of computers for the core design, hydrothermal design, core structure design, safety analysis and structural analysis of PWR plants, and for the nuclear design, safety analysis and heat flow analysis of FBR plants, the application to the support of design and the application to operation control are explained. (K.I.)
Implications of fusion power plant studies for materials requirements

International Nuclear Information System (INIS)

Cook, Ian; Ward, David; Dudarev, Sergei

2002-01-01

This paper addresses the key requirements for fusion materials, as these have emerged from studies of commercial fusion power plants. The objective of the international fusion programme is the creation of power stations that will have very attractive safety and environmental features and viable economics. Fusion power plant studies have shown that these objectives may be achieved without requiring extreme advances in materials. But it is required that existing candidate materials perform at least as well as envisaged in the environment of fusion neutrons, heat fluxes and particle fluxes. The development of advanced materials would bring further benefits. The work required entails the investigation of many intellectually exciting physics issues of great scientific interest, and of wider application than fusion. In addition to giving an overview, selected aspects of the science, of particular physics interest, are illustrated
Evaluation of earthquake resistance design for underground structures of nuclear power plant, (1)

International Nuclear Information System (INIS)

Tohma, Junichi; Kokusho, Kenji; Iwatate, Takahiro; Ohtomo, Keizo

1986-01-01

As to earthquake resistant design of underground civil engineering structures related with emergency cooling water system of nuclear power plant, it is required these structures must maintain the function of great important their own facilities during earthquakes, especially for design earthquake motion. In this study, shaft pipline, pit and duct for cooling sea water facilities were chosen as typical underground structures, and the authors deal with the seismic design method for calculation of the principal sectional force in these structures generated by design earthquake motion. Especially, comparative investigations concerned with response displacement method versus dynamic analysis methods (lumped mass analysis and finite element analysis) are discussed. (author)
Design and construction of nuclear power plants

International Nuclear Information System (INIS)

Meiswinkel, Ruediger; Meyer, Julian; Schnell, Juergen

2013-01-01

Despite all the efforts being put into expanding renewable energy sources, large-scale power stations will be essential as part of a reliable energy supply strategy for a longer period. Given that they are low on CO2 emissions, many countries are moving into or expanding nuclear energy to cover their baseload supply. Building structures required for nuclear installations whose protective function means they are classified as safety-related, have to meet particular construction requirements more stringent than those involved in conventional construction. This book gives a comprehensive overview from approval aspects given by nuclear and construction law, with special attention to the interface between plant and construction engineering, to a building structure classification. All life cycle phases are considered, with the primary focus on execution. Accidental actions on structures, the safety concept and design and fastening systems are exposed to a particular treatment. Selected chapters of the German concrete yearbook ''Beton-Kalender'' are now available in English. The new English BetonKalender Series delivers internationally useful engineering expertise and industrial know-how from Germany.
Component design considerations for gas turbine HTGR waste-heat power plant

International Nuclear Information System (INIS)

McDonald, C.F.; Vrable, D.L.

1976-01-01

Component design considerations are described for the ammonia waste-heat power conversion system of a large helium gas-turbine nuclear power plant under development by General Atomic Company. Initial component design work was done for a reference plant with a 3000-MW(t) High-Temperature Gas-Cooled Reactor (HTGR), and this is discussed. Advanced designs now being evaluated include higher core outlet temperature, higher peak system pressures, improved loop configurations, and twin 4000-MW(t) reactor units. Presented are the design considerations of the major components (turbine, condenser, heat input exchanger, and pump) for a supercritical ammonia Rankine waste heat power plant. The combined cycle (nuclear gas turbine and waste-heated plant) has a projected net plant efficiency of over 50 percent. While specifically directed towards a nuclear closed-cycle helium gas-turbine power plant (GT-HTGR), it is postulated that the bottoming waste-heat cycle component design considerations presented could apply to other low-grade-temperature power conversion systems such as geothermal plants
Analysis of the Possibility of Required Resources Estimation for Nuclear Power Plant Decommissioning Applying BIM

Energy Technology Data Exchange (ETDEWEB)

Jung, Insu [Korea Institute of construction Technology, Goyang (Korea, Republic of); Kim, Woojung [KHNP-Central Research Institute, Daejeon (Korea, Republic of)

2014-05-15

Estimation of decommissioning cost, decommissioning strategy, and decommissioning quantity at the time when entering into any decommissioning plans are some elements whose inputs are mandatory for nuclear power plant decommissioning. Ways to estimate decommissioning of required resources in the past have imposed great uncertainty since they analyze required resources at the construction stage, analyzing and consulting decommissioning required resources of overseas nuclear power plants. This study aims at analyzing whether required resources for decommissioning nuclear power plants can be estimated, applying BIM. To achieve this goal, this study analyzed the status quo of BIM such as definition, characteristics, and areas applied, and made use of them when drawing out study results by examining types and features of the tools realizing BIM. In order to review how BIM could be used for decommissioning nuclear power plants, the definition, characteristics and applied areas of BIM were discussed. BIM designs objects of the structures (walls, slabs, pillars, stairs, windows and doors, etc.) by 3D technology and endows attribute (function, structure and usage) information for each object, thereby providing visualized information of structures for participants in construction projects. Major characteristics of BIM attribute information are as follows: - Geometry: The information of objects is represented by measurable geometric information - Extensible object attributes: Objects include pre-defined attributes, and allow extension of other attributes. Any model that includes these attributes forms relationships with other various attributes in order to perform analysis and simulation. - All information including the attributes are integrated to ensure continuity, accuracy and accessibility, and all information used during the life cycle of structures are supported. This means that when information of required resources is added as another attributes other than geometric
Plant designer's view of the operator's role in nuclear plant safety

International Nuclear Information System (INIS)

Corcoran, W.R.; Church, J.F.; Cross, M.T.; Porter, N.J.

1981-01-01

The nuclear plant operator's role supports the design assumptions and equipment with four functional tasks. He must set up th plant for predictable response to disturbances, operate the plant so as to minimize the likelihood and severity of event initiators, assist in accomplishing the safety functions, and feed back operating experiences to reinforce or redefine the safety analyses' assumptions. The latter role enhances the operator effectiveness in the former three roles. The Safety Level Concept offers a different perspective that enables the operator to view his roles in nuclear plant safety. This paper outlines the operator's role in nuclear safety and classifies his tasks using the Safety Level Concept
10-MWe pilot-plant-receiver panel test requirements document solar thermal test facility

Energy Technology Data Exchange (ETDEWEB)

1978-08-25

Testing plans for a full-scale test receiver panel and supporting hardware which essentially duplicate both physically and functionally, the design planned for the Barstow Solar Pilot Plant are presented. Testing is to include operation during normal start and shutdown, intermittent cloud conditions, and emergencies to determine the panel's transient and steady state operating characteristics and performance under conditions equal to or exceeding those expected in the pilot plant. The effects of variations of input and output conditions on receiver operation are also to be investigated. Test hardware are described, including the pilot plant receiver, the test receiver assembly, receiver panel, flow control, electrical control and instrumentation, and structural assembly. Requirements for the Solar Thermal Test Facility for the tests are given. The safety of the system is briefly discussed, and procedures are described for assembly, installation, checkout, normal and abnormal operations, maintenance, removal and disposition. Also briefly discussed are quality assurance, contract responsibilities, and test documentation. (LEW)
Design Requirements, Epistemic Uncertainty and Solution Development Strategies in Software Design

DEFF Research Database (Denmark)

Ball, Linden J.; Onarheim, Balder; Christensen, Bo Thomas

2010-01-01

This paper investigates the potential involvement of “epistemic uncertainty” in mediating between complex design requirements and strategic switches in software design strategies. The analysis revealed that the designers produced an initial “first-pass” solution to the given design brief in a bre...... a view of software design as involving a mixed breadth-first and depth-first solution development approach, with strategic switching to depth-first design being triggered by requirement complexity and being mediated by associated feelings of uncertainty....
Risk Informed Assessment of Regulatory and Design Requirements for Future Nuclear Power Plants (Cooperative Agreement DE-FC03-99SF21902, Am. M004) Final Technical Report

International Nuclear Information System (INIS)

Ritterbusch, Stanley E.

2003-01-01

accidents would be an inherent part of the Probabilistic Safety Assessment for the plant and their evaluation would be probabilistic. Other first year accomplishments include (1) the conversion of an NRC database for cross-referencing NRC criteria and industry codes and standards to Microsoft 2000 software, (2) an assessment of the NRC's hearing process which concluded that the normal cross-examination during public hearings is not actually required by the U.S. Administrative Procedures Act, (3) the identification and listing of reliability data sources, and (4) interfacing with other industry groups (e.g., NEI and IAEA) and NRC at workshops for risk-informing regulations. The major accomplishments during the second year consisted of (1) issuance of the final report for Subtask 1.1, ''Identify Current Applicable Regulatory Requirements [and Industry Standards],'' (2) issuance of the final report for Subtask 1.2,'' Identify Structures, Systems, and Components and Their Associate d Costs for a Typical Plant,'' (3) extension of the new, highly risk-informed design and regulatory framework to non-light-water-reactor technology, (4) completion of more detailed thermal-hydraulic and probabilistic analyses of advanced conceptual reactor system/component designs, (6) initial evaluation and recommendations for improvement of the NRC design review process, and (7) initial development of the software format, procedures and statistical routines needed to store, analyze and retrieve the available reliability data. Final reports for Subtasks 1.1 (regulatory and design criteria) and 1.2 (costs for structures, systems, and components) were prepared and issued. A final report for Subtask 1.3 (Regulatory Framework) was drafted with the aim to issue it in Phase 3 (Year 3). One technical report was produced for Subtask 1.4 (methods development) and two technical reports were produced for Subtask 1.6 (sample problem analysis). An interim report on the NRC design review process (Subtask 1.7) was
Integrated design of SIGMA uranium enrichment plants

International Nuclear Information System (INIS)

Rivarola, Martin E.; Brasnarof, Daniel O.

1999-01-01

In the present work, we describe a preliminary analysis of the design feedbacks in a Uranium Enrichment Plant, using the SIGMA concept. Starting from the result of this analysis, a computer code has been generated, which allows finding the optimal configurations of plants, for a fixed production rate. The computer code developed includes the model of the Thermohydraulic loop of a SIGMA module. The model contains numerical calculations of the main components of the circuit. During the calculations, the main components are dimensioned, for a posterior cost compute. The program also makes an estimation of the enrichment gain of the porous membrane, for each separation stage. Once the dimensions of the main components are known, using the enrichment cascade calculation, the capital and operation costs of the plant could be determined. At this point it is simple to calculate a leveled cost of the Separative Work Unit (SWU). A numerical optimizer is also included in the program. This optimizer finds the optimal cascade configuration, for a given set of design parameters. The whole-integrated program permits to investigate in detail the feedback in the component design. Therefore, the sensibility of the more relevant parameters can be computed, with respect of the economical variables of the plant. (author)

External man-induced events in relation to nuclear power plant design

International Nuclear Information System (INIS)

1982-01-01

This Guide deals with the basic design requirements for nuclear power plants, and presents a general design approach for protection against the effects of man-induced events. Section 2 discusses the general design approach. Section 3 outlines the development of the basic information necessary for an evaluation of the adequacy of a design against the effects of aircraft crashes, fires, explosions, and the release of toxic gases or corrosive substances. Section 4 outlines the design logic for protection against external man-induced events. It indicates possible methods of ensuring overall plant safety, including protection against possible secondary effects. Included for each event are: a methodology for calculating the design input parameters from the data generated in the siting study, system protection considerations from the effects of this man-induced event, and criteria for judging the adequacy of the protection provided. Specific design guidance related to acts of sabotage is not provided in this Guide. It should be recognized, however, that for certain situations such acts can be important to safety and could constitute the controlling postulated initiating event for design. The list of events covered is not necessarily complete. However, important events on which enough work has already been done in various Member States to enable their effects to be converted into generally accepted design parameters are included. In addition, other man-induced events such as dam ruptures, ship collisions, construction accidents and the like are identified but no general guidelines for design can be specified for these at present. These events need to be considered on an ad hoc basis, in order to arrive at design input parameters for them
Test design requirements: Thermal conductivity probe testing

International Nuclear Information System (INIS)

Heath, R.E.

1985-01-01

This document establishes the test design requirements for development of a thermal conductivity probe test. The thermal conductivity probe determines in situ thermal conductivity using a line source transient heat conduction analysis. This document presents the rationale for thermal conductivity measurement using a thermal conductivity probe. A general test description is included. Support requirements along with design constraints are detailed to allow simple design of the thermal conductivity probe and test. The schedule and delivery requirements of the responsible test designer are also included. 7 refs., 1 fig
Design, Development and Evaluation of a Pneumatic Seeder for Automatic Planting of Seeds in Cellular Trays

Directory of Open Access Journals (Sweden)

E Movahedi

2014-04-01

Full Text Available For planting fine seeds in cellular trays, an automatic pneumatic seeder was designed, constructed and evaluated. CATIA software was used to design and analysis the system parts of the seeder. Different parts of the seeder, including vibrating seed hopper, vacuum boom, seed picking nozzles, seed tube, pneumatic system and electronic control unit for automation of the seeder, were designed and constructed. The area of nozzle orifice was used to calculate the required pressure of nozzle tip. The seeder was evaluated using two sizes of trays. Experiments were performed with five replications and the error of planting the seeds in the 105 and 390-cellular trays were 1.9 and 0.46 percent, respectively. The time of planting for 105 and 390 cellular trays reduced from 20 min (for manual seeding to 35 s and from 90 min to 160 s, respectively.
Techno-economic design optimization of solar thermal power plants

OpenAIRE

Morin, G.

2011-01-01

A holistic view is essential in the engineering of technical systems. This thesis presents an integrative approach for designing solar thermal power plants. The methodology is based on a techno-economic plant model and a powerful optimization algorithm. Typically, contemporary design methods treat technical and economic parameters and sub-systems separately, making it difficult or even impossible to realize the full optimization potential of power plant systems. The approach presented here ov...
Requirements and concepts for a nuclear plant surveillance and diagnostic system (NPSDS)

International Nuclear Information System (INIS)

Nicholson, P.J.; Lanning, D.D.

1980-01-01

An advanced plant surveillance and diagnostic system has been postulated for the purpose of aiding operator response to unanticipated plant upsets. The plant operator is described as an information processor that needs plant status information represented by symbolic outputs. These will be compatible with modern visual processing techniques, such as CRTs. Preferrred methods of estimating the state-of-the-plant and verifying measurements require on-line real-time models which are simple dynamical relationships based on energy and mass conservation laws. Implementation of on-line state estimation techniques using such models probably requires a distributed microcomputer system whose features are described. (auth)
Advanced light water reactor utility requirements document: Volume 1--ALWR policy and summary of top-tier requirements

International Nuclear Information System (INIS)

Anon.

1990-01-01

The U.S. utilities are leading an industry wide effort to establish the technical foundation for the design of the Advanced Light Water Reactor (ALWR). This effort, the ALWR Program, is being managed for the U.S. electric utility industry by the Electric Power Research Institute (EPRI) and includes participation and sponsorship of several international utility companies and close cooperation with the U.S. Department of Energy (DOE). The cornerstone of the ALWR Program is a set of utility design requirements which are contained in the ALWR Requirements Document. The purpose of the Requirement Document is to present a clear, complete statement of utility desires for their next generation of nuclear plants. The Requirements Document covers the entire plant up to the grid interface. It therefore is the basis for an integrated plant design, i.e., nuclear steam supply system and balance of plant, and it emphasizes those areas which are most important to the objective of achieving an ALWR which is excellent with respect to safety, performance, constructibility, and economics. The document applies to both Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs). The Requirements Document is organized in three volumes. Volume 1 summarizes AlWR Program policy statements and top-tier requirements. The top-tier design requirements are categorized by major functions, including safety and investment protection, performance, and design process and constructibility. There is also a set of general design requirements, such as simplification and proven technology, which apply broadly to the ALWR design, and a set of economic goals for the ALWR program. The top-tier design requirements are described further in Volume 1 and are formally invoked as requirements in Volumes 2 and 3
A design study for a fission product fixation plant and storage at Windscale

International Nuclear Information System (INIS)

Hill, K.M.; Ridley, G.; Adler, D.G.

1961-01-01

A flowsheet for a production plant has been worked out and the problems of designing a reliable production plant utilising it have been assessed. The properties of current glass compositions have been considered in relation to problems of permanent storage for periods of 800 to 1000 years and various storage proposals assessed. The performance of the preferred type of storage, i.e. dry storage in natural draught air-cooled vaults, has been estimated and consideration given to a number of variables affecting the optimum design. The design problems of a storage of the required life have been investigated in relation to suitable types of site and proposals made as to the types of construction to be used. Design work on tunnel loading facilities has been carried to the state of demonstrating feasibility. The capital costs and the economics of this type of waste storage have been investigated for unalloyed Magnox fuel and brief consideration given to alloyed and cermet fuels. Attention has been given to development problems of the whole system where they arise. (author)
Central receiver solar thermal power system, Phase 1. CDRL item 2. Pilot plant preliminary design report. Volume VI. Electrical power generation and master control subsystems and balance of plant

Energy Technology Data Exchange (ETDEWEB)

Hallet, Jr., R. W.; Gervais, R. L.

1977-10-01

The requirements, performance, and subsystem configuration for both the Commercial and Pilot Plant electrical power generation subsystems (EPGS) and balance of plants are presented. The EPGS for both the Commercial Plant and Pilot Plant make use of conventional, proven equipment consistent with good power plant design practices in order to minimize risk and maximize reliability. The basic EPGS cycle selected is a regenerative cycle that uses a single automatic admission, condensing, tandem-compound double-flow turbine. Specifications, performance data, drawings, and schematics are included. (WHK)
The impact of technical specification surveillance requirements and allowable outage times on plant availability

International Nuclear Information System (INIS)

Webster, S.A.; Finnicum, D.J.

1985-01-01

Surveillances required to be conducted by a plant's Technical Specifications have resulted in plant shutdowns and lost availability. This paper looks at shutdowns which have occurred due to required surveillance testing and insufficient repair time allowed by Technical Specifications. A loss of plant availability of almost 3% per plant year was found for U.S. pressurized water reactors during the five year period, 1979 to 1984. This figure excludes major problems which required plant shutdown whether or not mandated by the Technical Specifications. In addition to their affect on availability, such shutdowns can add to the challenges to plant safety systems and can affect plant aging by increasing the thermal cycles on plant components
Future needs for inelastic analysis in design of high-temperature nuclear plant components

International Nuclear Information System (INIS)

Corum, J.M.

1980-01-01

The role that inelastic analyses play in the design of high-temperature nuclear plant components is described. The design methodology, which explicitly accounts for nonlinear material deformation and time-dependent failure modes, requires a significant level of realism in the prediction of structural response. Thus, material deformation and failure modeling are, along with computational procedures, key parts of the methodology. Each of these is briefly discussed along with validation by comparisons with benchmark structural tests, and problem areas and needs are discussed for each
Thermal analysis, optimization and design of a Martian oxygen production plant

Science.gov (United States)

Iyer, Venkatesh A.; Sridhar, K. R.

1991-01-01

The objective is to optimally design the thermal components of a system that uses carbon dioxide (CO2) from the Martian atmosphere to produce oxygen (O2) for spacecraft propulsion and/or life-support. Carbon dioxide is thermally decomposed into carbon monoxide (CO) and O2 followed by the electrochemical separation of O2. The design of the overall system and its various individual components depends on, among other things, the fraction of the stoichiometric yield of O2 that can be realized in the system and the temperature of operation of the electrochemical separation membrane. The analysis indicates that a substantial reduction could be obtained in the mass and power requirements of the system if the unreacted CO2 were to be recycled. The concepts of an optimum temperature of the zirconia cell and impracticality of plant operation at low cell efficiencies are also discussed. The design of the thermal equipment is such that the mass and power requirements of the individual components and of the overall system are optimized.
Analysis and Design of the Logistics System for Rope Manufacturing Plant

Directory of Open Access Journals (Sweden)

Sun Xue

2017-01-01

Full Text Available In order to promote logistics system for manufacturing plant, this paper proposed a new design for the logistics system of a rope manufacturing plant. Through the analysis in the aspects of workshop facility layout, material handling and inventory management, the original logistics system of the plant is optimized. According to the comparison of the simulation results between original and optimized design, the optimized model has the higher productive efficiency. This can provide the references for the other manufacturing plant in analysis and design of the logistics system to improve plant efficiency.
Extreme external events in the design and assessment of nuclear power plants

International Nuclear Information System (INIS)

2003-03-01

The analysis of feedback experience from the operation of nuclear power plants (NPPs) in the past 20 years shows few cases of degradation of the plant safety initiated by external events. However, when these have occurred, the consequences have been serious, involving challenges to the defence in depth of the plant. Part of the problem involves the definition of the design basis parameters for some scenarios and differences among regulators on the methods for the protection of operational NPPs in relation to external events. This results in different engineering practices in Member States for the siting and design of NPPs. In the framework of the present revision of the IAEA safety standards on siting and design of NPPs, many initiatives have been implemented by the IAEA in recent years aimed at a systematic analysis of engineering practices in Member States. The most recent event in this connection was a Technical Committee Meeting (TCM) on Structural Safety of NPPs in Relation to Extreme External Loads, organized with the specific objective of evaluating the state of the art of NPP design in relation to external events. Such an analysis provided a technical background for the development of a common technical basis for an integrated approach in site evaluation, design and operation in relation to extreme external events. The scope included new and existing plants, as they are required to meet the same general safety principles, in spite of their peculiarities. The objective of this publication is to provide a technical background to drive regulators, plant owners and designers in the definition of a consistent strategy in selected safety issues on site evaluation, design and operation in relation to extreme external events. This publication is also of support to the IAEA in the development of safety standards since many Safety Guides dealing with related topics are under periodic review. Four major tasks were identified to comply with these general objectives
Forecasting manpower requirements for nuclear power plant construction

International Nuclear Information System (INIS)

Seltzer, N.; Schriver, W.R.

1978-01-01

This paper presents both the methodology and results of a segment of a comprehensive construction manpower demand forecasting system aimed at forecasting virtually all construction manpower requirements in the United States of America. The part of the system dealing with the demand for construction workers needed to build nuclear powered electricity generating plants is discussed here. The object of the system is to forecast manpower construction needs for each of 29 construction crafts on a monthly basis in each of 10 geographical regions of the United States. The method used is to establish profiles of the types of workers and time phasing required in the past. Profiling was done for different types of plants, different capacity classes, and different geographical locations. An appropriate worker profile matrix cannot simply be multiplied by the capacity of the proposed plant if the number of man-hours required per kilowatt of generating capacity is not constant. The value of this latter variable has changed considerably recently - presumably because of an increased awareness of environmental and safety considerations. Econometric techniques are used to forecast values for man-hours per kilowatt which are then multiplied by projected new capacity to be put in place. The resulting total man-hour requirement is then allocated over time and by craft through use of a worker profile matrix. The summary results indicate that 20 percent increases in man-hours required per kilowatt of capacity can be expected between 1977 and 1981. Total construction labour demand will rise from 65,700 work-years in 1977 to nearly 96,600 work-years in 1981. Forecasts of the actual number of different types of workers to be demanded in each month and in each region are available from the system. (author)
Design and fabrication of stainless steel components for long life of spent fuel reprocessing plants

International Nuclear Information System (INIS)

Natarajan, R.; Ramkumar, P.; Sundararaman, V.; Kamachi Mudali, U.; Baldev Raj; Shanmugam, K.

2010-01-01

Reprocessing of spent nuclear fuels based on the PUREX process is the proven process with many commercial plants operating satisfactorily worldwide. The process medium being nitric acid, austenitic stainless steel is the material of construction as it is the best commercially available material for meeting the conditions in the reprocessing plants. Because of the high radiation fields, contact maintenance of equipment and systems of these plants are very time consuming and costly unlike other chemical process plants. Though the plants constructed in the early years required extensive shut downs for replacement of equipment and systems within the first fifteen years of operation itself, development in the field of stainless steel metallurgy and fabrication techniques have made it possible to design the present day plants for an operating life period of forty years. A review of the operational experience of the PUREX process based aqueous reprocessing plants has been made in this paper and reveals that life limiting failures of equipment and systems are mainly due to corrosion while a few are due to stresses. Presently there are no standards for design specification of materials and fabrication of reprocessing plants like the nuclear power plants, where well laid down ASTM and ASME codes and standards are available which are based on the large scale operational feedbacks on pressure vessels for conventional and nuclear industries. (author)
Design of a uranium recovery pilot plant

International Nuclear Information System (INIS)

1984-01-01

The engineering design of a pilot plant of uranium recover, is presented. The diagrams and specifications of the equipments such as pipelines, pumps, values tanks, filters, engines, etc... as well as metallic structure and architetonic design is also presented. (author)
7 CFR 801.11 - Related design requirements.

Science.gov (United States)

2010-01-01

... 7 Agriculture 7 2010-01-01 2010-01-01 false Related design requirements. 801.11 Section 801.11... FOR GRAIN INSPECTION EQUIPMENT § 801.11 Related design requirements. (a) Suitability. The design... tolerances prescribed in §§ 801.3 through 801.10, be capable of repeating its results when the equipment is...
Improving human reliability through better nuclear power plant system design: Program for advanced nuclear power studies

International Nuclear Information System (INIS)

Golay, M.W.

1993-01-01

The project on ''Development of a Theory of the Dependence of Human Reliability upon System Designs as a Means of Improving Nuclear Power Plant Performance'' was been undertaken in order to address the problem of human error in advanced nuclear power plant designs. Lack of a mature theory has retarded progress in reducing likely frequencies of human errors. Work being pursued in this project is to perform a set of experiments involving human subjects who are required to operate, diagnose and respond to changes in computer-simulated systems, relevant to those encountered in nuclear power plants, which are made to differ in complexity in a systematic manner. The computer program used to present the problems to be solved also records the response of the operator as it unfolds
The plant design analyser and its applications

International Nuclear Information System (INIS)

Whitmarsh-Everiss, M.J.

1992-01-01

Consideration is given to the history of computational methods for the non-linear dynamic analysis of plant behaviour. This is traced from analogue to hybrid computers. When these were phased out simulation languages were used in the batch mode and the interactive computational capabilities were lost. These have subsequently been recovered using mainframe computing architecture in the context of small models using the Prototype Plant Design Analyser. Given the development of parallel processing architectures, the restriction on model size can be lifted. This capability and the use of advanced Work Stations and graphics software has enabled an advanced interactive design environment to be developed. This system is generic and can be used, with suitable graphics development, to study the dynamics and control behaviour of any plant or system for minimum cost. Examples of past and possible future uses are identified. (author)
Progress in XRCS-Survey plant instrumentation and control design for ITER

International Nuclear Information System (INIS)

Varshney, Sanjeev; Jha, Shivakant; Simrock, Stefan; Barnsley, Robin; Martin, Vincent; Mishra, Sapna; Patil, Prabhakant; Patel, Shreyas; Kumar, Vinay

2016-01-01

Highlights: • An identification of the major process functions system compliant to Plant Control Design Handbook (PCDH) has been made for XRCS-Survey plant I&C. • I&C Functional Breakdown Structure (FBS) and Operation Procedure (OP) have been drafted using Enterprise architect (EA). • I&C architecture, interface with ITER networks and Plants, configuration of cubicles are discussed towards nine design review deliverables. - Abstract: A real time, plasma impurity survey system based on X-ray Crystal Spectroscopy (XRCS) has been designed for ITER and will be made available in the set of first plasma diagnostics for measuring impurity ion concentrations and their in-flux. For the purpose of developing a component level design of XRCS-Survey plant I&C system that is compliant to the rules and guidelines defined in the Plant Control Design Handbook (PCDH), firstly an identification of the major process functions has been made. The preliminary plant I&C Functional Breakdown Structure (FBS) and Operation Procedure (OP) have been drafted using a system engineering tool, Enterprise Architect (EA). Conceptual I&C architecture, interface with the ITER networks and other Plants have been discussed along with the basic configuration of I&C cubicles aiming towards nine I&C deliverables for the design review.

Progress in XRCS-Survey plant instrumentation and control design for ITER

Energy Technology Data Exchange (ETDEWEB)

Varshney, Sanjeev, E-mail: sanjeev.varshney@iter-india.org [ITER-India, Institute for Plasma Research, Bhat, Gandhinagar, 382 428 (India); Jha, Shivakant [ITER-India, Institute for Plasma Research, Bhat, Gandhinagar, 382 428 (India); Simrock, Stefan; Barnsley, Robin; Martin, Vincent [ITER-Organization, Route de Vinon sur Verdon, CS 90 046, 13067 St. Paul-Lez-Durance, Cedex (France); Mishra, Sapna [ITER-India, Institute for Plasma Research, Bhat, Gandhinagar, 382 428 (India); Patil, Prabhakant [ITER-Organization, Route de Vinon sur Verdon, CS 90 046, 13067 St. Paul-Lez-Durance, Cedex (France); Patel, Shreyas; Kumar, Vinay [ITER-India, Institute for Plasma Research, Bhat, Gandhinagar, 382 428 (India)

2016-11-15

Highlights: • An identification of the major process functions system compliant to Plant Control Design Handbook (PCDH) has been made for XRCS-Survey plant I&C. • I&C Functional Breakdown Structure (FBS) and Operation Procedure (OP) have been drafted using Enterprise architect (EA). • I&C architecture, interface with ITER networks and Plants, configuration of cubicles are discussed towards nine design review deliverables. - Abstract: A real time, plasma impurity survey system based on X-ray Crystal Spectroscopy (XRCS) has been designed for ITER and will be made available in the set of first plasma diagnostics for measuring impurity ion concentrations and their in-flux. For the purpose of developing a component level design of XRCS-Survey plant I&C system that is compliant to the rules and guidelines defined in the Plant Control Design Handbook (PCDH), firstly an identification of the major process functions has been made. The preliminary plant I&C Functional Breakdown Structure (FBS) and Operation Procedure (OP) have been drafted using a system engineering tool, Enterprise Architect (EA). Conceptual I&C architecture, interface with the ITER networks and other Plants have been discussed along with the basic configuration of I&C cubicles aiming towards nine I&C deliverables for the design review.
Human factors design, verification, and validation for two types of control room upgrades at a nuclear power plant

Energy Technology Data Exchange (ETDEWEB)

Boring, Laurids Ronald [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2014-10-01

This paper describes the NUREG-0711 based human factors engineering (HFE) phases and associated elements required to support design, verification and validation (V&V), and implementation of a new plant process computer (PPC) and turbine control system (TCS) at a representative nuclear power plant. This paper reviews ways to take a human-system interface (HSI) specification and use it when migrating legacy PPC displays or designing displays with new functionality. These displays undergo iterative usability testing during the design phase and then undergo an integrated system validation (ISV) in a full scope control room training simulator. Following the successful demonstration of operator performance with the systems during the ISV, the new system is implemented at the plant, first in the training simulator and then in the main control room.
Human factors design, verification, and validation for two types of control room upgrades at a nuclear power plant

International Nuclear Information System (INIS)

Boring, Laurids Ronald

2014-01-01

This paper describes the NUREG-0711 based human factors engineering (HFE) phases and associated elements required to support design, verification and validation (V&V), and implementation of a new plant process computer (PPC) and turbine control system (TCS) at a representative nuclear power plant. This paper reviews ways to take a human-system interface (HSI) specification and use it when migrating legacy PPC displays or designing displays with new functionality. These displays undergo iterative usability testing during the design phase and then undergo an integrated system validation (ISV) in a full scope control room training simulator. Following the successful demonstration of operator performance with the systems during the ISV, the new system is implemented at the plant, first in the training simulator and then in the main control room.
Evaluation of the energy required for constructing and operating a fusion power plant

International Nuclear Information System (INIS)

Buende, R.

1982-09-01

The energy required for constructing and operating a tokamak fusion power plant is appraised with respect to the energy output during the lifetime of the plant. A harvesting factor is deduced as a relevant figure of energetic merit and is used for a comparison between fusion, fission, and coal-fired power plants. Because fusion power plants involve considerable uncertainties the comparison is supplemented by a sensitivity analysis. In comparison with Light Water Reactor plants fusion power plants appear to be rather favourable in this respect. The energy required for providing the fuel is relatively low for fusion plants, thus overcompensating the considerable higher amount of energy necessary for constructing the fusion power plant. (orig.)
Effect of increased regulation on capital costs and manual labor requirements of nuclear power plants

International Nuclear Information System (INIS)

Paik, S.; Schriver, W.R.

1981-01-01

An attempt is made to explain the impact of increasing governmental regulation on capital costs and labor requirements for constructing light water reactor (LWR) electric power plants. The principal factors contributing to these increases are: (1) market conditions and (2) increased regulation. General market conditions include additional costs attributable to price inflation of equipment, material, labor, and the increased cost of money. The central objective of this work is to estimate the impact of increasing regulation on plant costs and, conversely, on output. To do this it is necessary to isolate two opposing sets of forces which have been in operation during the period of major regulatory expansion: learning based upon plant design experience and economies of scale with increasing size (generating capacity) of newer plants. Conceptual models are specified to capture the independent effects of increasing regulation, learning, and economies of scale. Empirical results were obtained by estimating the models on data collected from industry experience during the 1967-1980 period. 23 refs
Design and Tuning of Wind Power Plant Voltage Controller with Embedded Application of Wind Turbines and STATCOMs

DEFF Research Database (Denmark)

Petersen, Lennart; Kryezi, Fitim; Iov, Florin

2017-01-01

This study addresses a detailed design and tuning of a wind power plant voltage control with reactive power contribution of wind turbines and static synchronous compensators (STATCOMs). First, small-signal models of a single wind turbine and STATCOM are derived by using the state-space approach....... A complete phasor model of the entire wind power plant is constructed, being appropriate for voltage control assessment. An exemplary wind power plant located in the United Kingdom and the corresponding grid code requirements are used as a base case. The final design and tuning process of the voltage...... controller results in a guidance, proposed for this particular control architecture. It provides qualitative outcomes regarding the parametrisation of each individual control loop and how to adjust the voltage controller depending on different grid stiffnesses of the wind power plant connection...
Planting and care of fine hardwood seedlings: Designing hardwood tree plantings for wildlife

Science.gov (United States)

Brian J. MacGowan

2003-01-01

Woody plants can be of value to many wildlife species. The species of tree or shrub, or the location, size, and shape of planting can all have an impact on wildlife. The purpose of this paper is to discuss the benefits of trees and shrubs for wildlife and how to design tree and shrub plantings for wildlife. Some of the practices may conflict with other management...
Cold vacuum drying facility design requirements

Energy Technology Data Exchange (ETDEWEB)

Irwin, J.J.

1997-09-24

This release of the Design Requirements Document is a complete restructuring and rewrite to the document previously prepared and released for project W-441 to record the design basis for the design of the Cold Vacuum Drying Facility.
Cold vacuum drying facility design requirements

International Nuclear Information System (INIS)

Irwin, J.J.

1997-01-01

This release of the Design Requirements Document is a complete restructuring and rewrite to the document previously prepared and released for project W-441 to record the design basis for the design of the Cold Vacuum Drying Facility
Design of Plant Gas Exchange Experiments in a Variable Pressure Growth Chamber

Science.gov (United States)

Corey, Kenneth A.

1996-01-01

Sustainable human presence in extreme environments such as lunar and martian bases will require bioregenerative components to human life support systems where plants are used for generation of oxygen, food, and water. Reduced atmospheric pressures will be used to minimize mass and engineering requirements. Few studies have assessed the metabolic and developmental responses of plants to reduced pressure and varied oxygen atmospheres. The first tests of hypobaric pressures on plant gas exchange and biomass production at the Johnson Space Center will be initiated in January 1996 in the Variable Pressure Growth Chamber (VPGC), a large, closed plant growth chamber rated for 10.2 psi. Experiments were designed and protocols detailed for two complete growouts each of lettuce and wheat to generate a general database for human life support requirements and to answer questions about plant growth processes in reduced pressure and varied oxygen environments. The central objective of crop growth studies in the VPGC is to determine the influence of reduced pressure and reduced oxygen on the rates of photosynthesis, dark respiration, evapotranspiration and biomass production of lettuce and wheat. Due to the constraint of one experimental unit, internal controls, called pressure transients, will be used to evaluate rates of CO2 uptake, O2 evolution, and H2O generation. Pressure transients will give interpretive power to the results of repeated growouts at both reduced and ambient pressures. Other experiments involve the generation of response functions to partial pressures of O2 and CO2 and to light intensity. Protocol for determining and calculating rates of gas exchange have been detailed. In order to build these databases and implement the necessary treatment combinations in short time periods, specific requirements for gas injections and removals have been defined. A set of system capability checks will include determination of leakage rates conducted prior to the actual crop
Design requirement on KALIMER blanket fuel assembly duct

International Nuclear Information System (INIS)

Hwang, Woan; Kang, H. Y.; Nam, C.; Kim, J. O.

1998-03-01

This document describes design requirements which are needed for designing the blanket fuel assembly duct of the KALIMER as design guidance. The blanket fuel assembly duct of the KALIMER consists of fuel rods, mounting rail, nosepiece, duct with pad, handling socket with pad. Blanket fuel rod consists of top end plug, bottom end plug with solid ferritic-martensitic steel rod and key way blanket fuel slug, cladding, and wire wrap. In the assembly, the rods are in a triangular pitch array, and the rod bundle is attached to the nosepiece with mounting rails. The bottom end of the assembly duct is formed by a long nosepiece which provides the lower restraint function and the paths for coolant inlet. This report contains functional requirements, performance and operational requirements, interfacing systems requirements, core restraint and interface requirements, design limits and strength requirements, system configuration and essential feature requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements. (author). 20 refs., 4 figs
Next Generation Nuclear Plant Pre-Conceptual Design Report

International Nuclear Information System (INIS)

Larry Demick; Doug Vandel

2007-01-01

management of the project using commercial processes, share the cost of design and construction with the government, and secure a commercial nuclear operating company to operate the plant. In FY-07, Pre-Conceptual Design (PCD) work was completed by the NGNP Project at the INL with the objective of developing a framework in which the design and technology development of the NGNP could progress and to begin to develop bases for selection of the specific design and operational characteristics of NGNP. This work was completed by three contractor teams with extensive experience in HTGR technology, nuclear power applications, and hydrogen production. The teams were led by Westinghouse Electric Company, LLC; AREVA NP, Inc.; and General Atomics. The scope of work included completion of special studies to address key aspects of the NGNP (e.g., reactor type, power levels, power conversion system [PCS] and heat transfer/transport system [HTS] designs, licensing and end product disposition). The results of these special studies were applied to the development by each contractor of a recommended design for NGNP and a commercial version of the HTGR. These were then used to estimate costs and schedule for design; construction; licensing; startup and testing; operation; and deactivation, decontamination, and decommissioning (DD and D) of the NGNP and an economic assessment for an Nth of a kind (NOAK) commercial plant. A primary objective of this work is to identify research and development (R and D), data needs, and future studies required to support selection of key characteristics of and to support the design and licensing processes for the NGNP. The Preliminary Project Management Plan (PPMP) for the NGNP identifies two planning options that weigh a range of programmatic risks and approaches to mitigating risk. The two options are compliant with the EPAct, but emphasize different approaches to technology development risks, design and construction risks, and the extent of demonstration in
ITER plant systems

International Nuclear Information System (INIS)

Kolbasov, B.; Barnes, C.; Blevins, J.

1991-01-01

As part of a series of documents published by the IAEA that summarize the results of the Conceptual Design Activities for the ITER project, this publication describes the conceptual design of the ITER plant systems, in particular (i) the heat transport system, (ii) the electrical distribution system, (iii) the requirements for radioactive equipment handling, the hot cell, and waste management, (iv) the supply system for fluids and operational chemicals, (v) the qualitative analyses of failure scenarios and methods of burn stability control and emergency shutdown control, (vi) analyses of tokamak building functions and design requirements, (vii) a plant layout, and (viii) site requirements. Refs, figs and tabs
Nuclear power plant design characteristics. Structure of nuclear power plant design characteristics in the IAEA Power Reactor Information System (PRIS)

International Nuclear Information System (INIS)

2007-03-01

One of the IAEA's priorities has been to maintain the Power Reactor Information System (PRIS) database as a viable and useful source of information on nuclear reactors worldwide. To satisfy the needs of PRIS users as much as possible, the PRIS database has included also a set of nuclear power plant (NPP) design characteristics. Accordingly, the PRIS Technical Meeting, organized in Vienna 4-7 October 2004, initiated a thorough revision of the design data area of the PRIS database to establish the actual status of the data and make improvements. The revision first concentrated on a detailed review of the design data completion and the composition of the design characteristics. Based on the results of the review, a modified set and structure of the unit design characteristics for the PRIS database has been developed. The main objective of the development has been to cover all significant plant systems adequately and provide an even more comprehensive overview of NPP unit designs stored in the PRIS database
Inferences from new plant design from fast flux test facility operation

International Nuclear Information System (INIS)

Peterson, R.E.; Peckinpaugh, C.L.; Simpson, D.E.

1985-04-01

Experience gained through operation of the Fast Flux Test Facility (FFTF) is now sufficiently extensive that this experience can be utilized in designing the next generation of liquid metal fast reactors. Experience with FFTF core and plant components is cited which can result in design improvements to achieve inherently safe, economic reactor plants. Of particular interest is the mixed oxide fuel system which has demonstrated large design margins. Other plant components have also demonstrated high reliability and offer capital cost reduction opportunities through design simplifications. The FFTF continues to be a valuable US resource which affords prototypic development and demonstration, contributing to public acceptability of future plants
Shield requirement estimation for pin storage room in fuel fabrication plant

International Nuclear Information System (INIS)

Shanthi, M.M.; Keshavamurthy, R.S.; Sivashankaran, G.

2012-01-01

Fast Reactor Fuel Cycle Facility (FRFCF) is an upcoming project in Kalpakkam. It has the facility to recycle the fuel from PFBR. It is an integrated facility, consists of fuel reprocessing plant, fuel fabrication plant (FFP), core subassembly plant, Reprocessed Uranium plant (RUP) and waste management plant. The spent fuel from PFBR would be reprocessed in fuel reprocessing plant. The reprocessed fuel material would be sent to fuel fabrication plant. The main activity of fuel fabrication plant is the production of MOX fuel pins. The fuel fabrication plant has a fuel pin storage room. The shield requirement for the pin storage room has been estimated by Monte Carlo method. (author)
Design of a Load Following Controller for APR+ Nuclear Plants

International Nuclear Information System (INIS)

Lee, Sim Won; Kim, Jae Hwan; Kim, Dong Su; Na, Man Gyun; Yu, Keuk Jong

2011-01-01

The load-following operation of an APR+ nuclear plants is required to restrain the adjustment of boric acid concentration and to efficiently control the control rods for the flexibility of the operation. Especially, axial flux distribution disproportion that is usually caused by load-following operation in a reactor core induces xenon oscillation because the absorption cross-section of xenon is extremely large and its effects in a reactor are delayed by the iodine precursor. Rapid and smooth power maneuvering has its benefits in view of the economical and safe operation of reactors, so it is required that the controller is efficiently designed. Therefore, the load-following operation of an APR+ nuclear plants needs the ultimate automatic control and the advanced control method that satisfies the conditions such as the flexibility, safety and convenience. A model predictive control (MPC) method is applied to design an automatic load-following controller for the integrated thermal power level and axial shape index (ASI) control for an APR+ nuclear plants. Some tracking controllers use only the current tracking command. On the other hand, since MPC considers future commands in addition to the current tracking command, the MPC can achieve better tracking performance. Therefore, the MPC has been applied very much to the control of industrial process systems. The basic concept of the MPC is to solve an optimization problem for generating finite future control inputs at current time and to implement as the current control input only the first control input among the solutions of the finite time steps. At the next time step, the second control input is not implemented and the procedure to solve the optimization problem is then repeated. The power level and the ASI are controlled by the regulating control banks and part-strength control banks together with the automatic adjustment of boric acid concentration. The 3-dimensional MASTER code, which models the APR+ nuclear
Design of a Load Following Controller for APR+ Nuclear Plants

Energy Technology Data Exchange (ETDEWEB)

Lee, Sim Won; Kim, Jae Hwan; Kim, Dong Su; Na, Man Gyun [Chosun University, Gwnagju (Korea, Republic of); Yu, Keuk Jong [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

2011-08-15

The load-following operation of an APR+ nuclear plants is required to restrain the adjustment of boric acid concentration and to efficiently control the control rods for the flexibility of the operation. Especially, axial flux distribution disproportion that is usually caused by load-following operation in a reactor core induces xenon oscillation because the absorption cross-section of xenon is extremely large and its effects in a reactor are delayed by the iodine precursor. Rapid and smooth power maneuvering has its benefits in view of the economical and safe operation of reactors, so it is required that the controller is efficiently designed. Therefore, the load-following operation of an APR+ nuclear plants needs the ultimate automatic control and the advanced control method that satisfies the conditions such as the flexibility, safety and convenience. A model predictive control (MPC) method is applied to design an automatic load-following controller for the integrated thermal power level and axial shape index (ASI) control for an APR+ nuclear plants. Some tracking controllers use only the current tracking command. On the other hand, since MPC considers future commands in addition to the current tracking command, the MPC can achieve better tracking performance. Therefore, the MPC has been applied very much to the control of industrial process systems. The basic concept of the MPC is to solve an optimization problem for generating finite future control inputs at current time and to implement as the current control input only the first control input among the solutions of the finite time steps. At the next time step, the second control input is not implemented and the procedure to solve the optimization problem is then repeated. The power level and the ASI are controlled by the regulating control banks and part-strength control banks together with the automatic adjustment of boric acid concentration. The 3-dimensional MASTER code, which models the APR+ nuclear
Cold vacuum drying facility design requirements

Energy Technology Data Exchange (ETDEWEB)

IRWIN, J.J.

1999-07-01

This document provides the detailed design requirements for the Spent Nuclear Fuel Project Cold Vacuum Drying Facility. Process, safety, and quality assurance requirements and interfaces are specified.
Cold vacuum drying facility design requirements

International Nuclear Information System (INIS)

IRWIN, J.J.

1999-01-01

This document provides the detailed design requirements for the Spent Nuclear Fuel Project Cold Vacuum Drying Facility. Process, safety, and quality assurance requirements and interfaces are specified

Study on in-service inspection and repair program and related plant design for Japan Sodium-Cooled Fast Reactor (JSFR)

International Nuclear Information System (INIS)

Kubo, Shigenobu; Suzuki, Shinichi; Kotake, Shoji; Nishiyama, Noboru; Uzawa, Masayuki

2011-01-01

Maintenance and repair program and conformity with them were investigated as a part of the conceptual design study of Japan Sodium-cooled Fast Reactor (JSFR). The maintenance program was set by taking the feature of sodium-cooled reactors and domestic practice of LWRs into account. Both of regulatory required inspection and voluntary inspection, which are conducted in the domestic LWRs, were counted. The regulatory required ISI program was based on that of the previous Japanese SFRs, LWRs (JSME S NA1) and liquid metal cooled reactors (ASME section XI division 3). Parts to be inspected, methods of inspection were identified for major structures and components. Concerning the repair program, we set three levels of repair requirements based on estimated frequency of defect and failure during the plant life time. For level 1, which might be occur several times during the plant life time, it is required to be easily repaired in a short period. Access routes and working space are considered in the component design and its arrangement. For level 2, which might be unlikely to occur during the plant life time, it is required to check that the repair work is feasible in a practical time range. For level 3, which frequency is negligible small, repair is not taken into account but the feasibility was investigated. The plant design shall be done so that all of above mentioned inspection and repair can be conducted. It is desired to ensure accessibility for all of the coolant and cover gas boundaries and the internal structures in order to cope with unforeseen troubles. Access routes for the reactor vessel and its internal structures, piping, pumps and intermediate heat exchangers and steam generators were investigated. As the results of that, possible ways for implementation of the maintenance and repair were identified. (author)
Requirements for class 1, 2, and 3 pressure-retaining systems and components in CANDU nuclear power plants

International Nuclear Information System (INIS)

1991-09-01

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

DEFF Research Database (Denmark)

Li, Xuemeng; Zhang, Zhinan; Ahmed-Kristensen, Saeema

2014-01-01

to be defined in a new context. This paper focuses on understanding the design requirement sources at the requirement elicitation phase. It aims at proposing an improved design requirement source classification considering emerging markets and presenting current methods for eliciting requirement for each source...
Design of nuclear power plants

International Nuclear Information System (INIS)

Lobo, C.G.

1987-01-01

The criteria of design and safety, applied internationally to systems and components of PWR type reactors, are described. The main criteria of the design analysed are: thermohydraulic optimization; optimized arrangement of buildings and components; low costs of energy generation; high level of standardization; application of specific safety criteria for nuclear power plants. The safety criteria aim to: assure the safe reactor shutdown; remove the residual heat and; avoid the release of radioactive elements for environment. Some exemples of safety criteria are given for Angra-2 and Angra-3 reactors. (M.C.K.) [pt
Safety of Nuclear Power Plants: Commissioning and Operation. Specific Safety Requirements (French Edition)

International Nuclear Information System (INIS)

2016-01-01

This publication describes the requirements to be met to ensure the safe operation of nuclear power plants. It takes into account developments in areas such as long term operation of nuclear power plants, plant ageing, periodic safety review, probabilistic safety analysis and risk informed decision making processes. In addition, the requirements are governed by, and must apply, the safety objective and safety principles that are established in the IAEA Safety Standards Series No. SF-1, Fundamental Safety Principles. A review of Safety Requirements publications was commenced in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan. The review revealed no significant areas of weakness and resulted in just a small set of amendments to strengthen the requirements and facilitate their implementation, which are contained in the present publication.
Top-tier requirements for KNGR

International Nuclear Information System (INIS)

Sung-Jae, Ch.; Kwangho, L.; Dong Wook, J.

1996-01-01

In 1992, Korea Electric Power Corporation (KEPCO) has launched the next generation reactor project to develop the standard design of an advanced pressurized water reactor by 2000. This advanced reactor aims to have the sufficient capability to be a safe, environmentally sound and economical energy source for 2000's in Korea. In conjunction with the project development, the program phase I is studied and it is in the Korean Next Generation Reactor (KNGR) first phase project that the requirements of this specification called ''Top-tier'' have been established. These functional requirements are of the first importance for the design, construction and operation of a nuclear power plant. These requirements are divided into safety requirements, serious accidents control, design base requirements, definition of the system characteristics, performance, construction feasibility, economical objectives, site parameters and design processes. The ''Top-tier'' requirements are concentrated on the improvement of the safety and reliability. Safety is one of the first priorities. In particular, the requirements for the design of the next reactors generation must include the capacity to control serious accidents because when an accident occurs, the protection degree is crucial. The KNGR requirements include the existing nuclear power plants competitiveness as well as those of the coal thermal plants. Moreover, when safety is reinforced, the economic competitiveness can be assured. At the present time, a subsequent specification for the KNGR considering the bases of the domestic technology and experimenting the running. (O.M.)
Seismic design criteria used for electrical raceway systems in commercial nuclear power plants

International Nuclear Information System (INIS)

Summers, P.B.; Manrique, M.A.; Nelson, T.A.

1991-01-01

This paper summarizes some of the seismic design approaches, relevant technical issues and criteria used over the years for design of electrical raceway systems at commercial nuclear power plant facilities. The approaches used for design and endorsed by the NRC can be seen to be quite varied. In recent years, considerably more rigor has been required for raceway design, as well as for the level of design basis documentation produced. However, there has also been a willingness by the NRC to accept rational approaches based on testing, analytical results or experience data, provided proper justification is given. Such rational approaches can simplify the significant task of analysis, design and construction of miles of raceways and thousands of raceway supports. Summarizing past practice and identifying relevant technical issues are an important first step in formalizing up-to-date criteria for new raceway designs
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.
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
Understand the Design Requirement in Companies

DEFF Research Database (Denmark)

Li, Xuemeng; Ahmed-Kristensen, Saeema

2015-01-01

requirements can lead to inappropriate products (Hall, et al., 2002). Understanding the nature of design requirements and the sources, from where they can or should be generated, is critical to before developing methods and processes to support this process. Requirement Engineering research, originated from...
Knowledge-based support system for requirement elaboration in design

International Nuclear Information System (INIS)

Furuta, Kazuo; Kondo, Shunsuke

1994-01-01

Design requirements are the seeds of every design activity, but elicitation and formalization of them are not easy tasks. This paper proposes a method to support designers in such requirement elaboration process with a computer. In this method the cognitive work space of designers is modeled by abstraction and structural hierarchies, and supporting functions of knowledge-based requirement elaboration, requirement classification and assessment of contentment status of requirements are provided on this framework. A prototype system was developed and tested using fast breeder reactor design. (author)
Three-dimensional computer aided design system for plant layout

International Nuclear Information System (INIS)

Yoshinaga, Toshiaki; Kiguchi, Takashi; Tokumasu, Shinji; Kumamoto, Kenjiro.

1986-01-01

The CAD system for three-dimensional plant layout planning, with which the layout of pipings, cable trays, air conditioning ducts and so on in nuclear power plants can be planned and designed effectively in a short period is reported. This system comprises the automatic routing system by storing the rich experience and know-how of designers in a computer as the knowledge, and deciding the layout automatically following the predetermined sequence by using these, the interactive layout system for reviewing the routing results from higher level and modifying to the optimum layout, the layout evaluation system for synthetically evaluating the layout from the viewpoint of the operability such as checkup and maintenance, and the data base system which enables these effective planning and design. In this report, the total constitution of this system and the technical features and effects of the individual subsystems are outlined. In this CAD system for three-dimensional plant layout planning, knowledge engineering, CAD/CAM, computer graphics and other latest technology were introduced, accordingly by applying this system to plant design, the design can be performed quickly, various case studies can be carried out at planning stage, and systematic and optimum layout planning becomes possible. (Kako, I.)
FutureGen 2.0 Oxy-Coal Combustion Carbon Capture Plant Pre-FEED Design and Cost

Energy Technology Data Exchange (ETDEWEB)

Flanigan, Tom; Pybus, Craig; Roy, Sonya; Lockwood, Frederick; McDonald, Denny; Maclnnis, Jim

2011-09-30

This report summarizes the results of the Pre-Front End Engineering Design (pre-FEED) phase of a proposed advanced oxy-combustion power generation plant to repower the existing 200 MWe Unit 4 at Ameren Energy Resources’ (AER) Meredosia Power Plant. AER has formed an alliance with Air Liquide Process and Construction, Inc. (ALPC) and Babcock & Wilcox Power Generation Group (B&W PGG) for the design, construction, and testing of the facility, and has contracted with URS Corporation (URS) for preliminary design and Owner’s engineering services. The Project employs oxy-combustion technology – combustion of coal with nearly pure oxygen and recycled flue gas (instead of air) – to capture approximately 90% of the flue gas CO2 for transport and sequestration by another Project. Plant capacity and configuration has been developed based on the B&W PGG-ALPC cool recycle process firing high-sulfur bituminous coal fuel, assuming baseload plant operation to maximize existing steam turbine capability, with limited consideration for plant redundancy and performance optimization in order to keep plant costs as low as practical. Activities and preliminary results from the pre-FEED phase addressed in this report include the following: Overall plant thermal performance; Equipment sizing and system configuration; Plant operation and control philosophy; Plant emissions and effluents; CO₂ production and recovery characteristics; Project cost estimate and economic evaluation; Integrated project engineering and construction schedule; Project risk and opportunity assessment; Development of Project permitting strategy and requirements During the Phase 2 of the Project, additional design details will be developed and the Phase 1 work products updated to support actual construction and operation of the facility in Phase 3. Additional information will be provided early in Phase 2 to support Ameren-Environmental in finalizing the appropriate permitting strategies and permit
Baseload Nitrate Salt Central Receiver Power Plant Design Final Report

Energy Technology Data Exchange (ETDEWEB)

Tilley, Drake [Abengoa Solar LLC, Lakewood, CO (United States); Kelly, Bruce [Abengoa Solar LLC, Lakewood, CO (United States); Burkholder, Frank [Abengoa Solar LLC, Lakewood, CO (United States)

2014-12-12

The objectives of the work were to demonstrate that a 100 MWe central receiver plant, using nitrate salt as the receiver coolant, thermal storage medium, and heat transport fluid in the steam generator, can 1) operate, at full load, for 6,400 hours each year using only solar energy, and 2) satisfy the DOE levelized energy cost goal of $0.09/kWhe (real 2009 $). To achieve these objectives the work incorporated a large range of tasks relating to many different aspects of a molten salt tower plant. The first Phase of the project focused on developing a baseline design for a Molten Salt Tower and validating areas for improvement. Tasks included a market study, receiver design, heat exchanger design, preliminary heliostat design, solar field optimization, baseline system design including PFDs and P&IDs and detailed cost estimate. The baseline plant met the initial goal of less than $0.14/kWhe, and reinforced the need to reduce costs in several key areas to reach the overall $0.09/kWhe goal. The major improvements identified from Phase I were: 1) higher temperature salt to improve cycle efficiency and reduce storage requirements, 2) an improved receiver coating to increase the efficiency of the receiver, 3) a large receiver design to maximize storage and meet the baseload hours objective, and 4) lower cost heliostat field. The second Phase of the project looked at advancing the baseline tower with the identified improvements and included key prototypes. To validate increasing the standard solar salt temperature to 600 °C a dynamic test was conducted at Sandia. The results ultimately proved the hypothesis incorrect and showed high oxide production and corrosion rates. The results lead to further testing of systems to mitigate the oxide production to be able to increase the salt temperature for a commercial plant. Foster Wheeler worked on the receiver design in both Phase I and Phase II looking at both design and lowering costs utilizing commercial fossil boiler
Report of the US Nuclear Regulatory Commission Piping Review Committee. Volume 2. Evaluation of seismic designs: a review of seismic design requirements for Nuclear Power Plant Piping

Energy Technology Data Exchange (ETDEWEB)

1985-04-01

This document reports the position and recommendations of the NRC Piping Review Committee, Task Group on Seismic Design. The Task Group considered overlapping conservation in the various steps of seismic design, the effects of using two levels of earthquake as a design criterion, and current industry practices. Issues such as damping values, spectra modification, multiple response spectra methods, nozzle and support design, design margins, inelastic piping response, and the use of snubbers are addressed. Effects of current regulatory requirements for piping design are evaluated, and recommendations for immediate licensing action, changes in existing requirements, and research programs are presented. Additional background information and suggestions given by consultants are also presented.
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)
Designing fault-tolerant real-time computer systems with diversified bus architecture for nuclear power plants

International Nuclear Information System (INIS)

Behera, Rajendra Prasad; Murali, N.; Satya Murty, S.A.V.

2014-01-01

Fault-tolerant real-time computer (FT-RTC) systems are widely used to perform safe operation of nuclear power plants (NPP) and safe shutdown in the event of any untoward situation. Design requirements for such systems need high reliability, availability, computational ability for measurement via sensors, control action via actuators, data communication and human interface via keyboard or display. All these attributes of FT-RTC systems are required to be implemented using best known methods such as redundant system design using diversified bus architecture to avoid common cause failure, fail-safe design to avoid unsafe failure and diagnostic features to validate system operation. In this context, the system designer must select efficient as well as highly reliable diversified bus architecture in order to realize fault-tolerant system design. This paper presents a comparative study between CompactPCI bus and Versa Module Eurocard (VME) bus architecture for designing FT-RTC systems with switch over logic system (SOLS) for NPP. (author)
Design and Evaluation of a Secure Virtual Power Plant.

Energy Technology Data Exchange (ETDEWEB)

Johnson, Jay Tillay [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2017-09-01

For three years, Sandia National Laboratories, Georgia Institute of Technology, and University of Illinois at Urbana-Champaign investigated a smart grid vision in which renewable-centric Virtual Power Plants (VPPs) provided ancillary services with interoperable distributed energy resources (DER). This team researched, designed, built, and evaluated real-time VPP designs incorporating DER forecasting, stochastic optimization, controls, and cyber security to construct a system capable of delivering reliable ancillary services, which have been traditionally provided by large power plants or other dedicated equipment. VPPs have become possible through an evolving landscape of state and national interconnection standards, which now require DER to include grid-support functionality and communications capabilities. This makes it possible for third party aggregators to provide a range of critical grid services such as voltage regulation, frequency regulation, and contingency reserves to grid operators. This paradigm (a) enables renewable energy, demand response, and energy storage to participate in grid operations and provide grid services, (b) improves grid reliability by providing additional operating reserves for utilities, independent system operators (ISOs), and regional transmission organization (RTOs), and (c) removes renewable energy high-penetration barriers by providing services with photovoltaics and wind resources that traditionally were the jobs of thermal generators. Therefore, it is believed VPP deployment will have far-reaching positive consequences for grid operations and may provide a robust pathway to high penetrations of renewables on US power systems. In this report, we design VPPs to provide a range of grid-support services and demonstrate one VPP which simultaneously provides bulk-system energy and ancillary reserves.
Optimizing NSSS power and turbine/generator performance for standardized nuclear power plant designs in tropical climates

International Nuclear Information System (INIS)

Parece, M.V.; Stack, T.G.; Huffman, A.D.

2007-01-01

The EPR was developed by AREVA as a standardized nuclear power plant design that could be deployed throughout the world. The first EPR is currently being constructed at Olkiluoto, Finland. Many of the plant systems for this first-of-a-kind unit are optimized for the climate and heat rejection method (once-through cooling) used at Olkiluoto. Two such systems are the Nuclear Steam Supply System (NSSS) and the Turbine/Generator (T/G) system. To achieve the EPR's target net electrical output for tropical climates and various condenser heat rejection methods, design studies were performed that showed that the NSSS and T/G system designs developed for the Olkiluoto site conditions required modification. The business case for EPR on U.S. sites where average ambient temperature is above 60 F, implies an economical design that provides an average net electrical output of at least 1600 MWe. It has been shown through parametric studies that the key features of the design needed to achieve this goal are: -) rated core thermal power of 4590 MWth, which is supported by plant systems, structures and components; -) the use of mechanical draft cooling towers rather than natural draft cooling towers; -) a low pressure turbine design with reduced exhaust annulus area; and -) a multi-pressure condenser configuration
Feasibility design study. Land-based OTEC plants. Final report

Energy Technology Data Exchange (ETDEWEB)

Brewer, J. H.; Minor, J.; Jacobs, R.

1979-01-01

The purpose of this study has been to determine the feasibility of installing 10 MWe (MegaWatt-electric) and 40 MWe land-based OTEC demonstration power plants at two specific sites: Keahole Point on the western shore of the island of Hawaii; and Punta Tuna, on the southeast coast of the main island of Puerto Rico. In addition, the study has included development of design parameters, schedules and budgets for the design, construction and operation of these plants. Seawater systems (intake and discharge pipes) were to be sized so that flow losses were equivalent to those expected with a platform-based OTEC power plant. The power module (components and general arrangement was established based on the TRW design. Results are presented in detail. (WHK)

Design of 90-mgd wastewater reclamation plant

International Nuclear Information System (INIS)

Cain, C.B.; Kluesener, J.W.; Lazarus, E.

1981-01-01

Ninety MGD of municipal secondary effluent is to be reclaimed for use as cooling water in the Palo Verde Nuclear Generating Station in Arizona. The water reclamation plant design criteria are presented for nitrification, lime-soda softening, filtration, and sludge processing. In-pipe storage is used for flow equalization. Forced air down-draft trickling filters with computer-controlled recycle are used for nitrification. Effluent recirculation is used to maintain constant sludge blanket dept in the upflow solids contact clarifier softening units. An influent flow-splitting, gravity-backwash, dual-media filter is used for final suspended solids removal. Rheological data was used in the hydraulic design of sludge piping. Computerized automatic flushing also aids in preventing sludge pumping problems. Reclamation plant design and construction are integrated in one Engineering-Procurement-Construction project conducted by a single company. 5 refs
Task oriented design of robot kinematics using grid method and its application to nuclear power plant

International Nuclear Information System (INIS)

Chang, Pyung-Hun; Park, Joon-Young

2002-01-01

This paper presents a Task Oriented Design method for robot kinematics based on grid method, widely used in finite difference method and heat transfer/fluid flow analyses. This approach drastically reduces complexities and computational burden due to previous approaches. More specifically, the grid method with a new formulation simplifies the design to a problem of three-design-variable unit grid, which does not require to solve inverse/forward kinematics. The effectiveness of the grid method has been confirmed through a kinematics design of a robot for nuclear power plants. (author)
Thermal insulation system design and fabrication specification (nuclear) for the Clinch River Breeder Reactor plant

International Nuclear Information System (INIS)

1978-01-01

This specification defines the design, analysis, fabrication, testing, shipping, and quality requirements of the Insulation System for the Clinch River Breeder Reactor Plant (CRBRP), near Oak Ridge, Tennessee. The Insulation System includes all supports, convection barriers, jacketing, insulation, penetrations, fasteners, or other insulation support material or devices required to insulate the piping and equipment cryogenic and other special applications excluded. Site storage, handling and installation of the Insulation System are under the cognizance of the Purchaser
Design aspects of radiation protection for nuclear power plants

International Nuclear Information System (INIS)

1985-01-01

This Safety Guide deals with the provisions to be made in the design of thermal neutron reactor power plants to protect site personnel and the public from undue exposure to ionizing radiation during operational states and accident conditions. The effective radiation protection is a combination of good design, high quality construction and proper operation. The document gives guidance on how to satisfy the objectives contained in Subsection 2.2 and Section 9 of the Code of Practice on Design for Safety of Nuclear Power Plants
A Course on Operational Considerations in Wastewater Treatment Plant Design. Student Manual.

Science.gov (United States)

Stottler, Stag and Associates, San Antonio, TX.

This manual was designed to furnish information for upgrading the design of wastewater treatment plant facilities and to serve as a resource for establishing criteria for upgrading these plants. The manual also furnishes information for modifying plant design to compensate for current organic and hydraulic overloads and/or to meet more stringent…
Nuclear challenges and progress in designing stellarator power plants

International Nuclear Information System (INIS)

El-Guebaly, L.

2007-01-01

, such as recycling (within the nuclear industry) and clearance (or unconditional release to the commercial market). The ARIES-CS bioshield, cryostat, and individual magnet constituents qualify for clearance, representing ∼80% of the total waste volume. We developed a recycling approach for the non-clearable, in-vessel components using a combination of conventional and advanced remote handling equipment that can handle high doses of 3000 Sv/h or more. Several additional nuclear-related tasks received considerable attention during the ARIES-CS design process. These include the radial build definition, the well-optimized in-vessel components that satisfy the top-level requirements, the carefully selected nuclear and engineering parameters to produce an economic optimum, and the overarching safety constraints to deliver a safe and reliable power plant. This paper provides a brief historical overview of the progress in designing stellarator power plants and a perspective to the successful integration of the nuclear activity into the final ARIES-CS design
Design and analysis of heat recovery system in bioprocess plant

International Nuclear Information System (INIS)

Anastasovski, Aleksandar; Rašković, Predrag; Guzović, Zvonimir

2015-01-01

Highlights: • Heat integration of a bioprocess plant is studied. • Bioprocess plant produces yeast and ethyl-alcohol. • The design of a heat recovery system is performed by batch pinch analysis. • Direct and indirect heat integration approaches are used in process design. • The heat recovery system without a heat storage opportunity is more profitable. - Abstract: The paper deals with the heat integration of a bioprocess plant which produces yeast and ethyl-alcohol. The referent plant is considered to be a multiproduct batch plant which operates in a semi-continuous mode. The design of a heat recovery system is performed by batch pinch analysis and by the use of the Time slice model. The results obtained by direct and indirect heat integration approaches are presented in the form of cost-optimal heat exchanger networks and evaluated by different thermodynamic and economic indicators. They signify that the heat recovery system without a heat storage opportunity can be considered to be a more profitable solution for the energy efficiency increase in a plant
Extraction of uranium from seawater: chemical process and plant design feasibility study

International Nuclear Information System (INIS)

Campbell, M.H.; Frame, J.M.; Dudey, N.D.; Kiel, G.R.; Mesec, V.; Woodfield, F.W.; Binney, S.E.; Jante, M.R.; Anderson, R.C.; Clark, G.T.

1979-02-01

A major assessment was made of the uranium resources in seawater. Several concepts for moving seawater to recover the uranium were investigated, including pumping the seawater and using natural ocean currents or tides directly. The optimal site chosen was on the southeastern Puerto Rico coast, with the south U.S. Atlantic coast as an alternate. The various processes for extracting uranium from seawater were reviewed, with the adsorption process being the most promising at the present time. Of the possible adsorbents, hydrous titanium oxide was found to have the best properties. A uranium extraction plant was conceptually designed. Of the possible methods for contacting the seawater with the adsorbent, a continuous fluidized bed concept was chosen as most practical for a pumped system. A plant recovering 500 tonnes of U 3 O 8 per year requires 5900 cubic meters per second of seawater to be pumped through the adsorbent beds for a 70% overall recovery efficiency. Total cost of the plant was estimated to be about $6.2 billion. A computer model for the process was used for parametric sensitivity studies and economic projections. Several design case variations were developed. Other topics addressed were the impact of co-product recovery, environmental considerations, etc
Extraction of uranium from seawater: chemical process and plant design feasibility study

Energy Technology Data Exchange (ETDEWEB)

Campbell, M.H.; Frame, J.M.; Dudey, N.D.; Kiel, G.R.; Mesec, V.; Woodfield, F.W.; Binney, S.E.; Jante, M.R.; Anderson, R.C.; Clark, G.T.

1979-02-01

A major assessment was made of the uranium resources in seawater. Several concepts for moving seawater to recover the uranium were investigated, including pumping the seawater and using natural ocean currents or tides directly. The optimal site chosen was on the southeastern Puerto Rico coast, with the south U.S. Atlantic coast as an alternate. The various processes for extracting uranium from seawater were reviewed, with the adsorption process being the most promising at the present time. Of the possible adsorbents, hydrous titanium oxide was found to have the best properties. A uranium extraction plant was conceptually designed. Of the possible methods for contacting the seawater with the adsorbent, a continuous fluidized bed concept was chosen as most practical for a pumped system. A plant recovering 500 tonnes of U/sub 3/O/sub 8/ per year requires 5900 cubic meters per second of seawater to be pumped through the adsorbent beds for a 70% overall recovery efficiency. Total cost of the plant was estimated to be about $6.2 billion. A computer model for the process was used for parametric sensitivity studies and economic projections. Several design case variations were developed. Other topics addressed were the impact of co-product recovery, environmental considerations, etc.
Requirements' Role in Mobilizing and Enabling Design Conversation

Science.gov (United States)

Bergman, Mark

Requirements play a critical role in a design conversation of systems and products. Product and system design exists at the crossroads of problems, solutions and requirements. Requirements contextualize problems and solutions, pointing the way to feasible outcomes. These are captured with models and detailed specifications. Still, stakeholders need to be able to understand one-another using shared design representations in order to mobilize bias and transform knowledge towards legitimized, desired results. Many modern modeling languages, including UML, as well as detailed, logic-based specifications are beyond the comprehension of key stakeholders. Hence, they inhibit, rather than promote design conversation. Improved design boundary objects (DBO), especially design requirements boundary objects (DRBO), need to be created and refined to improve the communications between principals. Four key features of design boundary objects that improve and promote design conversation are discussed in detail. A systems analysis and design case study is presented which demonstrates these features in action. It describes how a small team of analysts worked with key stakeholders to mobilize and guide a complex system design discussion towards an unexpected, yet desired outcome within a short time frame.
Characterisation of Liquefaction Effects for Beyond-Design Basis Safety Assessment of Nuclear Power Plants

Science.gov (United States)

Bán, Zoltán; Győri, Erzsébet; János Katona, Tamás; Tóth, László

2015-04-01

Preparedness of nuclear power plants to beyond design base external effects became high importance after 11th of March 2011 Great Tohoku Earthquakes. In case of some nuclear power plants constructed at the soft soil sites, liquefaction should be considered as a beyond design basis hazard. The consequences of liquefaction have to be analysed with the aim of definition of post-event plant condition, identification of plant vulnerabilities and planning the necessary measures for accident management. In the paper, the methodology of the analysis of liquefaction effects for nuclear power plants is outlined. The case of Nuclear Power Plant at Paks, Hungary is used as an example for demonstration of practical importance of the presented results and considerations. Contrary to the design, conservatism of the methodology for the evaluation of beyond design basis liquefaction effects for an operating plant has to be limited to a reasonable level. Consequently, applicability of all existing methods has to be considered for the best estimation. The adequacy and conclusiveness of the results is mainly limited by the epistemic uncertainty of the methods used for liquefaction hazard definition and definition of engineering parameters characterizing the consequences of liquefaction. The methods have to comply with controversial requirements. They have to be consistent and widely accepted and used in the practice. They have to be based on the comprehensive database. They have to provide basis for the evaluation of dominating engineering parameters that control the post-liquefaction response of the plant structures. Experience of Kashiwazaki-Kariwa plant hit by Niigata-ken Chuetsu-oki earthquake of 16 July 2007 and analysis of site conditions and plant layout at Paks plant have shown that the differential settlement is found to be the dominating effect in case considered. They have to be based on the probabilistic seismic hazard assessment and allow the integration into logic
Human factors engineering in Clinch River Breeder plant design

International Nuclear Information System (INIS)

Planchon, H.P. Jr.; Kaushal, N.N.; Snider, J.

1982-01-01

The Clinch River Breeder Reactor Plant (CRBRP) Project formed a Control Room Task Force to ensure that lessons learned from the Three Mile Island accident are incorporated into the design. The charter for the Control Room Task Force was to review plant operations from the control room. The focus was on the man-machine interface to ensure that the systems' designs and operator actions meshed to properly support plant operation during normal and off-normal conditions. Specific items included for review are described. This paper describes the methodology utilized to accomplish the Task Forces' objectives and the results of the review
Effects of foundation conditions on plant design

International Nuclear Information System (INIS)

Ehasz, J.L.

1975-01-01

Design considerations for nuclear plant foundations are examined including site stability, bearing capacity and settlement, dynamic response, and structural modeling techniques for dynamic analysis. (U.S.)
Trends in radiation protection: possible effects on fusion power plant design

International Nuclear Information System (INIS)

Eurajoki, Tapani; Frias, Manuel Pascual; Orlandi, Sergio

2003-01-01

Since the design of fusion power plants involves long-term issues, ranging over several decades, it is useful to try to foresee under what kind of regulations the first fusion plants are to be operated. Application of present radiological regulations and practice to a fusion power plant concept is considered. The current design phase of fusion power plants motivates the top-down dose assessment, but it is crucial to aim at bottom-up assessments to ensure radiation doses as low as reasonably achievable. Since several issues, relating both to our knowledge on radiation as well as to the practice of radiation protection, may change in the future, it is necessary to continuously follow the development in the further design of fusion power plants
Feasibility study on floating nuclear power plant (1). Conceptual design study of FNPP. Contract research

International Nuclear Information System (INIS)

Yabuuchi, Noriaki; Shimazaki, Junya; Ochiai, Masaaki; Takahashi, Masao; Nakazawa, Toshio

2001-02-01

Offshore siting methods for a nuclear power plant are classified into three types as a floating type, a settled type and a land reclamation type. The floating nuclear power plant (FNPP) has a number of advantages, such as seismic isolation, standardization of design and manufacturing, and reduction of construction period. It is, however, required for FNPP to establish the safety standards, which are different from ones for land based nuclear power plant. Investigations for this subject have not been conducted sufficiently. In this report, design study on a concept for FNPP and a review on stability evaluation for the floating platform, which were performed in order to study the safety concept of the FNPP are described. The basic concept of the FNPP are described. The basic concept for FNPP is as follows: The FNPP is sited approximately 1 - 2km off the sea coast on the open sea with water depth of about 20m and it is moored of protected sea by the breakwater, it provide a floating platform for a 1,100MWe class PWR plant. The results of design study show that the floating platform for 1,100MWe class PWR plant of 300m (L) x 80m (W) x 35m (H), and displacement of approximately 300,000 ton can be constructed in a dockyard. This floating platform guarded by the breakwater is found to be stable enough to install the nuclear power plant from the analysis simulating the movement of the platform due to sea wave or wind. (author)
Feasibility study on floating nuclear power plant (1). Conceptual design study of FNPP. Construct research

Energy Technology Data Exchange (ETDEWEB)

Yabuuchi, Noriaki; Shimazaki, Junya; Ochiai, Masaaki [Department of Nuclear Energy System, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan); Takahashi, Masao [Niigata Engineering Co. Ltd., Tokyo (Japan); Nakazawa, Toshio [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Sato, Kazuo [Ishikawajima-Harima Heavy Industries Co. Ltd., Tokyo (Japan)

2001-02-01

Offshore siting methods for a nuclear power plant are classified into three types as a floating type, a settled type and a land reclamation type. The floating nuclear power plant (FNPP) has a number of advantages, such as seismic isolation, standardization of design and manufacturing, and reduction of construction period. It is, however, required for FNPP to establish the safety standards, which are different from ones for land based nuclear power plant. Investigations for this subject have not been conducted sufficiently. In this report, design study on a concept for FNPP and a review on stability evaluation for the floating platform, which were performed in order to study the safety concept of the FNPP are described. The basic concept of the FNPP are described. The basic concept for FNPP is as follows: The FNPP is sited approximately 1 - 2km off the sea coast on the open sea with water depth of about 20m and it is moored of protected sea by the breakwater, it provide a floating platform for a 1,100MWe class PWR plant. The results of design study show that the floating platform for 1,100MWe class PWR plant of 300m (L) x 80m (W) x 35m (H), and displacement of approximately 300,000 ton can be constructed in a dockyard. This floating platform guarded by the breakwater is found to be stable enough to install the nuclear power plant from the analysis simulating the movement of the platform due to sea wave or wind. (author)
Risk-based design of process plants with regard to domino effects and land use planning

Energy Technology Data Exchange (ETDEWEB)

Khakzad, Nima, E-mail: nkhakzad@gmail.com [Safety and Security Science Group (S3G), Faculty of Technology, Policy and Management, TU Delft, Delft (Netherlands); Reniers, Genserik [Safety and Security Science Group (S3G), Faculty of Technology, Policy and Management, TU Delft, Delft (Netherlands); Antwerp Research Group on Safety and Security (ARGoSS), Faculty of Applied Economics, Universiteit Antwerpen, Antwerp (Belgium); Research Group CEDON, Campus Brussels, KULeuven, Brussels (Belgium)

2015-12-15

Highlights: • A Bayesian network methodology has been developed to estimate the total probability of major accidents in chemical plants. • Total probability of accidents includes the probability of individual accidents and potential domino effects. • The methodology has been extended to calculate on-site and off-site risks. • The results of the risk analysis have been used in a multi-criteria decision analysis technique to risk-based design of chemical plants. - Abstract: Land use planning (LUP) as an effective and crucial safety measure has widely been employed by safety experts and decision makers to mitigate off-site risks posed by major accidents. Accordingly, the concept of LUP in chemical plants has traditionally been considered from two perspectives: (i) land developments around existing chemical plants considering potential off-site risks posed by major accidents and (ii) development of existing chemical plants considering nearby land developments and the level of additional off-site risks the land developments would be exposed to. However, the attempts made to design chemical plants with regard to LUP requirements have been few, most of which have neglected the role of domino effects in risk analysis of major accidents. To overcome the limitations of previous work, first, we developed a Bayesian network methodology to calculate both on-site and off-site risks of major accidents while taking domino effects into account. Second, we combined the results of risk analysis with Analytic Hierarchical Process to design an optimal layout for which the levels of on-site and off-site risks would be minimum.
Design-Basis Flood Estimation for Site Characterization at Nuclear Power Plants in the United States of America

International Nuclear Information System (INIS)

Prasad, Rajiv; Hibler, Lyle F.; Coleman, Andre M.; Ward, Duane L.

2011-01-01

The purpose of this document is to describe approaches and methods for estimation of the design-basis flood at nuclear power plant sites. Chapter 1 defines the design-basis flood and lists the U.S. Nuclear Regulatory Commission's (NRC) regulations that require estimation of the design-basis flood. For comparison, the design-basis flood estimation methods used by other Federal agencies are also described. A brief discussion of the recommendations of the International Atomic Energy Agency for estimation of the design-basis floods in its member States is also included.
Design-Basis Flood Estimation for Site Characterization at Nuclear Power Plants in the United States of America

Energy Technology Data Exchange (ETDEWEB)

Prasad, Rajiv; Hibler, Lyle F.; Coleman, Andre M.; Ward, Duane L.

2011-11-01

The purpose of this document is to describe approaches and methods for estimation of the design-basis flood at nuclear power plant sites. Chapter 1 defines the design-basis flood and lists the U.S. Nuclear Regulatory Commission's (NRC) regulations that require estimation of the design-basis flood. For comparison, the design-basis flood estimation methods used by other Federal agencies are also described. A brief discussion of the recommendations of the International Atomic Energy Agency for estimation of the design-basis floods in its member States is also included.
Analysis of near optimum design for small and medium size nuclear power plants

International Nuclear Information System (INIS)

Ahmed, A.A.

1977-01-01

Market surveys in recent years have shown that a significant market would exist among the developing nations of the world for nuclear power plants that would be classified as small to medium sized, provided that these small plants could produce electricity at a unit price comparable to that of equivalent sized fossil fired plants. Nuclear plants in the range of 100 MWe to 500 MWe would fit more effectively into the relatively smaller grids of most developing nations than would the 900 MWe to 1300 MWe units now being constructed in the large industrial nations. Worldwide re-evaluation of the worth of fossil fuels has prompted a re-examination of the competitive position of small to medium sized nuclear generating units compared to comparable fossil fired units, especially in the context of units specifically optimized for the size range of interest, rather than of designs that are simply scaled down versions of the currently available larger units. Since the absolute cost of electricity is more sensitive to external factors such as cost of money, national inflation rate and time required for licensing and construction than to details of design or perhaps even to choice of fuels, and since the cost of electricity generated in small to medium sized fossil fired units is periodically compared to that of scaled down versions of conventional large nuclear units, the point of view taken here is one of comparing the relative generating costs of smaller nuclear units of optimum design with the corresponding costs of scaled down versions of current large nuclear generating units

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