AP1000: Meeting economic goals in a competitive world. Annex 7

International Nuclear Information System (INIS)

Davis, G.; Cummins, E.; Winters, J.

2002-01-01

In the U.S., conditions are becoming more favorable for considering the nuclear option again for new baseload generation. While oil and natural gas prices have risen, the cost of operating the existing fleet of nuclear plants has decreased. Furthermore, an advanced 1000 MWe nuclear plant that will be even more cost-competitive with fossil fuels and natural gas will be available by 2005. Westinghouse, in an effort to further improve on the AP600's cost competitiveness, has developed the AP1000, a two-loop, 1000 MWe, advanced pressurized water reactor (PWR) with passive safety features and extensive plant simplifications to enhance the construction, operation, and maintenance. Like the AP600, the AP1000 uses proven technology that builds on over 30 years of operating PWR experience. Westinghouse has completed design studies that demonstrate that it is feasible to increase the power output of the AP600 to at least 1000 MWe, maintaining its current design configuration and licensing basis. To maximize the cost savings, the AP1000 has been designed within the space constraints of the AP600, while retaining the credibility of proven components and substantial safety margins. The affect on the plant's overnight cost of the increased major components that is required to uprate the AP600 to 1000 MWe is small. This overall cost addition is on the order of 11 percent, while the overall power increase is almost 80 percent. This paper describes the changes made to uprate the AP600 and gives an overview of the plant design. (author)

AP1000 - the new standard for nuclear power

International Nuclear Information System (INIS)

Lipman, Daniel S.

2006-01-01

Full text of publication follows: The AP1000 is the only Generation III+ reactor to receive Final Design Approval (FDA) from the Nuclear Regulatory Commission, and is expected to receive its Design Certification by the end of the year. Building on the proven features of current generation nuclear plants, the AP1000 combines experience with innovation into a design that surpasses current standards of safety and reliability. Use of passive safety features results in a simpler and more compact design that enhances safety, simplifies O and M requirements, and reduces capital and operating costs. At 1117 Mwe, the AP1000 is well suited for almost any grid system and will be fully competitive with combined-cycle gas and comparable fossil fuel plants. The AP1000 is ready to help launch a renaissance in new nuclear plant construction throughout the world. Maturity of Design: In excess of 1300 man-years and $400 million in development funding have been expended on the AP1000. It has undergone extensive, part scale testing at the system, sub-system and component level, in addition to a series of part scale integrated tests. The AP1000 is the most analyzed of the next generation reactors. Simplicity of Design/Economics: The AP1000 uses simplified and innovative passive safety systems to an unprecedented extent. Simplified passive safety systems provide reliable operation, reduced capital costs, and enhanced plant safety with large plant operating margins. The AP1000 features improved reliability through simplicity rather than addition of redundant safety trains. This simpler design is easier and less costly to operate and maintain than larger, more complex plants, while less equipment and smaller buildings translate into lower capital costs and shorter construction durations. After construction, economic benefit will be found in reduced operating and maintenance costs, largely due to reduced operating and maintenance staffing requirements. Construction aspects

European utility requirements (EUR) volume 3 assessment for AP1000

International Nuclear Information System (INIS)

Saiu, G.; Demetri, K.J.

2005-01-01

The EUR (European Utility Requirements) Volume 3 is intended to report the Plant Description, the Compliance Assessment to EUR Volumes 1 and 2, and finally, the Specific Requirements for each specific Nuclear Power Plant Design considered by the EUR. Five subsets of EUR Volume 3, based on EUR Revision B, are already published; all of which are next generation plant designs being developed for Europe beyond 2000. They include : 1) EP1000 - Passive Pressurized Light Water Reactor (3-Loop, 1000 MWe) 2) EPR - Evolutionary Pressurized Light Water Reactor (1500 MWe) 3) BWR90/90+ - Evolutionary Boiling Water Reactor (1400 MWe) 4) ABWR - Evolutionary Boiling Water Reactor (1400 MWe) 5) SWR 1000 - Boiling Water Reactor With Passive Features (1000 MWe) In addition, the following subsets are currently being developed: 1) AP1000 - Passive Pressurized Light Water Reactor (2-Loop, 1117 MWe) 2) VVER AES 92 - Pressurized Water Reactor With Passive Features (1000 MWe) The purpose of this paper is to provide an overview of the program, which started in January 2004 with the EUR group to prepare an EUR Volume 3 Subset for the AP1000 nuclear plant design. The AP1000 EUR compliance assessment, to be performed against EUR Revision C requirements, is an important step for the evaluation of the AP1000 design for application in Europe. The AP1000 compliance assessment is making full use of AP1000 licensing documentation, EPP Phase 2 design activities and EP1000 EUR detailed compliance assessment. As of today, nearly all of the EUR Chapters have been discussed within the EUR Coordination Group. Based on the results of the compliance assessment, it can be stated that the AP1000 design shows a good level of compliance with the EUR Revision C requirements. Nevertheless, the compliance assessment has highlighted areas for where the AP1000 plant deviates from the EUR. The EPP design group has selected the most significant ones for performing detailed studies to quantify the degree of compliance

Reliability analysis on passive residual heat removal of AP1000 based on Grey model

Energy Technology Data Exchange (ETDEWEB)

Qi, Shi; Zhou, Tao; Shahzad, Muhammad Ali; Li, Yu [North China Electric Power Univ., Beijing (China). School of Nuclear Science and Engineering; Beijing Key Laboratory of Passive Safety Technology for Nuclear Energy, Beijing (China); Jiang, Guangming [Nuclear Power Institute of China, Chengdu (China). Science and Technology on Reactor System Design Technology Laboratory

2017-06-15

It is common to base the design of passive systems on the natural laws of physics, such as gravity, heat conduction, inertia. For AP1000, a generation-III reactor, such systems have an inherent safety associated with them due to the simplicity of their structures. However, there is a fairly large amount of uncertainty in the operating conditions of these passive safety systems. In some cases, a small deviation in the design or operating conditions can affect the function of the system. The reliability of the passive residual heat removal is analysed.

AP1000{sup R} severe accident features and post-Fukushima considerations

Energy Technology Data Exchange (ETDEWEB)

Scobel, J. H.; Schulz, T. L.; Williams, M. G. [Westinghouse Electric Company, LLC, 1000 Westinghouse Dr., Cranberry Township, PA 16066 (United States)

2012-07-01

The AP1000{sup R} passive nuclear power plant is uniquely equipped to withstand an extended station blackout scenario such as the events following the earthquake and tsunami at Fukushima without compromising core and containment integrity. The AP1000 plant shuts down the reactor, cools the core, containment and spent fuel pool for more than 3 days using passive systems that do not require AC or DC power or operator actions. Following this passive coping period, minimal operator actions are needed to extend the operation of the passive features to 7 days using installed equipment. To provide defense-in-depth for design extension conditions, the AP1000 plant has engineered features that mitigate the effects of core damage. Engineered features retain damaged core debris within the reactor vessel as a key feature. Other aspects of the design protect containment integrity during severe accidents, including unique features of the AP1000 design relative to passive containment cooling with water and air, and hydrogen management. (authors)

AP1000 will meet the challenges of near-term deployment

International Nuclear Information System (INIS)

Matzie, Regis A.

2008-01-01

The world demand for energy is growing rapidly, particularly in developing countries that are trying to raise the standard of living for billions of people, many of whom do not have access to electricity or clean water. Climate change and the concern for increased emissions of green house gases have brought into question the future primary reliance of fossil fuels. With the projected worldwide increase in energy demand, concern for the environmental impact of carbon emissions, and the recent price volatility of fossil fuels, nuclear energy is undergoing a rapid resurgence. This 'nuclear renaissance' is broad based, reaching across Asia, North America, Europe, as well as selected countries in Africa and South America. Many countries have publicly expressed their intentions to pursue the construction of new nuclear energy plants. Some countries that have previously turned away from commercial nuclear energy are reconsidering the advisability of this decision. This renaissance is facilitated by the availability of more advanced reactor designs than are operating today, with improved safety, economy, and operations. One such design, the Westinghouse AP1000 advanced passive plant, has been a long time in the making! The development of this passive technology started over two decades ago from an embryonic belief that a new approach to design was needed to spawn a nuclear renaissance. The principal challenges were seen as ensuring reactor safety by requiring less reliance on operator actions and overcoming the high plant capital cost of nuclear energy. The AP1000 design is based on the use of innovative passive technology and modular construction, which require significantly less equipment and commodities that facilitate a more rapid construction schedule. Because Westinghouse had the vision and the perseverance to continue the development of this passive technology, the AP1000 design is ready to meet today's challenge of near-term deployment

AP1000{sup TM} plant modularization

Energy Technology Data Exchange (ETDEWEB)

Cantarero L, C.; Demetri, K. J. [Westinghouse Electric Co., 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States); Quintero C, F. P., E-mail: cantarc@westinghouse.com [Westinghouse Electric Spain, Padilla 17, 28006 Madrid (Spain)

2016-09-15

The AP1000{sup TM} plant is an 1100 M We pressurized water reactor (PWR) with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance and safety. Modules are used extensively in the design of the AP1000 plant nuclear island. The AP1000 plant uses modern, modular-construction techniques for plant construction. The design incorporates vendor-designed skids and equipment packages, as well as large, multi-ton structural modules and special equipment modules. Modularization allows traditionally sequential construction tasks to be completed simultaneously. Factory-built modules can be installed at the site in a planned construction schedule. The modularized AP1000 plant allows many more construction activities to proceed in parallel. This reduces plant construction calendar time, thus lowering the costs of plant financing. Furthermore, performing less work onsite significantly reduces the amount of skilled field-craft labor, which costs more than shop labor. In addition to labor cost savings, doing more welding and fabrication in a factory environment raises the quality of work, allowing more scheduling flexibility and reducing the amount of specialized tools required onsite. The site layout for the AP1000 plant has been established to support modular construction and efficient operations during construction. The plant layout is compact, using less space than previous conventional plant layouts. This paper provides and overview of the AP1000 plant modules with an emphasis on structural modules. Currently the Westinghouse AP1000 plant has four units under construction in China and four units under construction in the United States. All have shown successful fabrication and installation of various AP1000 plant modules. (Author)

AP1000"T"M plant modularization

International Nuclear Information System (INIS)

Cantarero L, C.; Demetri, K. J.; Quintero C, F. P.

2016-09-01

The AP1000"T"M plant is an 1100 M We pressurized water reactor (PWR) with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance and safety. Modules are used extensively in the design of the AP1000 plant nuclear island. The AP1000 plant uses modern, modular-construction techniques for plant construction. The design incorporates vendor-designed skids and equipment packages, as well as large, multi-ton structural modules and special equipment modules. Modularization allows traditionally sequential construction tasks to be completed simultaneously. Factory-built modules can be installed at the site in a planned construction schedule. The modularized AP1000 plant allows many more construction activities to proceed in parallel. This reduces plant construction calendar time, thus lowering the costs of plant financing. Furthermore, performing less work onsite significantly reduces the amount of skilled field-craft labor, which costs more than shop labor. In addition to labor cost savings, doing more welding and fabrication in a factory environment raises the quality of work, allowing more scheduling flexibility and reducing the amount of specialized tools required onsite. The site layout for the AP1000 plant has been established to support modular construction and efficient operations during construction. The plant layout is compact, using less space than previous conventional plant layouts. This paper provides and overview of the AP1000 plant modules with an emphasis on structural modules. Currently the Westinghouse AP1000 plant has four units under construction in China and four units under construction in the United States. All have shown successful fabrication and installation of various AP1000 plant modules. (Author)

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

International Nuclear Information System (INIS)

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

2016-01-01

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

AP1000{sup R} nuclear power plant safety overview for spent fuel cooling

Energy Technology Data Exchange (ETDEWEB)

Gorgemans, J.; Mulhollem, L.; Glavin, J.; Pfister, A.; Conway, L.; Schulz, T.; Oriani, L.; Cummins, E.; Winters, J. [Westinghouse Electric Company LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

2012-07-01

The AP1000{sup R} plant is an 1100-MWe class pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, safety and costs. The AP1000 design uses passive features to mitigate design basis accidents. The passive safety systems are designed to function without safety-grade support systems such as AC power, component cooling water, service water or HVAC. Furthermore, these passive features 'fail safe' during a non-LOCA event such that DC power and instrumentation are not required. The AP1000 also has simple, active, defense-in-depth systems to support normal plant operations. These active systems provide the first level of defense against more probable events and they provide investment protection, reduce the demands on the passive features and support the probabilistic risk assessment. The AP1000 passive safety approach allows the plant to achieve and maintain safe shutdown in case of an accident for 72 hours without operator action, meeting the expectations provided in the U.S. Utility Requirement Document and the European Utility Requirements for passive plants. Limited operator actions are required to maintain safe conditions in the spent fuel pool via passive means. In line with the AP1000 approach to safety described above, the AP1000 plant design features multiple, diverse lines of defense to ensure spent fuel cooling can be maintained for design-basis events and beyond design-basis accidents. During normal and abnormal conditions, defense-in-depth and other systems provide highly reliable spent fuel pool cooling. They rely on off-site AC power or the on-site standby diesel generators. For unlikely design basis events with an extended loss of AC power (i.e., station blackout) or loss of heat sink or both, spent fuel cooling can still be provided indefinitely: - Passive systems, requiring minimal or no operator actions, are sufficient for at least 72 hours under all

Enhancing AP1000 reactor accident management capabilities for long term accidents

International Nuclear Information System (INIS)

Jiang Pingting; Liu Mengying; Duan Chengjie; Liao Yehong

2015-01-01

Passive safety actions are considered as main measures under severe accident in AP1000 power plant. However, risk is still existed. According to PSA, several probable scenarios for AP1000 nuclear power plant are analyzed in this paper with MAAP the severe accident analysis code. According to the analysis results, several deficiencies of AP1000 severe accident management are found. The long term cooling and containment depressurization capability for AP1000 power plant appear to be most important factors under such accidents. Then, several temporary strategies for AP1000 power plant are suggested, including PCCWST temporary water supply strategy after 72h, temporary injection strategy for IRWST, hydrogen relief action in fuel building, which would improve the safety of AP1000 power plant. At last, assessments of effectiveness for these strategies are performed, and the results are compared with analysis without these strategies. The comparisons showed that correct actions of these strategies would effectively prevent the accident process of AP1000 power plant. (author)

AP1000 plant construction in China: Ansaldo Nucleare contribution

International Nuclear Information System (INIS)

Frogheri, Monica; Saiu, Gianfranco

2009-01-01

On 24th of July 2007 Westinghouse Electric Co. signed landmark contracts with China's State Nuclear Power Technology Corporation (SNPTC), to provide four AP1000 nuclear power plants in China. The AP1000 is a two-loop 1117 MWe Pressurized Water Reactor (PWR). It is based on proven technology, but with an emphasis on safety features that rely on natural driving forces, such as pressurized gas, gravity flow, natural circulation flow and convection. Ansaldo Nucleare has provided a significant support to the passive plant technology development and, starting from 2000, is cooperating with Westinghouse to development of the AP1000 Plant. In the frame of the AP1000 Chinese agreement, Ansaldo Nucleare, in Joint Venture with Mangiarotti Nuclear, has signed a contract with Westinghouse for the design and the supply of innovative components to be installed in the first AP1000 unit to be constructed at the Sanmen site. The contract includes: the design of the steel containment vessel, preparation of construction and fabrication, specifications, design and supply of SCV mechanical penetrations, air locks and equipment hatches. Moreover, Ansaldo Nucleare is in charge of the final design of the AP1000 PRHR-HX and together with Mangiarotti Nuclear will supply the component for the Sanmen Unit 1 NPP. The paper presents an overview of the design and manufacturing activities performed by Ansaldo Nucleare and its partners for the AP1000 plant in China. (authors)

Study on time management of FCD concrete pouring AP1000 NPP and its application

International Nuclear Information System (INIS)

Wu Jie; Yang Ming; Cong Jiuyuan

2010-01-01

Haiyang nuclear power plant adopts the advanced third-generation nuclear power technology, AP1000, the design concept of passive system and the methods of modularization and 'open-top' construction greatly shortens the construction period. At the same time, higher requirements are put forward to the management of construction, quality and progress control. This paper will apply the statistical process control theory and method to the time management of FCD (First Concrete Day) to Unit 1 of Haiyang AP1000 nuclear power plant, and it brings up the mathematical model of time management based on verification through modeling, data analysis, model optimization and the actual construction work. The theory and method studied in this paper can not be only applied to the FCD concrete pouring for Unit 1 of Haiyang AP1000 nuclear power plant, but also have great referential and guiding significance to the continuous concreting of the mass concrete of the follow-up similar construction. (authors)

Learning through delivery, Westinghouse AP1000 plant construction

International Nuclear Information System (INIS)

Gorgemans, J.; Hinman, R.D.; Steuck, C.M.; Greco, P.L.

2014-01-01

The AP1000 plant, which is a 1100 MWe class pressurized water reactor with passive safety features, is designed around a conventional 2 loop, 2 steam generator primary system configuration with 2 hot legs, 4 reactor coolant pumps directly mounted in the steam generator lower head and 4 cold legs. A particular feature of AP1000 is its modular construction to minimize the time and cost of construction. Modular construction allows activities to be run in parallel, it allows more activities to be performed in a controlled factory instead of in the field, and it provides a better level of quality. The AP1000 plant design includes 106 structural modules and 52 mechanical modules. Structural modules include all penetrations for piping, cable trays, HVAC duct runs, and all reinforcement for pipe, equipment hangers, and supports. Structural modules are shipped in sub-modules to support transportation by rail or truck or barge. Mechanical modules contain equipment such as pumps, tanks, heat exchangers, air-handling units, and filters along with interconnecting pipes, valves, instruments, wiring and support services. Modular construction requires strong coordination between engineering, supply chain and construction. A total of 8 AP1000 units are currently under construction in China and in the United States. The lessons learned and best practices of each new AP1000 construction are systematically incorporated into the standard design. (A.C.)

The AP1000{sup R} China projects move forward to construction completion and equipment installation

Energy Technology Data Exchange (ETDEWEB)

Harrop, G. [Westinghouse Electric Company LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

2012-07-01

The AP1000 design is the only Generation III+ technology to receive design certification from the U.S. Nuclear Regulatory Commission. This evolutionary design provides the highest safety and performance standards and has several distinct advantages over other designs, including improved operations and reduced construction schedule risks through the use of modern, modular, engineering principles that allow construction and fabrication tasks traditionally performed in sequence to be undertaken in parallel. Since the first granting of Design Certification in 2005 by the NRC, the AP1000 design has been modified to meet emergent NRC requirements such as those requiring the design to withstand the impact of an aircraft crash. Both domestic and foreign utilities have turned to the Westinghouse AP1000 plant design to meet their near - and long-term sustainable energy needs. The first ever deployment of this advanced U.S. nuclear power technology began in China in 2007 with the award of a contract to build four AP1000 units, constructed in pairs at the coastal sites of Sanmen (Zhejiang Province) and Haiyang (Shandong Province). Currently, all four units are at an advanced stage of construction. The commercial operation date for Sanmen Unit 1 is November 2013 followed by Haiyang Unit 1 being operational in May 2014. Construction and equipment manufacture is at an advanced stage. Sanmen Unit 1 equipment that has been delivered includes the reactor vessel, the reactor vessel closure head, the passive residual heat removal heat exchanger, the integrated head package, the polar crane, and the refueling machine. The steam generators are also completed. The RV was installed within the containment vessel building in September 2011. The installation of this major equipment will allow the setting of the containment vessel top head. Haiyang Unit 1 is also achieving significant progress. Significant benefits continue to be realized as a result of lessons learned and experience gained

Characterization of liquid entrainment in the AP1000 automatic depressurization system from APEX tests

International Nuclear Information System (INIS)

Richard F Wright; Terry L Schulz; Jose N Reyes; John Groome

2005-01-01

Full text of publication follows: The AP1000 is a 1000 MWe advanced nuclear power plant that uses passive safety features to enhance plant safety and to provide significant and measurable improvements in plant simplification, reliability, investment protection and plant costs. The AP1000 relies heavily on the 600 MWe AP600 which received design certification in 1999. A critical part of the AP600 design certification process involved the testing of the passive safety systems. A one-fourth height, one-fourth pressure test facility, APEX-600, was constructed at the Oregon State University to study design basis events, and to provide a body of data to be used to validate the computer models used to analyze the AP600. This facility was extensively modified to reflect the design changes for AP1000 including higher power in the electrically heated rods representing the reactor core, and changes in the size of the pressurizer, core makeup tanks and automatic depressurization system. The APEX-1000 test facility was used to perform design basis accident simulations and separate effects tests to support the AP1000 design certification process. In the event of a LOCA, the AP1000 passive core cooling system provides sources of core makeup water along with an automatic depressurization system (ADS) consisting of several stages of valves which reduce the reactor coolant system pressure in a controlled manner. The final stage of this system, ADS-4, consists of four large valves that open off the hot legs, reducing the pressure to allow gravity injection from the in-containment refueling water storage tank (IRWST) and eventually the containment sump. The 67% increase in power from AP600 to AP1000 results in proportionally larger steam velocities exiting the core. Higher steam velocities could increases the potential for significant liquid entrainment out the ADS-4 lines, affecting the liquid inventory in the reactor. Tests were performed in APEX-1000 to characterize the two

The Design of Cooling System Model on The AP1000 Containment

International Nuclear Information System (INIS)

Daddy Setyawan; Yerri Noer Kartiko; Aryadi Suwono; Ari Darmawan Pasek; Nathanael P Tandian; Efrizon Umar

2009-01-01

The policy of national energy leads to the utilization of new energy as nuclear energy, and also contains some efforts to increase reactor safety and optimizing in the design of safety system component such as passive cooling system on reactor containment tank. Because of this, the assessment of safety level to passive safety system needs to be made. To increase the understanding it, the design of cooling system model on containment tank should be done to get safety level on cooling system in the AP1000 containment. To reach the similar model with reality and inexpensive cost, we should make assessment about similarity and dimensionless number. While the heat transfer of air natural circulation and water spray cooling system are a result of gravity approach, we can calculate Grashof modification number and Reynolds number respectively. By this approach, we have a factor of forty for laboratory model. From this model, we hope that we get characteristic correlation to heat transfer on the containment of AP1000 for both air natural circulation and water spray result from gravity. Finally, we can assess the safety level of passive cooling system on the AP1000 containment. (author)

AP1000 Design for Security

International Nuclear Information System (INIS)

Long, L.B.; Cummins, W.E.; Winters, J.W.

2006-01-01

Nuclear power plants are protected from potential security threats through a combination of robust structures around the primary system and other vital equipment, security systems and equipment, and defensive strategy. The overall objective for nuclear power plant security is to protect public health and safety by ensuring that attacks or sabotage do not challenge the ability to safely shutdown the plant or protect from radiological releases. In addition, plants have systems, features and operational strategies to cope with external conditions, such as loss of offsite power, which could be created as part of an attack. Westinghouse considered potential security threats during design of the AP1000 PWR. The differences in plant configuration, safety system design, and safe shutdown equipment between existing plants and AP1000 affect potential vulnerabilities. This paper provides an evaluation of AP1000 with respect to vulnerabilities to security threats. The AP1000 design differs from the design of operating PWRs in the US in the configuration and the functional requirements for safety systems. These differences are intentional departures from conventional PWR designs which simplify plant design and enhance overall safety. The differences between the AP1000 PWR and conventional PWRs can impact vulnerabilities to security threats. The NRC addressed security concerns as part of their reviews for AP1000 Design Certification, and did not identify any security issues of concern. However, much of the detailed security design information for the AP1000 was deferred to the combined Construction and Operating License (COL) phase as many of the security issues are site-specific. Therefore, NRC review of security issues related to the AP1000 is not necessarily complete. Further, since the AP1000 plant design differs from existing PWRs, it is not obvious that the analyses and assessments prepared for existing plants also apply to the AP1000. We conclude that, overall, the AP1000

ANALISIS KONDISI TERAS REAKTOR DAYA MAJU AP1000 PADA KECELAKAAN SMALL BREAK LOCA

Directory of Open Access Journals (Sweden)

Andi Sofrany Ekariansyah

2015-06-01

, mixture level, temperatur kelongsong, small break LOCA, RELAP5.   ABSTRACT ANALYSIS ON THE CORE CONDITION OF AP1000 ADVANCED POWER REACTOR DURING SMALL BREAK LOCA. Accident due to the loss of coolant from the reactor boundary is an anticipated design basis event in the design of power reactor adopting the Generation II up to IV technology. Small break LOCA leads to more significant impact on safety compared to the large break LOCA as shown in the Three-Mile Island (TMI. The focus of this paper is the small break LOCA analysis on the Generation III+ advanced power reactor of AP1000 by simulating three different initiating events, which are inadvertent opening of Automatic Depressurization System (ADS, double-ended break on one of Direct Vessel Injection (DVI pipe, and 10 inch diameter split break on one of cold leg pipe. Methodology used is by simulating the events on the AP1000 model developed using RELAP5/SCDAP/Mod3.4. The impact analyzed is the core condition during the small break LOCA consisting of the mixture level occurrence and the fuel cladding temperature transient. The results show that the mixture level for all small break LOCA events are above the active core height, which indicates no core uncovery event. The development of the mixture level affect the fuel cladding temperature transient, which shows a decreasingly trend after the break, and the effectifeness of core cooling. Those results are identical with the results of other code of NOTRUMP. The resulted core cooling is also due to the function of coolant injection from passive safety feature, which is unique in the AP1000 design. In overall, the result of analysis shows that the AP1000 model developed by the RELAP5 can be used for analysis of design basis accident considered in the AP1000 advanced power reactor. Keywords: analysis, mixture level, fuel cladding temperature, small break LOCA, RELAP5.

AP1000R licensing and deployment in the United States

International Nuclear Information System (INIS)

Jordan, R. P.; Russ, P. A.; Filiak, P. P.; Castiglione, L. L.

2012-01-01

In recent years, both domestic and foreign utilities have turned to the standardized Westinghouse AP1000 plant design in satisfying their near - and long-term - sustainable energy needs. As direct support to these actions, licensing the AP1000 design has played a significant role by providing one of the fundamental bases in clearing regulatory hurdles leading to the start of new plant construction. Within the U.S. alone, Westinghouse AP1000 licensing activities have reached unprecedented milestones with the approvals of both AP1000 Design Certification and Southern Company's combined construction permit and operating license (COL) application directly supporting the construction of two new nuclear plants in Georgia. Further COL application approvals are immediately pending for an additional two AP1000 plants in South Carolina. And, across the U.S. nuclear industry spectrum, there are 10 other COL applications under regulatory review representing some 16 new plants at 10 sites. In total, these actions represent the first wave of new plant licensing under the regulatory approval process since 1978. Fundamental to the Nuclear Regulatory Commission's AP1000 Design Certification is the formal recognition of the AP1000 passive safety design through regulatory acceptance rulemaking. Through recognition and deployment of the AP1000 Design Certification, the utility licensee / operator of this reactor design are now offered an opportunity to use a simplified 'one-step' combined license process, thereby managing substantial back-end construction schedule risk from regulatory and intervention delays. Application of this regulatory philosophy represents both acceptance and encouragement of standardized reactor designs like the AP1000. With the recent AP1000 Design Certification and utility COL acceptances, the fundamental licensing processes of this philosophy have successfully proven the attainment of significant milestones with the next stage licensing actions directed

Significant advantages of the safety-first concept in construction, operation, and maintenance of the Westinghosue AP1000 reactor

International Nuclear Information System (INIS)

Cummins, E.; Benitz, K.

2004-01-01

In June 2003, the U.S. Nuclear Regulatory Commission (USNRC) published a draft opinion about safety of the AP1000 Westinghouse pressurized water reactor with 'passive safety' features. The report constitutes an important milestone in the development of the next generation of safe and cost-efficient nuclear power plants. A new AP1000 can be absolutely competitive with fossil fired power plants and may be able to revive the construction of new nuclear power plants worldwide. The reason for designing the AP1000 were safety considerations. The use of passive safety systems at the same time entails a considerable reduction in the costs of design, maintenance, and operation of an AP1000 plant. Independent experts confirmed that an AP1000 can be erected within three years or even less. The estimated electricity generating costs of an AP1000 plant in the United States amount to US Cent 3.2 to 3.6 per kilowatthour. (orig.)

Westinghouse AP1000® PWR: Meeting Customer Commitments and Market Needs

International Nuclear Information System (INIS)

Shulyak, Nick

2014-01-01

Westinghouse Electric Company once again sets a new industry standard with the AP1000 reactor. Historically, Westinghouse plant designs and technology have forged the cutting edge of worldwide nuclear technology. Today, about 50 percent of the world's 440 nuclear plants are based on Westinghouse technology. The AP1000 is the safest and most economical nuclear power plant available in the worldwide commercial marketplace, and is the only Generation III+ reactor to receive Design Certification from the U.S. Nuclear Regulatory Commission (NRC). The AP1000 features proven technology, innovative passive safety systems and offers: Unequalled safety, Economic competitiveness, Improved and more efficient operations. The AP1000 builds and improves upon the established technology of major components used in current Westinghouse-designed plants with proven, reliable operating experience over the past 50 years. These components include: Steam generators, Digital instrumentation and controls, Fuel, Pressurizers, Reactor vessels. Simplification was a major design objective for the AP1000. The simplified plant design includes overall safely systems, normal operating systems, the control room, construction techniques, and instrumentation and control systems. The result is a plant that is easier and less expensive to build, operate and maintain. The AP1000 design saves money and time with an accelerated construction time period of approximately 36 months, from the pouring of first concrete to the loading of fuel. Also, the innovative AP1000 features: 50% fewer safety-related valves, 80% less safety-related piping, 85% less control cable, 35% fewer pumps , 45% less seismic building volume. Eight AP1000 units under construction worldwide-Four units in China-Four units in the United States. (author)

Dynamic Response of AP1000 Nuclear Island Due to Safe Shutdown Earthquake Loading

Directory of Open Access Journals (Sweden)

Gan Buntara S.

2017-01-01

Full Text Available AP1000 is a standard nuclear power plant developed by Westinghouse and its partners by using an advanced passive safety feature. Among the five principle building structures, namely the nuclear island, turbine building, annex building, diesel generator building and radwaste building, the safety of the nuclear island building is the most concerned. This paper investigates the dynamic response of the nuclear island building of the AP1000 plant subjected to safe shutdown earthquake loadings. A finite element model for the building, which is assumed to be built in a hard-rock base, is developed and its dynamic response is computed with the aid of the commercial finite element package ANSYS. The dynamic characteristics, including the natural frequencies, the vibration modes, and the time histories for displacements, velocities, and accelerations of the building are obtained for two typical safe shutdown earthquakes, El Centro and Kobe earthquakes. The dynamic behavior of the building due to the earthquakes and its safety is examined and highlighted.

Application case study of AP1000 automatic depressurization system (ADS) for reliability evaluation by GO-FLOW methodology

Energy Technology Data Exchange (ETDEWEB)

Hashim, Muhammad, E-mail: hashimsajid@yahoo.com; Hidekazu, Yoshikawa, E-mail: yosikawa@kib.biglobe.ne.jp; Takeshi, Matsuoka, E-mail: mats@cc.utsunomiya-u.ac.jp; Ming, Yang, E-mail: myang.heu@gmail.com

2014-10-15

Highlights: • Discussion on reasons why AP1000 equipped with ADS system comparatively to PWR. • Clarification of full and partial depressurization of reactor coolant system by ADS system. • Application case study of four stages ADS system for reliability evaluation in LBLOCA. • GO-FLOW tool is capable to evaluate dynamic reliability of passive safety systems. • Calculated ADS reliability result significantly increased dynamic reliability of PXS. - Abstract: AP1000 nuclear power plant (NPP) utilized passive means for the safety systems to ensure its safety in events of transient or severe accidents. One of the unique safety systems of AP1000 to be compared with conventional PWR is the “four stages Automatic Depressurization System (ADS)”, and ADS system originally works as an active safety system. In the present study, authors first discussed the reasons of why four stages ADS system is added in AP1000 plant to be compared with conventional PWR in the aspect of reliability. And then explained the full and partial depressurization of RCS system by four stages ADS in events of transient and loss of coolant accidents (LOCAs). Lastly, the application case study of four stages ADS system of AP1000 has been conducted in the aspect of reliability evaluation of ADS system under postulated conditions of full RCS depressurization during large break loss of a coolant accident (LBLOCA) in one of the RCS cold legs. In this case study, the reliability evaluation is made by GO-FLOW methodology to determinate the influence of ADS system in dynamic reliability of passive core cooling system (PXS) of AP1000, i.e. what will happen if ADS system fails or successfully actuate. The GO-FLOW is success-oriented reliability analysis tool and is capable to evaluating the systems reliability/unavailability alternatively to Fault Tree Analysis (FTA) and Event Tree Analysis (ETA) tools. Under these specific conditions of LBLOCA, the GO-FLOW calculated reliability results indicated

EP 1000 -The European Passive Plant

International Nuclear Information System (INIS)

Cummins, Ed; Oyarzabal, Mariano; Saiu, Gianfranco

1998-01-01

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

Study on modular construction management in AP1000 nuclear plant project

International Nuclear Information System (INIS)

Fang Xiaopeng; Shen Wenrong; Sun Kebin; Wei Zhong

2010-01-01

The construction of AP1000 Nuclear Power Plant (NPP) has commenced in China. The AP1000 NPP features a passive design concept and modular construction technology. Based on the management of the construction of current AP1000 NNP, this paper describes the effects on Nuclear Island (NI) construction project management resulting from modular construction technology, as well as new construction techniques and methods. This paper puts forward new requirements for construction schedule management of the nuclear island construction at different levels. The AP1000 NI construction logic features the parallel construction of civil and structural erection as the main approach, with the integrated schedule of module fabrication, assembly and installation as support. The structural modules of AP1000 project are prefabricated in shop, delivered to site as sub-modules and assembled to integrated structural module. The assembled module is transported to the construction site, hoisted and finally set in NI. This paper illustrates how to ensure the construction quality of structural modules by analyzing the interface process and key links in the quality control program, and introduces how to ensure the safety of heavy structural components during various construction phases by evaluating and analyzing the construction safety process. This paper also makes an analysis of the safe environment for the assembly and installation of Containment Vessel, the management of product protection and personnel safety inside the Containment Building during 'Open Top' construction, raises to implement effective protection for the numerous pre-set mechanical modules and equipments, as well as personnel safety protection programs and measures. The modular construction feature of AP1000 NPP design requires technique improvement and management innovation during the NI construction. This paper makes a study and research on the control management of schedule, quality and safety of AP1000 NPP NI

AP1000{sup R} licensing and deployment in the United States

Energy Technology Data Exchange (ETDEWEB)

Jordan, R. P.; Russ, P. A.; Filiak, P. P.; Castiglione, L. L. [Westinghouse Electric Company LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

2012-07-01

In recent years, both domestic and foreign utilities have turned to the standardized Westinghouse AP1000 plant design in satisfying their near - and long-term - sustainable energy needs. As direct support to these actions, licensing the AP1000 design has played a significant role by providing one of the fundamental bases in clearing regulatory hurdles leading to the start of new plant construction. Within the U.S. alone, Westinghouse AP1000 licensing activities have reached unprecedented milestones with the approvals of both AP1000 Design Certification and Southern Company's combined construction permit and operating license (COL) application directly supporting the construction of two new nuclear plants in Georgia. Further COL application approvals are immediately pending for an additional two AP1000 plants in South Carolina. And, across the U.S. nuclear industry spectrum, there are 10 other COL applications under regulatory review representing some 16 new plants at 10 sites. In total, these actions represent the first wave of new plant licensing under the regulatory approval process since 1978. Fundamental to the Nuclear Regulatory Commission's AP1000 Design Certification is the formal recognition of the AP1000 passive safety design through regulatory acceptance rulemaking. Through recognition and deployment of the AP1000 Design Certification, the utility licensee / operator of this reactor design are now offered an opportunity to use a simplified 'one-step' combined license process, thereby managing substantial back-end construction schedule risk from regulatory and intervention delays. Application of this regulatory philosophy represents both acceptance and encouragement of standardized reactor designs like the AP1000. With the recent AP1000 Design Certification and utility COL acceptances, the fundamental licensing processes of this philosophy have successfully proven the attainment of significant milestones with the next stage licensing

Introduction to sump screen downstream effect analysis of AP1000 nuclear power plant

International Nuclear Information System (INIS)

Zhang Qinghua; Liu Yu; Chai Guohan

2010-01-01

The design of AP1000 takes into account the potential impact of debris clogging on sump screen. In this article, the technical background of sump screen issue and the design characteristics of AP1000 to address the sump screen blockage issue are introduced. The article focuses on the 'downstream effect' analysis method, acceptance criteria and analysis result of AP1000 sump screen. Although the design of AP1000 is different with traditional PWR, the author expects to bring some reference to advance the downstream effect analysis in China through the introduction. (authors)

Analysis of the passive heat removal enhancement for AP1000 containment due to the partially wetted coverage

Energy Technology Data Exchange (ETDEWEB)

Li, Cheng, E-mail: 510395453@qq.com [State Nuclear Power Technology Research & Development Center, 102209 Beijing (China); Li, Le [Tsinghua University, Institute of Nuclear and New Energy Technology, 100084 Beijing (China); Li, Junming [Tsinghua University, Key Laboratory for Thermal Science and Power Engineering of Ministry of Education, Department of Thermal Engineering, Beijing 100084 (China); Zhang, Yajun [Tsinghua University, Institute of Nuclear and New Energy Technology, 100084 Beijing (China); Li, Zhihui [State Nuclear Power Technology Research & Development Center, 102209 Beijing (China)

2017-03-15

distributions. Results show that the maximum heat transfer enhancement can reach 63% and this enhancement peak appears when the dry strip fraction approximately equals 90%. The influences of dry coverage and dry-wet strip width were carefully discussed. It indicates that the heat transfer enhancement for small dry-wet strip is linear with dry strip fraction (0 < α < 80%). Finally, according to the wetted coverage ranges for AP1000 passive containment cooling system, an empirical correlation for heat transfer enhancement is given and it can be used to improve PCCS analysis.

Analysis of the passive heat removal enhancement for AP1000 containment due to the partially wetted coverage

International Nuclear Information System (INIS)

Li, Cheng; Li, Le; Li, Junming; Zhang, Yajun; Li, Zhihui

2017-01-01

distributions. Results show that the maximum heat transfer enhancement can reach 63% and this enhancement peak appears when the dry strip fraction approximately equals 90%. The influences of dry coverage and dry-wet strip width were carefully discussed. It indicates that the heat transfer enhancement for small dry-wet strip is linear with dry strip fraction (0 < α < 80%). Finally, according to the wetted coverage ranges for AP1000 passive containment cooling system, an empirical correlation for heat transfer enhancement is given and it can be used to improve PCCS analysis.

AP1000R design robustness against extreme external events - Seismic, flooding, and aircraft crash

International Nuclear Information System (INIS)

Pfister, A.; Goossen, C.; Coogler, K.; Gorgemans, J.

2012-01-01

Both the International Atomic Energy Agency (IAEA) and the U.S. Nuclear Regulatory Commission (NRC) require existing and new nuclear power plants to conduct plant assessments to demonstrate the unit's ability to withstand external hazards. The events that occurred at the Fukushima-Dai-ichi nuclear power station demonstrated the importance of designing a nuclear power plant with the ability to protect the plant against extreme external hazards. The innovative design of the AP1000 R nuclear power plant provides unparalleled protection against catastrophic external events which can lead to extensive infrastructure damage and place the plant in an extended abnormal situation. The AP1000 plant is an 1100-MWe pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance and safety. The plant's compact safety related footprint and protection provided by its robust nuclear island structures prevent significant damage to systems, structures, and components required to safely shutdown the plant and maintain core and spent fuel pool cooling and containment integrity following extreme external events. The AP1000 nuclear power plant has been extensively analyzed and reviewed to demonstrate that it's nuclear island design and plant layout provide protection against both design basis and extreme beyond design basis external hazards such as extreme seismic events, external flooding that exceeds the maximum probable flood limit, and malicious aircraft impact. The AP1000 nuclear power plant uses fail safe passive features to mitigate design basis accidents. The passive safety systems are designed to function without safety-grade support systems (such as AC power, component cooling water, service water, compressed air or HVAC). The plant has been designed to protect systems, structures, and components critical to placing the reactor in a safe shutdown condition within the steel containment vessel which is

AP1000 passive core cooling system pre-operational tests procedure definition and simulation by means of Relap5 Mod. 3.3 computer code

International Nuclear Information System (INIS)

Lioce, D.; Asztalos, M.; Alemberti, A.; Barucca, L.; Frogheri, M.; Saiu, G.

2012-01-01

Highlights: ► Two AP1000 Core Make-up Tanks pre-operational tests procedures have been defined. ► The two tests have been simulated by means of the Relap5 computer code. ► Results show the tests can be successfully performed with the selected procedures. - Abstract: The AP1000 ® plant is an advanced Pressurized Water Reactor designed and developed by Westinghouse Electric Company which relies on passive safety systems for core cooling, containment isolation and containment cooling, and maintenance of main control room emergency habitability. The AP1000 design obtained the Design Certification by NRC in January 2006, as Appendix D of 10 CFR Part 52, and it is being built in two locations in China. The AP1000 plant will be the first commercial nuclear power plant to rely on completely passive safety systems for core cooling and its licensing process requires the proper operation of these systems to be demonstrated through some pre-operational tests to be conducted on the real plant. The overall objective of the test program is to demonstrate that the plant has been constructed as designed, that the systems perform consistently with the plant design, and that activities culminating in operation at full licensed power including initial fuel load, initial criticality, and power increase to full load are performed in a controlled and safe manner. Within this framework, Westinghouse Electric Company and its partner Ansaldo Nucleare S.p.A. have strictly collaborated, being Ansaldo Nucleare S.p.A. in charge of the simulation of some pre-operational tests and supporting Westinghouse in the definition of tests procedures. This paper summarizes the work performed at Ansaldo Nucleare S.p.A. in collaboration with Westinghouse Electric Company for the Core Makeup Tank (CMT) tests, i.e. the CMTs hot recirculation test and the CMTs draindown test. The test procedure for the two selected tests has been defined and, in order to perform the pre-operational tests simulations, a

AP1000 passive core cooling system pre-operational tests procedure definition and simulation by means of Relap5 Mod. 3.3 computer code

Energy Technology Data Exchange (ETDEWEB)

Lioce, D., E-mail: donato.lioce@aen.ansaldo.it [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy); Asztalos, M., E-mail: asztalmj@westinghouse.com [Westinghouse Electric Company, Cranberry Twp, PA 16066 (United States); Alemberti, A., E-mail: alessandro.alemberti@aen.ansaldo.it [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy); Barucca, L. [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy); Frogheri, M., E-mail: monicalinda.frogheri@aen.ansaldo.it [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy); Saiu, G., E-mail: gianfranco.saiu@aen.ansaldo.it [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy)

2012-09-15

Highlights: Black-Right-Pointing-Pointer Two AP1000 Core Make-up Tanks pre-operational tests procedures have been defined. Black-Right-Pointing-Pointer The two tests have been simulated by means of the Relap5 computer code. Black-Right-Pointing-Pointer Results show the tests can be successfully performed with the selected procedures. - Abstract: The AP1000{sup Registered-Sign} plant is an advanced Pressurized Water Reactor designed and developed by Westinghouse Electric Company which relies on passive safety systems for core cooling, containment isolation and containment cooling, and maintenance of main control room emergency habitability. The AP1000 design obtained the Design Certification by NRC in January 2006, as Appendix D of 10 CFR Part 52, and it is being built in two locations in China. The AP1000 plant will be the first commercial nuclear power plant to rely on completely passive safety systems for core cooling and its licensing process requires the proper operation of these systems to be demonstrated through some pre-operational tests to be conducted on the real plant. The overall objective of the test program is to demonstrate that the plant has been constructed as designed, that the systems perform consistently with the plant design, and that activities culminating in operation at full licensed power including initial fuel load, initial criticality, and power increase to full load are performed in a controlled and safe manner. Within this framework, Westinghouse Electric Company and its partner Ansaldo Nucleare S.p.A. have strictly collaborated, being Ansaldo Nucleare S.p.A. in charge of the simulation of some pre-operational tests and supporting Westinghouse in the definition of tests procedures. This paper summarizes the work performed at Ansaldo Nucleare S.p.A. in collaboration with Westinghouse Electric Company for the Core Makeup Tank (CMT) tests, i.e. the CMTs hot recirculation test and the CMTs draindown test. The test procedure for the two

Application of computerized procedure system for AP1000

International Nuclear Information System (INIS)

Wen Fang

2012-01-01

With the tendency of digitalisation in instrumentation and control system of nuclear power plants, AP1000, as well as other advanced PWRs, is provided with the technical basis for the application of CPS (computerized procedure system). This paper makes a brief introduction on CPS construction and function. CPS, as an advanced procedure system, does not only have the function of electronic indication for operation procedures, but also have the ability to monitor plant data, process the data and then present the status of the procedure steps to the reactor operator. In addition, advantages of CPS compared with paper-based procedures and comparison with digital operating system of one M310 plus advanced nuclear power plant are described. Moreover, based on current situation, this paper offers several suggestions on CPS localization for Sanmen AP1000 nuclear power project. Besides, the last part of this paper discusses problems we might meet during the process of CPS localization. (author)

Westinghouse AP 1000 program status

International Nuclear Information System (INIS)

Doehnert, B.

2002-01-01

The project 1000 is presented and features are discussed in the paper. Design maturity is characterized by 1300 man-year / $400 million design and testing effort, more than 12 000 design documents completed; 3D computer model developed. It includes structures, equipment, small / large pipe, cable trays, ducts etc. Licensing Maturity is determined by a very thorough and complete NRC review of AP600; 110 man-year effort (NRC) over 6 years, $30 million; independent, confirmatory plant analysis; independent, confirmatory plant testing (ROSA, OSU); over 7400 questions answered, no open items; over 380 meeting with NRC, 43 meetings with ACRS. NRC Design Certification is issued in December 1999. Reasons for developing AP 1000 and design changes are presented. Economic analysis shows an expectation for payback within 20 years. AP1000 provides 75% power uprate for 15% increment in capital cost. AP1000 meets new plant economic targets in the near term

Spent fuel pool spray cooling system for the AP1000 {sup registered}

Energy Technology Data Exchange (ETDEWEB)

Vujic, Zoran; Sassen, Felix; Tietsch, Wolfgang [Westinghouse Electric Germany GmbH, Mannheim (Germany)

2013-07-01

The AP1000 {sup registered} plant design features multiple, diverse lines of defense to ensure spent fuel cooling can be maintained for Design Basis Events and Beyond Design Basis Accidents (BDBA). The AP1000 {sup registered} plant lines of defense with respect to Spent Fuel Pool (SFP) cooling are as follows: 1. During normal and abnormal conditions, defense-in-depth and duty systems provide highly reliable SFP cooling, supplied by offsite AC power or the onsite Standby Diesel Generators. 2. For unlikely events with extended loss of AC power (i.e. station black-out) and/or loss of heat sink, spent fuel cooling can be still provided indefinitely by: 2a. Passive systems, requiring minimal or no operator actions, sufficient for at least 72 hours under all possible loading conditions. 2b. After 3 days, several different means are provided to continue SFP cooling using installed plant equipment as well as off-site equipment with built-in connections. 3. Even for BDBA with postulated SFP damage and multiple failures in the passive safety-related systems and in the defense-in-depth active systems, the AP1000 {sup registered} SFP Spray System provides an additional line of defense to prevent spent fuel damage. (orig.)

The AP1000R nuclear power plant innovative features for extended station blackout mitigation

International Nuclear Information System (INIS)

Vereb, F.; Winters, J.; Schulz, T.; Cummins, E.; Oriani, L.

2012-01-01

Station Blackout (SBO) is defined as 'a condition wherein a nuclear power plant sustains a loss of all offsite electric power system concurrent with turbine trip and unavailability of all onsite emergency alternating current (AC) power system. Station blackout does not include the loss of available AC power to buses fed by station batteries through inverters or by alternate AC sources as defined in this section, nor does it assume a concurrent single failure or design basis accident...' in accordance with Reference 1. In this paper, the innovative features of the AP1000 plant design are described with their operation in the scenario of an extended station blackout event. General operation of the passive safety systems are described as well as the unique features which allow the AP1000 plant to cope for at least 7 days during station blackout. Points of emphasis will include: - Passive safety system operation during SBO - 'Fail-safe' nature of key passive safety system valves; automatically places the valve in a conservatively safe alignment even in case of multiple failures in all power supply systems, including normal AC and battery backup - Passive Spent Fuel Pool cooling and makeup water supply during SBO - Robustness of AP1000 plant due to the location of key systems, structures and components required for Safe Shutdown - Diverse means of supplying makeup water to the Passive Containment Cooling System (PCS) and the Spent Fuel Pool (SFP) through use of an engineered, safety-related piping interface and portable equipment, as well as with permanently installed onsite ancillary equipment. (authors)

Comparison of CPR1000 and AP1000 rod position indication systems

International Nuclear Information System (INIS)

Lei Qing

2009-01-01

This paper introduces the structure, the function, the digital detection principle of reactor control rod position and monitoring systems in CPR1000 and AP1000, comparing with the characteristics of the system design. The results show that the operation mode and function of AP1000 Rod position indication system are similar to that of CPR1000, but AP1000 rod position system provides higher reliability, and reduces the numbers of containment electrical penetrations and is with better characteristics than that of CPR1000, since it incorporated the redundancy design and data communication. (authors)

The AP1000{sup R} nuclear power plant innovative features for extended station blackout mitigation

Energy Technology Data Exchange (ETDEWEB)

Vereb, F.; Winters, J.; Schulz, T.; Cummins, E.; Oriani, L. [Westinghouse Electric Company LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

2012-07-01

Station Blackout (SBO) is defined as 'a condition wherein a nuclear power plant sustains a loss of all offsite electric power system concurrent with turbine trip and unavailability of all onsite emergency alternating current (AC) power system. Station blackout does not include the loss of available AC power to buses fed by station batteries through inverters or by alternate AC sources as defined in this section, nor does it assume a concurrent single failure or design basis accident...' in accordance with Reference 1. In this paper, the innovative features of the AP1000 plant design are described with their operation in the scenario of an extended station blackout event. General operation of the passive safety systems are described as well as the unique features which allow the AP1000 plant to cope for at least 7 days during station blackout. Points of emphasis will include: - Passive safety system operation during SBO - 'Fail-safe' nature of key passive safety system valves; automatically places the valve in a conservatively safe alignment even in case of multiple failures in all power supply systems, including normal AC and battery backup - Passive Spent Fuel Pool cooling and makeup water supply during SBO - Robustness of AP1000 plant due to the location of key systems, structures and components required for Safe Shutdown - Diverse means of supplying makeup water to the Passive Containment Cooling System (PCS) and the Spent Fuel Pool (SFP) through use of an engineered, safety-related piping interface and portable equipment, as well as with permanently installed onsite ancillary equipment. (authors)

AP1000 design and construction integration

International Nuclear Information System (INIS)

Winters, James W.; Clelland, Jill A.

2004-01-01

Construction costs of commercial nuclear generating plants must be reduced in order to expand the future use of nuclear energy. Two of the drivers of plant construction costs are the cost of financing during the construction duration and the substantial amount of skilled craft labor hours needed on site during construction. The application of information technology (IT) has been used to understand and reduce both of these drivers by establishing parallel construction paths using modules and integrating construction sequence review into the design process. In a program sponsored by EPRI, Westinghouse has modeled the construction of AP1000 in '4D' to show its viability, to improve its logic, to improve the plant design for constructibility and overall to reduce time and risk in the construction schedule. The design of most of AP1000 was constrained to be a duplicate of AP600 except where components required expansion for the higher power level. As a result, the construction schedule for AP1000 is as mature and as robust as that for AP600. Two areas important to the construction of AP1000 did require some design work because they could not remain the same as AP1000. First, the turbine building had to be redesigned to accommodate the larger turbine and its support systems. Again, as much of the AP600 design and philosophy as possible was retained. The building required enlargement and the basemat, foundations, steel structure and structural modules required modification. As concrete, steel, and equipment were defined by the designers, they were matched to the original AP600 turbine building schedule. This forced designers to assemble files to be consistent with building assembly activities and to think about constructibility as they defined the final design. Second, the reinforcement structure within the concrete under and supporting the containment vessel required detail design. Westinghouse was fortunate to have the constructor Obayashi of Japan recommend a detailed

AP1000{sup R} design robustness against extreme external events - Seismic, flooding, and aircraft crash

Energy Technology Data Exchange (ETDEWEB)

Pfister, A.; Goossen, C.; Coogler, K.; Gorgemans, J. [Westinghouse Electric Company LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

2012-07-01

Both the International Atomic Energy Agency (IAEA) and the U.S. Nuclear Regulatory Commission (NRC) require existing and new nuclear power plants to conduct plant assessments to demonstrate the unit's ability to withstand external hazards. The events that occurred at the Fukushima-Dai-ichi nuclear power station demonstrated the importance of designing a nuclear power plant with the ability to protect the plant against extreme external hazards. The innovative design of the AP1000{sup R} nuclear power plant provides unparalleled protection against catastrophic external events which can lead to extensive infrastructure damage and place the plant in an extended abnormal situation. The AP1000 plant is an 1100-MWe pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance and safety. The plant's compact safety related footprint and protection provided by its robust nuclear island structures prevent significant damage to systems, structures, and components required to safely shutdown the plant and maintain core and spent fuel pool cooling and containment integrity following extreme external events. The AP1000 nuclear power plant has been extensively analyzed and reviewed to demonstrate that it's nuclear island design and plant layout provide protection against both design basis and extreme beyond design basis external hazards such as extreme seismic events, external flooding that exceeds the maximum probable flood limit, and malicious aircraft impact. The AP1000 nuclear power plant uses fail safe passive features to mitigate design basis accidents. The passive safety systems are designed to function without safety-grade support systems (such as AC power, component cooling water, service water, compressed air or HVAC). The plant has been designed to protect systems, structures, and components critical to placing the reactor in a safe shutdown condition within the steel

The power of simplification: Operator interface with the AP1000R during design-basis and beyond design-basis events

International Nuclear Information System (INIS)

Williams, M. G.; Mouser, M. R.; Simon, J. B.

2012-01-01

The AP1000 R plant is an 1100-MWe pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, safety and cost. The passive safety features are designed to function without safety-grade support systems such as component cooling water, service water, compressed air or HVAC. The AP1000 passive safety features achieve and maintain safe shutdown in case of a design-basis accident for 72 hours without need for operator action, meeting the expectations provided in the European Utility Requirements and the Utility Requirement Document for passive plants. Limited operator actions may be required to maintain safe conditions in the spent fuel pool (SFP) via passive means. This safety approach therefore minimizes the reliance on operator action for accident mitigation, and this paper examines the operator interaction with the Human-System Interface (HSI) as the severity of an accident increases from an anticipated transient to a design basis accident and finally, to a beyond-design-basis event. The AP1000 Control Room design provides an extremely effective environment for addressing the first 72 hours of design-basis events and transients, providing ease of information dissemination and minimal reliance upon operator actions. Symptom-based procedures including Emergency Operating Procedures (EOPs), Abnormal Operating Procedures (AOPs) and Alarm Response Procedures (ARPs) are used to mitigate design basis transients and accidents. Use of the Computerized Procedure System (CPS) aids the operators during mitigation of the event. The CPS provides cues and direction to the operators as the event progresses. If the event becomes progressively worse or lasts longer than 72 hours, and depending upon the nature of failures that may have occurred, minimal operator actions may be required outside of the control room in areas that have been designed to be accessible using components that have been designed

OLP embedment design method research for AP1000 nuclear plant

International Nuclear Information System (INIS)

Li Cheng; Li Shaoping; Liu Jianwei

2013-01-01

Background: One of the most advanced nuclear power technology, the first AP1000 reactor is under construction in China. Modularization is one of the main characteristics for AP1000 nuclear plant building. Module wall with steel face plate is used instead of reinforced concrete structure wall. A number of OLP embedments need to be installed into the module wall to connect other structures such as pipes, equipment, operation platforms and any other component attached to the module wall. Therefore, the design of embedment is very important in AP1000 structural design. Purpose: A finite element analysis method and tool for embedment design is needed for convenience. Methods: This paper applies the self-developed GTStrudl command template and VBA macro program for embedment capacity calculation and evaluation based on Microsoft Excel to the embedment design. Results: A Microsoft Excel template for embedment design is developed. Conclusions: The analysis method and template brings reasonable results and may provide some help and use for reference for the engineering practice. (authors)

The power of simplification: Operator interface with the AP1000{sup R} during design-basis and beyond design-basis events

Energy Technology Data Exchange (ETDEWEB)

Williams, M. G.; Mouser, M. R.; Simon, J. B. [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

2012-07-01

The AP1000{sup R} plant is an 1100-MWe pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, safety and cost. The passive safety features are designed to function without safety-grade support systems such as component cooling water, service water, compressed air or HVAC. The AP1000 passive safety features achieve and maintain safe shutdown in case of a design-basis accident for 72 hours without need for operator action, meeting the expectations provided in the European Utility Requirements and the Utility Requirement Document for passive plants. Limited operator actions may be required to maintain safe conditions in the spent fuel pool (SFP) via passive means. This safety approach therefore minimizes the reliance on operator action for accident mitigation, and this paper examines the operator interaction with the Human-System Interface (HSI) as the severity of an accident increases from an anticipated transient to a design basis accident and finally, to a beyond-design-basis event. The AP1000 Control Room design provides an extremely effective environment for addressing the first 72 hours of design-basis events and transients, providing ease of information dissemination and minimal reliance upon operator actions. Symptom-based procedures including Emergency Operating Procedures (EOPs), Abnormal Operating Procedures (AOPs) and Alarm Response Procedures (ARPs) are used to mitigate design basis transients and accidents. Use of the Computerized Procedure System (CPS) aids the operators during mitigation of the event. The CPS provides cues and direction to the operators as the event progresses. If the event becomes progressively worse or lasts longer than 72 hours, and depending upon the nature of failures that may have occurred, minimal operator actions may be required outside of the control room in areas that have been designed to be accessible using components that have been

Effect of air condition on AP-1000 containment cooling performance in station black out accident

International Nuclear Information System (INIS)

Hendro Tjahjono

2015-01-01

AP1000 reactor is a nuclear power plant generation III+ 1000 MWe which apply passive cooling concept to anticipate accidents triggered by the extinction of the entire supply of electrical power or Station Black Out (SBO). In the AP1000 reactor, decay heat disposal mechanism conducted passively through the PRHR-IRWST and subsequently forwarded to the reactor containment. Containment externally cooled through natural convection in the air gap and through evaporation cooling water poured on the outer surface of the containment wall. The mechanism of evaporation of water into the air outside is strongly influenced by the conditions of humidity and air temperature. The purpose of this study was to determine the extent of the influence of the air condition on cooling capabilities of the AP1000 containment. The method used is to perform simulations using Matlab-based analytical calculation model capable of estimating the power of heat transferred. The simulation results showed a decrease in power up to 5% for relative humidity rose from 10% to 95%, while the variation of air temperature of 10°C to 40°C, the power will decrease up to 15%. It can be concluded that the effect of air temperature increase is much more significant in lowering the containment cooling ability compared with the increase of humidity. (author)

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

International Nuclear Information System (INIS)

Brettschuh, W.

1999-01-01

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

WGOTHIC analysis of AP1000 passive containment cooling water

International Nuclear Information System (INIS)

Ye Cheng; Wang Yong; Zheng Mingguang; Wang Guodong; Zhang Di; Ni Chenxiao; Wang Minglu

2013-01-01

The WGOTHIC code was used to analyze the influence of the containment cooling water inventory to containment safety for different cases. The results show that if passive containment cooling system fails, the pressure in containment is beyond design limit after 1000 s; if cooling water can't be supplied after 72 h, the pressure in containment is beyond design limit after 0.9 d; if cooling water can't be supplied after 19.6 d, the pressure in containment is beyond design limit but less than the breakdown pressure; if cooling water is supplied for 30 d, the air cooling can remove the decay heat without any aid. It is a reference for making emergency plan and improving containment design. (authors)

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)

Control parameter optimization for AP1000 reactor using Particle Swarm Optimization

International Nuclear Information System (INIS)

Wang, Pengfei; Wan, Jiashuang; Luo, Run; Zhao, Fuyu; Wei, Xinyu

2016-01-01

Highlights: • The PSO algorithm is applied for control parameter optimization of AP1000 reactor. • Key parameters of the MSHIM control system are optimized. • Optimization results are evaluated though simulations and quantitative analysis. - Abstract: The advanced mechanical shim (MSHIM) core control strategy is implemented in the AP1000 reactor for core reactivity and axial power distribution control simultaneously. The MSHIM core control system can provide superior reactor control capabilities via automatic rod control only. This enables the AP1000 to perform power change operations automatically without the soluble boron concentration adjustments. In this paper, the Particle Swarm Optimization (PSO) algorithm has been applied for the parameter optimization of the MSHIM control system to acquire better reactor control performance for AP1000. System requirements such as power control performance, control bank movement and AO control constraints are reflected in the objective function. Dynamic simulations are performed based on an AP1000 reactor simulation platform in each iteration of the optimization process to calculate the fitness values of particles in the swarm. The simulation platform is developed in Matlab/Simulink environment with implementation of a nodal core model and the MSHIM control strategy. Based on the simulation platform, the typical 10% step load decrease transient from 100% to 90% full power is simulated and the objective function used for control parameter tuning is directly incorporated in the simulation results. With successful implementation of the PSO algorithm in the control parameter optimization of AP1000 reactor, four key parameters of the MSHIM control system are optimized. It has been demonstrated by the calculation results that the optimized MSHIM control system parameters can improve the reactor power control capability and reduce the control rod movement without compromising AO control. Therefore, the PSO based optimization

Numerical simulation of AP1000 LBLOCA with SCDAP/RELAP 4.0 code

International Nuclear Information System (INIS)

Xie Heng

2017-01-01

The risk of large-break loss of coolant accident (LBLOCA) is that core will be exposed once the accident occurs, and may cause core damages. New phenomena may occur in LBLOCA due to passive safety injection adopted by AP1000. This paper used SCDAP/RELAP5 4.0 to build the numerical model of AP1000 and double-end guillotine of cold leg is simulated. Reactor coolant system and passive core cooling system were modeled by RELAP5 modular. HEAT STRUCTURE component of RELAP5 was used to simulate the fuel rod. The reflood option in RELAP5 was chosen to be activated or not to study the effect of axial heat conduction. Results show that the axial heat conduction plays an important role in the reflooding phase and can effectively shorten reflood process. An alternative core model is built by SCDAP modular. It is found that the SCDAP model predicts higher maximum peak cladding temperature and longer reflood process than RELAP5 model. Analysis shows that clad oxidation heat plays a key role in the reflood. From the simulation results, it can be concluded that the cladding will keep intact and fission product will not be released from fuel to coolant in LBLOCA. (author)

The Westinghouse AP1000®: Passive, Proven Technology to Meet European Energy Demands

International Nuclear Information System (INIS)

Haspel, N.

2015-01-01

Even though several years ago nuclear power was merely considered to be an “optimistic future assessment”, the world-wide renaissance of nuclear power has become reality! The economical and climate-friendly nuclear power generation is internationally regarded to be in an evident upturn. The 435 nuclear power plants in operation worldwide are being modernized and the capacity is increased continuously. Furthermore, to date, 42 power plants are under construction, another 81 are already being applied for and or definitely planned. The global total net capacity out of nuclear power will increase accordingly in the upcoming years from currently 372 to more than 500 GWe, which presents an increase of more than one third. Westinghouse’s contribution hereto is considerable: At the present time, 4 power plants of the series AP1000 ® are under construction. To begin with, 2 units each are under construction at the Chinese sites Sanmen and Haiyang, another 4 per site are being planned. In the USA, Westinghouse has been contracted with a Engineering, Procurement and Construction (EPC) project for a total of 4 power plant units at the Vogtle and V.C. Summer. Also in Europe, the plans to construct new plants are meanwhile very specific and many countries have formally established the marginal conditions for new nuclear projects. The AP1000 ® , with its medium output capacity, is ideally positioned for many markets and can – as a twin unit – also cover large capacity demands. At the present time, Westinghouse, with its AP1000 ® , participates in the so-called GDA (Generic Design Assessment) process in Great Britain, where the British regulatory authorities conduct an assessment and evaluation of the safety aspects of this plant design in a defined multilevel process. The successful conclusion of this process ultimately leads to a “Design Acceptance Confirmation”, which will basically make the construction of the plant in Great Britain possible. (author)

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

International Nuclear Information System (INIS)

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

2003-01-01

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

Current status of generation III nuclear power and assessment of AP1000 developed by Westinghouse

International Nuclear Information System (INIS)

Zhang Mingchang

2005-01-01

In order to make greater contributions to the environment, new nuclear power systems will be needed to meet the increase of electricity demand and to replace plants to be decommissioned. A series of new designs, so called Generation III and Generation III +, are being developed to ensure their deployment in a Near-Term Deployment Road-map in US by 2010 and in Europe by 2015. The AP1000, developed by Westinghouse, is a two-loop 1000 MWe PWR with passive safety features and extensive simplifications to enhance its competitiveness in cost and tariff. It is the first Generation III + plant receiving the Final Design Approval by the US NRC. This paper briefly describes AP1000 design features and technical specifications, and presents a more detailed design evaluation with reference to relevant literatures. Both the opportunity and challenges for nuclear power development in China during the first decade of the 21 st century in a historic transition from Gen II to Gen III are analyzed. The key is to balance risks and benefits if the first AP1000 to be settled down in China. (author)

Essence and characteristics of the Westinghouse technology AP1000

International Nuclear Information System (INIS)

Llovet, Ricardo

2014-01-01

The AP1000 nuclear power plant can place the reactor in a Safe Shutdown Condition within the first 72 hours of a Station Blackout, without the use of AC power or operator action •With some operator action after 3 days, the AP1000 nuclear power plant continues to maintain reactor core cooling and Spent Fuel Pool cooling indefinitely •The AP1000 nuclear power plant has superior coping capabilities as well as significantly reduced risk for core damage

On integration and innovation of sino-foreign safety culture in Haiyang AP1000 Project

International Nuclear Information System (INIS)

Li Ruipu; Song Fengwei

2010-01-01

The undergoing Haiyang Nuclear Power Plant is not only introducing the top-advanced AP1000 nuclear technology, but also the mature HSE management system from U.S.A. It's very important for both sides to communicate, comprehend and acculturation of both different culture. After over 1 year discussion and practice, the experts of Westinghouse Consortium and Chinese HSE engineers have established an distinctive safety culture of AP1000 Project initially, demonstrating the followings: Exemplary actions of the expat experts and the SNPTC leaders, the high level standard HSE procedures, HSE audit, various training, HSE inspection all-around, the safety performance assessment for prospective index, JHA/JSA , emergency system, humanism rewards and punishment etc.. Haiyang SPMO has made Three-Step master plan for AP1000 project HSE Routine by analysis the site problems and the difference between Chinese and American, that is, from 2008 to 2020, when nuclear power achieve to independent, safety culture of Haiyang AP1000 will change from 'dependent' to 'independent', until the last 'interdependent'. (authors)

Westinghouse plans global new builds for AP1000

International Nuclear Information System (INIS)

Mitev, Lubomir

2014-01-01

Interview with Danny Roderick, Westinghouse Electric Company, President and Chief Executive Officer since September 2012, about perspectives and future plans for AP1000 new build worldwide. Within three to four years there wille be 'shovels in the ground' for three new AP1000 reactors in the UK, as well as new units in China and Bulgaria. Four AP1000 reactors are under construction in the United States at Vogtle and VC Summer, and soon at Turkey Point. Additionally Danny Roderick spoke about the acquisition of NuGen, technology transfer, the influence of the Ukraine crises on the nuclear market in East Europe and the future need for more nuclear worldwide and in the UK and Bulgaria.

Westinghouse plans global new builds for AP1000

Energy Technology Data Exchange (ETDEWEB)

Mitev, Lubomir [NucNet, Brussels (Belgium)

2014-10-15

Interview with Danny Roderick, Westinghouse Electric Company, President and Chief Executive Officer since September 2012, about perspectives and future plans for AP1000 new build worldwide. Within three to four years there wille be 'shovels in the ground' for three new AP1000 reactors in the UK, as well as new units in China and Bulgaria. Four AP1000 reactors are under construction in the United States at Vogtle and VC Summer, and soon at Turkey Point. Additionally Danny Roderick spoke about the acquisition of NuGen, technology transfer, the influence of the Ukraine crises on the nuclear market in East Europe and the future need for more nuclear worldwide and in the UK and Bulgaria.

Characteristics and design improvement of AP1000 automatic depressurization system

International Nuclear Information System (INIS)

Jin Fei

2012-01-01

Automatic depressurization system, as a specialty of AP1000 Design, enhances capability of mitigating design basis accidents for plant. Advancement of the system is discussed by comparing with traditional PWR design and analyzing system functions, such as depressurizing and venting. System design improvement during China Project performance is also described. At the end, suggestions for the system in China Project are listed. (author)

Containment integrity analysis for the (W) advanced AP600

International Nuclear Information System (INIS)

Gagnon, A.F.; Howe, K.S.

1989-01-01

This paper reports that since 1987, Westinghouse has been performing containment cooling analyses in support of the Advanced AP600 plant design. This program was intended to verify the feasibility of the passive containment cooling system features of the AP600 design. To support this design, containment analyses of the AP600 containment for a large break LOCA and a large Steam Line Break were performed. The transient results indicate the feasibility of the passive containment design by demonstrating the capability to remove sufficient heat to limit containment atmosphere conditions to within acceptable limits following these postulated accidents. These results also indicate that the PCCS can reduce containment pressure to less than one-quarter design pressure at 24 hours following the most severe accident scenario thereby minimizing containment leakage concerns

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

International Nuclear Information System (INIS)

Moraleda Lopez, A.

1997-01-01

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

Discussion of QA grading for AP1000 NP plant

International Nuclear Information System (INIS)

Luo Shuiyun; Zhang Qingchuan

2012-01-01

The grading method of quality assurance for the following AP1000 project is presented based on the Westinghouse classification principle, referring to the classification method of the AP1000 self-reliance supporting project and considering the factors of classification, which can meet the requirements of domestic nuclear safety regulation and standard of the QA classification. (authors)

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

International Nuclear Information System (INIS)

2016-09-01

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

AP1000 core design with 50% MOX loading

International Nuclear Information System (INIS)

Fetterman, Robert J.

2009-01-01

The European uility requirements (EUR) document states that the next generation European passive plant (EPP) reactor core design shall be optimized for UO 2 fuel assemblies, with provisions made to allow for up to 50% mixed-oxide (MOX) fuel assemblies. The use of MOX in the core design will have significant impacts on key physics parameters and safety analysis assumptions. Furthermore, the MOX fuel rod design must also consider fuel performance criterion important to maintaining the integrity of the fuel rod over its intended lifetime. The purpose of this paper is to demonstrate that the AP1000 is capable of complying with the EUR requirement for MOX utilization without significant changes to the design of the plant. The analyses documented within will compare a 100% UO 2 core design and a mixed MOX/UO 2 core design, discussing relevant results related to reactivity management, power margin and fuel rod performance

AP1000 core design with 50% MOX loading

International Nuclear Information System (INIS)

Fetterman, Robert J.

2008-01-01

The European Utility Requirements (EUR) document states that the next generation European Passive Plant (EPP) reactor core design shall be optimized for UO 2 fuel assemblies, with provisions made to allow for up to 50% mixed-oxide (MOX) fuel assemblies. The use of MOX in the core design will have significant impacts on key physics parameters and safety analysis assumptions. Furthermore, the MOX fuel rod design must also consider fuel performance criterion important to maintaining the integrity of the fuel rod over its intended lifetime. The purpose of this paper is to demonstrate that the AP1000 is capable of complying with the EUR requirement for MOX utilization without significant changes to the design of the plant. The analyses documented within will compare a 100% UO 2 core and a mixed MOX / UO 2 core design, discussing relevant results related to reactivity management, power margin and fuel rod performance. (authors)

AP1000 Shield Building Dynamic Response for Different Water Levels of PCCWST Subjected to Seismic Loading considering FSI

Directory of Open Access Journals (Sweden)

Daogang Lu

2015-01-01

Full Text Available Huge water storage tank on the top of many buildings may affect the safety of the structure caused by fluid-structure interaction (FSI under the earthquake. AP1000 passive containment cooling system water storage tank (PCCWST placed at the top of shield building is a key component to ensure the safety of nuclear facilities. Under seismic loading, water will impact the wall of PCCWST, which may pose a threat to the integrity of the shield building. In the present study, an FE model of AP1000 shield building is built for the modal and transient seismic analysis considering the FSI. Six different water levels in PCCWST were discussed by comparing the modal frequency, seismic acceleration response, and von Mises stress distribution. The results show the maximum von Mises stress emerges at the joint of shield building roof and water around the air inlet. However, the maximum von Mises stress is below the yield strength of reinforced concrete. The results may provide a reference for design of the AP1000 and CAP1400 in the future.

Analysis of an accident type sbloca in reactor contention AP1000 with 8.0 Gothic code; Analisis de un accidente tipo Sbloca en la contencion del reactor AP1000 con el codigo Gothic 8.0

Energy Technology Data Exchange (ETDEWEB)

Goni, Z.; Jimenez Varas, G.; Fernandez, K.; Queral, C.; Montero, J.

2016-08-01

The analysis is based on the simulation of a Small Break Loss-of-Coolant-Accident in the AP1000 nuclear reactor using a Gothic 8.0 tri dimensional model created in the Science and Technology Group of Nuclear Fision Advanced Systems of the UPM. The SBLOCA has been simulated with TRACE 5.0 code. The main purpose of this work is the study of the thermo-hydraulic behaviour of the AP1000 containment during a SBLOCA. The transients simulated reveal close results to the realistic behaviour in case of an accident with similar characteristics. The pressure and temperature evolution enables the identification of the accident phases from the RCS point of view. Compared to the licensing calculations included in the AP1000 Safety Analysis, it has been proved that the average pressure and temperature evolution is similar, yet lower than the licensing calculations. However, the temperature and inventory distribution are significantly heterogeneous. (Author)

HSE management for AP1000 nuclear plant construction in EPC mode

International Nuclear Information System (INIS)

He Xiaogang; Wei Zhong

2010-01-01

As a new nuclear type, AP1000 will become the development direction of Chinese nuclear project. EPC General Contract mode is favored by nuclear owners both at home and abroad. Therefore, there is necessity for studying HSE management system and method suitable for AP1000 nuclear plant construction (ANPC) based on combination of AP1000 construction characters in EPC mode. This can not only ensure safety for ANPC but also positively promote national nuclear power development. For this reason, based on site HSE management of the first AP1000 nuclear plant under construction, HSE management system and method for ANPC in EPC mode was proposed after analysis of the character of EPC mode and ANPC character. It is hoped that it will be helpful for safe construction for ANPC. (authors)

AP1000 core design with 50% MOX loading

Energy Technology Data Exchange (ETDEWEB)

Fetterman, Robert J. [Westinghouse Electric Company, LLC, Pittsburgh, PA (United States)

2008-07-01

The European Utility Requirements (EUR) document states that the next generation European Passive Plant (EPP) reactor core design shall be optimized for UO{sub 2} fuel assemblies, with provisions made to allow for up to 50% mixed-oxide (MOX) fuel assemblies. The use of MOX in the core design will have significant impacts on key physics parameters and safety analysis assumptions. Furthermore, the MOX fuel rod design must also consider fuel performance criterion important to maintaining the integrity of the fuel rod over its intended lifetime. The purpose of this paper is to demonstrate that the AP1000 is capable of complying with the EUR requirement for MOX utilization without significant changes to the design of the plant. The analyses documented within will compare a 100% UO{sub 2} core and a mixed MOX / UO{sub 2} core design, discussing relevant results related to reactivity management, power margin and fuel rod performance. (authors)

AP1000 core design with 50% MOX loading

Energy Technology Data Exchange (ETDEWEB)

Fetterman, Robert J. [Westinghouse Electric Company, LLC, Pittsburgh, PA (United States)], E-mail: fetterrj@westinghouse.com

2009-04-15

The European uility requirements (EUR) document states that the next generation European passive plant (EPP) reactor core design shall be optimized for UO{sub 2} fuel assemblies, with provisions made to allow for up to 50% mixed-oxide (MOX) fuel assemblies. The use of MOX in the core design will have significant impacts on key physics parameters and safety analysis assumptions. Furthermore, the MOX fuel rod design must also consider fuel performance criterion important to maintaining the integrity of the fuel rod over its intended lifetime. The purpose of this paper is to demonstrate that the AP1000 is capable of complying with the EUR requirement for MOX utilization without significant changes to the design of the plant. The analyses documented within will compare a 100% UO{sub 2} core design and a mixed MOX/UO{sub 2} core design, discussing relevant results related to reactivity management, power margin and fuel rod performance.

The modularization construction of piping system installation in AP1000 plant

International Nuclear Information System (INIS)

Lu Song; Wang Yuan; Wei Junming

2012-01-01

Modularization construction is the main technique used in AP1000 plants, the piping Modularization installation will impact directly to the module construction as the important part of the Modularization construction. After the piping system has took the modularization design in AP1000 plants, some installation works of piping system has moved from the site to fabrication shop. With improving the construction quality and minimizing the time frame of project, the critical paths can be optimized. This paper has analyzed the risk and challenge that met during the modularization construction period of piping systems though introducing the characteristic of modularization construction for AP1000 piping systems, and get construction experiences from the First AP1000 plants in the world, then it will be the firmly basics for the wide application of modularization construction in the future. (authors)

Dynamic Analysis of AP1000 Shield Building Considering Fluid and Structure Interaction Effects

Directory of Open Access Journals (Sweden)

Qiang Xu

2016-02-01

Full Text Available The shield building of AP1000 was designed to protect the steel containment vessel of the nuclear reactor. Therefore, the safety and integrity must be ensured during the plant life in any conditions such as an earthquake. The aim of this paper is to study the effect of water in the water tank on the response of the AP1000 shield building when subjected to three-dimensional seismic ground acceleration. The smoothed particle hydrodynamics method (SPH and finite element method (FEM coupling method is used to numerically simulate the fluid and structure interaction (FSI between water in the water tank and the AP1000 shield building. Then the grid convergence of FEM and SPH for the AP1000 shield building is analyzed. Next the modal analysis of the AP1000 shield building with various water levels (WLs in the water tank is taken. Meanwhile, the pressure due to sloshing and oscillation of the water in the gravity drain water tank is studied. The influences of the height of water in the water tank on the time history of acceleration of the AP1000 shield building are discussed, as well as the distributions of amplification, acceleration, displacement, and stresses of the AP1000 shield building. Research on the relationship between the WLs in the water tank and the response spectrums of the structure are also taken. The results show that the high WL in the water tank can limit the vibration of the AP1000 shield building and can more efficiently dissipate the kinetic energy of the AP1000 shield building by fluid-structure interaction.

Advances in passive cooling design and performance analysis

International Nuclear Information System (INIS)

Woodcock, J.

1994-01-01

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

Comparison and analysis for item classifications between AP1000 and traditional PWR

International Nuclear Information System (INIS)

Luo Shuiyun; Liu Xiaoyan

2012-01-01

The comparison and analysis for the safety classification, seismic category, code classification and QA classification between AP1000 and traditional PWR were presented. The safety could be guaranteed and the construction and manufacture costs could be cut down since all sorts of AP1000 classifications. It is suggested that the QA classification and the QA requirements correspond to the national conditions should be drafted in the process of AP1000 domestication. (authors)

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-03-01

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

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

International Nuclear Information System (INIS)

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

2016-01-01

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

Influence of performance improvement of AP1000 nuclear island main equipment forging on manufacturing

International Nuclear Information System (INIS)

Liu Zhiying

2013-01-01

In order to comply with the 60-year design life of an AP1000 nuclear power station, higher strength and ductility requirements have been made on AP1000 nuclear island SG forgings than on CPR1000 nuclear island main equipment. In addition, bigger size of AP1000 nuclear island SG forgings increases the difficulty of manufacturing them. Insufficient recognition of these changes may cause unstable quality of forgings and possible quality problems in follow-up welding procedure. On the basis of comparison and analysis of AP1000 nuclear island SG forgings and CPR1000 nuclear island forgings, this thesis suggests clear directions for the actions we need to take. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-03-15

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

Westinghouse AP1000 Electrical Generation Costs - Meeting Marketplace Requirements

International Nuclear Information System (INIS)

Paulson, C. Keith

2002-01-01

completion. These plants are the Westinghouse advanced passive designs - AP600 and AP1000 - both of which have verifiable engineering design packages that are more than 50 percent complete. (author)

Standardized safety management of AP1000 nuclear power plant

International Nuclear Information System (INIS)

Li Xingwen; Cao Zhiqiang; Cong Jiuyuan

2011-01-01

In 2002, China published and implemented the Law of the People's Republic of China on Work Safety and promulgated a series of guidelines and policies, which strengthened the safety management supervision. Standardization of safety, as another important step on safety supervision, comes after safety assesment and safety production licensing system, is also a permanent solution. Standardization of safety is a strategic, long term and fundamental work, which is also the basic access to achieving scientific safety management and increasing the inherent safety of an enterprise. Haiyang AP1000 nuclear power plant, adopting the modularized, 'open-top' and parallel construction means, overturned the traditional construction theory of installation work comes after the civil work and greatly shorten the construction period. At the same time, the notable increase of oversize module transportation and lifting and parallel construction raises higher demands for safety management. This article combines the characteristics and difficulties of safety management for Haiyang AP1000 nuclear power plant, puts forward ideas and methods for standardized safety management, and could also serve as reference to the safety management for other AP1000 projects. (authors)

Westinghouse Advances in Passive Plant Safety

International Nuclear Information System (INIS)

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

1993-01-01

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

Characteristics and application study of AP1000 NPPs equipment reliability classification method

International Nuclear Information System (INIS)

Guan Gao

2013-01-01

AP1000 nuclear power plant applies an integrated approach to establish equipment reliability classification, which includes probabilistic risk assessment technique, maintenance rule administrative, power production reliability classification and functional equipment group bounding method, and eventually classify equipment reliability into 4 levels. This classification process and result are very different from classical RCM and streamlined RCM. It studied the characteristic of AP1000 equipment reliability classification approach, considered that equipment reliability classification should effectively support maintenance strategy development and work process control, recommended to use a combined RCM method to establish the future equipment reliability program of AP1000 nuclear power plants. (authors)

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)

The study on neutron and photon distribution of AP1000 reactor by MCNP code

International Nuclear Information System (INIS)

Chen Defeng; Shen Mingqi

2014-01-01

The core and reactor structural of AP1000 was modeled by the MCNP calculation program which is based on the Monte Carlo method in this paper, the neutron and photon distribution of AP1000 reactor core was calculated by the conditions of reactor critical. The results show that the AP1000 reactor neutron and photon distribution is in accordance with the critical design of PWR. (authors)

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)

Study on the operation mode for indigenization and standardization of AP1000 technology

International Nuclear Information System (INIS)

Gao Zhihu; Cheng Huiping

2014-01-01

This paper describes the importance and necessity of developing standardized AP1000 technology, and analyzes the problems faced and measures to be taken. The operation mode, known as the Committee of AP1000 Standardized Design and Innovation Management, was first put forward in China. And the paper also discusses how to arouse the enthusiasm of the owners, designers, manufactures, as well as construction, installation, regulation and other parties, how to exploit the advantage of whole industry to promote the localization and standardization of AP1000 technology. (authors)

Safety features and research needs of westinghouse advanced reactors

International Nuclear Information System (INIS)

Carelli, M.D.; Winters, J.W.; Cummins, W.E.; Bruschi, H.J.

2002-01-01

The three Westinghouse advanced reactors - AP600, AP1000 and IRIS - are at different levels of readiness. AP600 has received a Design Certification, its larger size version AP1000 is currently in the design certification process and IRIS has just completed its conceptual design and will initiate soon a licensing pre-application. The safety features of the passive designs AP600/AP1000 are presented, followed by the features of the more revolutionary IRIS, a small size modular integral reactor. A discussion of the IRIS safety by design approach is given. The AP600/AP1000 design certification is backed by completed testing and development which is summarized, together with a research program currently in progress which will extend AP600 severe accident test data to AP1000 conditions. While IRIS will of course rely on applicable AP600/1000 data, a very extensive testing campaign is being planned to address all the unique aspects of its design. Finally, IRIS plans to use a risk-informed approach in its licensing process. (authors)

The investigation of Passive Accident Mitigation Scheme for advanced PWR NPP

International Nuclear Information System (INIS)

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

2015-01-01

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

Comparison Study of Water Demineralization System for the OPR 1000 and AP 1000 Nuclear Power Plant

International Nuclear Information System (INIS)

Dedy Priambodo; Siti Alimah; Erlan Dewita

2009-01-01

OPR 1000 adopts demineralization method based on ion exchanger resin and AP 1000 adopt the method that based on Reverse Osmosis (RO)-Electrodeionization (EDI). The Ion exchange process is a reversible chemical reaction of a solution and an insoluble solid. Ion exchanger use resin as polluter ions capture and will be regenerated after its saturated. RO is method using pressure to force a solution through a membrane, retaining the solute on one side and allowing the pure solvent to pass to the other side. Whereas, EDI is a combination of ion exchange and electrodialysis. The ions is taken by ion exchange resin, and then it is discharged utilizing electric potential difference. Due to water splitting phenomena in EDI, make resin will never be saturated, so the RO-EDI process is water demineralization system that use little chemical, more simple installation, capable to maintain demineralization water product quality and environmental friendly. Thereby, The RO-EDI water demineralization system is more advance then ion exchange technology. (author)

Research on quality assurance classification methodology for domestic AP1000 nuclear power projects

International Nuclear Information System (INIS)

Bai Jinhua; Jiang Huijie; Li Jingyan

2012-01-01

To meet the quality assurance classification requirements of domestic nuclear safety codes and standards, this paper analyzes the quality assurance classification methodology of domestic AP1000 nuclear power projects at present, and proposes the quality assurance classification methodology for subsequent AP1000 nuclear power projects. (authors)

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

International Nuclear Information System (INIS)

Wider, H.

2005-01-01

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

Analysis of risk management during AP1000 equipment technology transfer and localization

International Nuclear Information System (INIS)

Gao Yongjun; Guan Rui

2009-01-01

This article analyzes the risk factors existing in AP1000 equipment technology transfer and localization process by describing the invitation for bid, tender evaluation and contract negotiation process of the third-generation nuclear power plant technology introduction project of China and discusses the classification, evaluation and analysis methods of risks, and puts forward some referential suggestions for the successful introduction of equipment technology for AP1000 nuclear project. (authors)

AP1000 shield building: a constructability challenge

International Nuclear Information System (INIS)

Di Giuseppe, Giovanni; Bonanno, Domenico

2010-01-01

The AP1000 Shield Building, an enhanced structure which surrounds the containment vessel, consists of standard Reinforced Concrete (RC) and composite Steel and Concrete (SC) construction. In the SC module the surface steel plates, (with attached shear studs and angles) filled with concrete, act as the steel reinforcement in concrete. This is a relatively new design technology that required the appropriate use of structural codes, supplemented with information from applicable tests on similar composite steel and concrete construction. Being a newer design concept, existing codes do not provide explicit guidance on SC construction so a review of literature and test data on composite structures similar to AP1000 shield building was done in order to confirm the technical basis for the design. The SC walls, air inlet structure and roof of the Shield Building will be constructed using modular construction practices and then transported to site and lifted into place. These modules, working also as permanent form-work, will be filled with high strength Self- Consolidating Concrete. (SCC) This paper provides a focused and integrated presentation of the enhanced shield building design methodology, testing, constructability and inspection. (authors)

The passive safety systems of the Swr 1000

International Nuclear Information System (INIS)

Neumann, D.

2001-01-01

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

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.

Westinghouse and nuclear renaissance. The Westinghouse AP1000 - a technology solution for Slovakia

International Nuclear Information System (INIS)

Kirst, M.

2009-01-01

The Westinghouse AP1000 nuclear reactor design has been chosen by both China and the United States as the preferred technology in their new reactor programs. With four reactors in China and six in the United States under contract, in addition to the only Generation III+ design with NRC certification as well as the European Utility Requirements certification, the AP1000 has both a strong global customer base and regulatory certainty to facilitate its adoption in the Slovak Republic. (author)

A Calculation Method for the Sloshing Impact Pressure Imposed on the Roof of a Passive Water Storage Tank of AP1000

Directory of Open Access Journals (Sweden)

Daogang Lu

2016-01-01

Full Text Available There is a large water storage tank installed at the top of containment of AP1000, which can supply the passive cooling. In the extreme condition, sloshing of the free surface in the tank may impact on the roof under long-period earthquake. For the safety assessment of structure, it is necessary to calculate the impact pressure caused by water sloshing. Since the behavior of sloshing impacted on the roof is involved into a strong nonlinear phenomenon, it is a little difficult to calculate such pressure by theoretical or numerical method currently. But it is applicable to calculate the height of sloshing in a tank without roof. In the present paper, a simplified method was proposed to calculate the impact pressure using the sloshing wave height, in which we first marked the position of the height of roof, then produced sloshing in the tank without roof and recorded the maximum wave height, and finally regarded approximately the difference between maximum wave height and roof height as the impact pressure head. We also designed an experiment to verify this method. The experimental result showed that this method overpredicted the impact pressure with a certain error of no more than 35%. By the experiment, we conclude that this method is conservative and applicable for the engineering design.

The AP600 advanced simplified nuclear power plant. Results of the test program and progress made toward final design approval

International Nuclear Information System (INIS)

Bruschi, H.J.

1996-01-01

At the 1994 Pacific Basin Conference, Mr. Bruschi presented a paper describing the AP600, Westinghouse's advanced light water reactor design with passive safety features. Since then, a rigorous test program was completed and AP600 became the most thoroughly tested advanced reactor system design in history. Westinghouse is now well on its way toward receiving Final Design Approval from the U.S. Nuclear Regulatory Commission for AP600. In this paper, the results of the test program will be discussed and an update on prospects for building the plant will be covered. (author)

The AP600 advanced simplified nuclear power plant. Results of the test program and progress made toward final design approval

Energy Technology Data Exchange (ETDEWEB)

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

1996-10-01

At the 1994 Pacific Basin Conference, Mr. Bruschi presented a paper describing the AP600, Westinghouse`s advanced light water reactor design with passive safety features. Since then, a rigorous test program was completed and AP600 became the most thoroughly tested advanced reactor system design in history. Westinghouse is now well on its way toward receiving Final Design Approval from the U.S. Nuclear Regulatory Commission for AP600. In this paper, the results of the test program will be discussed and an update on prospects for building the plant will be covered. (author)

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

International Nuclear Information System (INIS)

Carsten, Pasler

2007-01-01

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

Preliminary safety evaluation for CSR1000 with passive safety system

International Nuclear Information System (INIS)

Wu, Pan; Gou, Junli; Shan, Jianqiang; Zhang, Bo; Li, Xiang

2014-01-01

Highlights: • The basic information of a Chinese SCWR concept CSR1000 is introduced. • An innovative passive safety system is proposed for CSR1000. • 6 Transients and 3 accidents are analysed with system code SCTRAN. • The passive safety systems greatly mitigate the consequences of these incidents. • The inherent safety of CSR1000 is enhanced. - Abstract: This paper describes the preliminary safety analysis of the Chinese Supercritical water cooled Reactor (CSR1000), which is proposed by Nuclear Power Institute of China (NPIC). The two-pass core design applied to CSR1000 decreases the fuel cladding temperature and flattens the power distribution of the core at normal operation condition. Each fuel assembly is made up of four sub-assemblies with downward-flow water rods, which is favorable to the core cooling during abnormal conditions due to the large water inventory of the water rods. Additionally, a passive safety system is proposed for CSR1000 to increase the safety reliability at abnormal conditions. In this paper, accidents of “pump seizure”, “loss of coolant flow accidents (LOFA)”, “core depressurization”, as well as some typical transients are analysed with code SCTRAN, which is a one-dimensional safety analysis code for SCWRs. The results indicate that the maximum cladding surface temperatures (MCST), which is the most important safety criterion, of the both passes in the mentioned incidents are all below the safety criterion by a large margin. The sensitivity analyses of the delay time of RCPs trip in “loss of offsite power” and the delay time of RMT actuation in “loss of coolant flowrate” were also included in this paper. The analyses have shown that the core design of CSR1000 is feasible and the proposed passive safety system is capable of mitigating the consequences of the selected abnormalities

ANALISIS SKENARIO KEGAGALAN SISTEM UNTUK MENENTUKAN PROBABILITAS KECELAKAAN PARAH AP1000

Directory of Open Access Journals (Sweden)

D.T. Sony Tjahyani

2014-03-01

Full Text Available Kejadian Fukushima telah menunjukkan bahwa kecelakaan parah dapat terjadi, maka dari itu sangatlah penting untuk menganalisis tingkat keselamatan pada reaktor daya. Berdasarkan rekomendasi expert mission IAEA setelah kejadian Fukushima, perlu dilakukan upaya untuk meminimalisasi terjadinya kecelakaan parah yaitu dengan melakukan proses pendinginan yang maksimal. Dalam konsep keselamatan fasilitas nuklir, khususnya reaktor daya telah diterapkan konsep keselamatan berlapis (Defence in Depth, DiD. Konsep keselamatan tersebut terdiri atas 5 level pertahanan yang bertujuan mencegah dan mengurangi lepasan produk fisi ke masyarakat dan lingkungan pada saat reaktor daya mengalami kecelakaan. Dalam reaktor telah didesain sistem atau tindakan yang mempunyai fungsi untuk mengatasi setiap level tersebut. Tujuan dari analisis ini adalah menentukan probabilitas kecelakaan parah dengan melakukan skenario kegagalan sistem dalam proses pendinginan di reaktor. Sebagai obyek analisis adalah reaktor daya AP1000, karena jenis reaktor ini sedang banyak dibangun saat ini. Skenario dilakukan dengan mengasumsikan beberapa kombinasi kegagalan sistem yang termasuk dalam DiD level 2 dan 3. Kegagalan sistem kemudian dianalisis dengan menggunakan analisis pohon kegagalan berdasarkan perangkat lunak SAPHIRE ver. 6.76. Dari analisis didapatkan probabilitas gagal dari kelompok sistem DiD level 2 dan 3 pada AP1000 masih di bawah batas kriteria dari IAEA yaitu lebih kecil dari 10-2, serta probabilitas kecelakaan parah didapatkan sebesar 6,17 x 10-10. Berdasarkan analisis ini disimpulkan bahwa AP1000 mempunyai tingkat keselamatan yang cukup tinggi, karena melalui skenario kegagalan sistem didapatkan probabilitas kecelakaan parah yang sangat kecil.   ABSTRACT Fukushima accident has shown that severe accident could be occurred, therefore it is important to analyze safety level of nuclear power plants. Based on the recommendations of IAEA expert mission after the Fukushima accident

Spanish program of advanced Nuclear Power Plants

International Nuclear Information System (INIS)

Marco, M.; Redon, R.

1993-01-01

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

Key developments of the EP1000 design

International Nuclear Information System (INIS)

Noviello, L.

1999-01-01

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 Phase I of the European Passive Plant program involved the evaluation of the Westinghouse 600 MWe AP600 and 1000 MWe Simplified Pressurized Water Reactor (SPWR) designs against the European Utility Requirements (EUR), and when necessary, the investigation of possible modifications to achieve compliance with the EUR. In Phase 1 of the program, which has been completed in 1996, the following major tasks were accomplished: The impacts of the European Utility Requirements (EUR) on the Westinghouse nuclear island design were evaluated. A 1000 MWe passive plant reference design (EP1000) was developed which conforms to the EUR and is expected to be licensable in Europe. With respect to the NSSS and containment, the EP1000 reference design closely follows those of the Westinghouse SPWR design, while the AP600 design has been taken as the basis for the design of the auxiliary systems. Extensive design and testing efforts have been made for the AP600 and SPWR during the respective multi-year programs. While the results of these programs have been and will continue to be utilised, at the maximum extent, to minimise the work to be performed on the EP1000 design, the compliance with EUR is a key design requirement for the EP1000 The ultimate objective of Phase 2 of the program is to develop design details and perform supporting analyses to produce a Safety Case Report (SCR) for submittal to European Safety Authorities. The first part of Phase 2, hereafter referred as Phase 2A, started at the beginning of 1997 and will be completed at the end of 1998. Scope of this phase of the program is to develop the design modifications of important systems and structures so to comply with the

The PANDA tests for the SWR 1000 passive containment cooling system

International Nuclear Information System (INIS)

Dreier, J.; Aubert, C.; Huggenberger, M.; Strassberger, H.J.; Yadigaroglu, G.

1999-01-01

Since 1992, Siemens has been developing the SWR 1000, a new boiling water reactor with passive safety features. This development has been performed in close co-operation with the German nuclear utilities and with support from various European partners. Within the European Union sponsored project 'BWR R+D Cluster for Innovative Passive Safety Systems' and a bilateral contract between Siemens and the Paul Scherrer Institute, the passive containment cooling system of the SWR 1000 design has been investigated in the large-scale PANDA test facility at the Paul Scherrer Institute. A series of six tests were performed to simulate transients selected to cover a range of failure assumptions and accident severity, including core heat up and hydrogen generation. The results graphically demonstrate the self regulating character of the passive heat removal systems and their effectiveness, even under severe load, in limiting the containment pressurisation. Some tentative conclusions for the SWR 1000 are drawn, to be established by detailed analyses of the data, to support models and codes for application to plant transients. (author)

The PANDA tests for the SWR 1000 passive containment cooling system

International Nuclear Information System (INIS)

Dreier, J.; Aubert, C.; Huggenberger, M.; Strassberger, H.J.; Meseth, J.; Yadigaroglu, G.

1999-01-01

Since 1992, Siemens has been developing the SWR 1000, a new boiling water reactor with passive safety features. This development has been performed in close co-operation with the German nuclear utilities and with support from various European partners. Within the European Union sponsored project 'BWR R and D Cluster for Innovative Passive Safety Systems' and a bilateral contract between Siemens and the Paul Scherrer Institute, the passive containment cooling system of the SWR 1000 design has been investigated in the large-scale PANDA test facility at the Paul Scherrer Institute. A series of six tests were performed to simulate transients selected to cover a range of failure assumptions and accident severity, including core heat up and hydrogen generation. The results graphically demonstrate the self regulating character of the passive heat removal systems and their effectiveness, even under severe load, in limiting the containment pressurisation. Some tentative conclusions for the SWR1000 are drawn, to be established by detailed analyses of the data, to support models and codes for application to plant transients. (author)

Advanced Photon Source (APS)

Data.gov (United States)

Federal Laboratory Consortium — The Advanced Photon Source (APS) at the U.S. Department of Energy's Argonne National Laboratoryprovides this nation's (in fact, this hemisphere's) brightest storage...

77 FR 74696 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on AP-1000...

Science.gov (United States)

2012-12-17

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on AP-1000; Notice of Meeting The ACRS Subcommittee on AP-1000 will hold a meeting on January 18, 2013, Room T-2B1, 11545 Rockville Pike, Rockville, Maryland. The entire meeting will be open to public...

High-inertia hermetically sealed main coolant pump for next generation passive nuclear power plants

International Nuclear Information System (INIS)

Kujawski, Joseph M.; Nair, Bala R.; Vijuk, Ronald P.

2003-01-01

The main coolant pump for the Westinghouse AP1000 advanced passive nuclear power plant represents a significant scale-up in power, flow capacity, and physical size from its predecessor designed for the smaller AP600 power plant. More importantly, the AP1000 pump incorporates several innovative features that contribute to improved efficiency, operational reliability, and plant safety. The features include an internals design which provides the highest hydraulic efficiency achieved in commercial nuclear power plant applications. Another feature is the use of a distributed inertial mass system in the rotating assembly to develop the high rotational inertia to meet the extended system flow coastdown requirement for core heat removal in the event of loss of power to the pumps. This advanced canned motor pump also incorporates the latest development in higher operating voltage, providing plant designers with the ability to eliminate plant transformers and operate directly on the site electrical bus in many cases. The salient features of the pump design and performance data are presented in this paper. (author)

Finite mixture models for sensitivity analysis of thermal hydraulic codes for passive safety systems analysis

Energy Technology Data Exchange (ETDEWEB)

Di Maio, Francesco, E-mail: francesco.dimaio@polimi.it [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Nicola, Giancarlo [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Zio, Enrico [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Chair on System Science and Energetic Challenge Fondation EDF, Ecole Centrale Paris and Supelec, Paris (France); Yu, Yu [School of Nuclear Science and Engineering, North China Electric Power University, 102206 Beijing (China)

2015-08-15

Highlights: • Uncertainties of TH codes affect the system failure probability quantification. • We present Finite Mixture Models (FMMs) for sensitivity analysis of TH codes. • FMMs approximate the pdf of the output of a TH code with a limited number of simulations. • The approach is tested on a Passive Containment Cooling System of an AP1000 reactor. • The novel approach overcomes the results of a standard variance decomposition method. - Abstract: For safety analysis of Nuclear Power Plants (NPPs), Best Estimate (BE) Thermal Hydraulic (TH) codes are used to predict system response in normal and accidental conditions. The assessment of the uncertainties of TH codes is a critical issue for system failure probability quantification. In this paper, we consider passive safety systems of advanced NPPs and present a novel approach of Sensitivity Analysis (SA). The approach is based on Finite Mixture Models (FMMs) to approximate the probability density function (i.e., the uncertainty) of the output of the passive safety system TH code with a limited number of simulations. We propose a novel Sensitivity Analysis (SA) method for keeping the computational cost low: an Expectation Maximization (EM) algorithm is used to calculate the saliency of the TH code input variables for identifying those that most affect the system functional failure. The novel approach is compared with a standard variance decomposition method on a case study considering a Passive Containment Cooling System (PCCS) of an Advanced Pressurized reactor AP1000.

AP1000 station blackout study with and without depressurization using RELAP5/SCDAPSIM

Energy Technology Data Exchange (ETDEWEB)

Trivedi, A.K. [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India); Allison, C. [Innovative Systems Software Idaho Falls, ID 83406 (United States); Khanna, A., E-mail: akhanna@iitk.ac.in [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India); Munshi, P. [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India)

2016-10-15

Highlights: • A representative RELAP5/SCDAPSIM model of AP1000 has been developed. • Core is modeled using SCDAP. • A SBO for the AP1000 has been simulated for high pressure (no depressurization) and low pressure (depressurization). • Significant differences in the damage progression have been observed for the two cases. • Results also reinforced the fact that surge line fails before vessel failure in case of high pressure scenario. - Abstract: Severe accidents like TMI-2, Chernobyl, Fukushima made it inevitable to analyze station blackout (SBO) for all the old as well as new designs although it is not a regulatory requirement in most of the countries. For such improbable accidents, a SBO for the AP1000 using RELAP5/SCDAPSIM has been simulated. Many improvements have been made in fuel damage progression models of SCDAP after the Fukushima accident which are now being tested for the new reactor designs. AP1000 is a 2-loop pressurized water reactor (PWR) with all the emergency core cooling systems based on natural circulation. Its core design is very similar to 3-loop PWR with 157 fuel assemblies. The primary circuit pumps, pressurizer and steam generators (with necessary secondary side) are modeled using RELAP5. The core has been divided into 20 axial nodes and 6 radial rings; the corresponding six groups of assemblies have been modeled as six pipe components with proportionate flow area. Fuel assemblies are modeled using SCDAP fuel and control components. SCDAP has 2d-heat conduction and radiative heat transfer, oxidation and complete severe fuel damage progression models. The final input deck achieved all the steady state thermal hydraulic conditions comparable to the design control document of AP1000. To quantify the core behavior, under unavailability of all safety systems, various time profiles for SBO simulations @ high pressure and low pressure have been compared. This analysis has been performed for 102% (3468 MWt) of the rated core power. The

AP1000 - update on projects in US and China

Energy Technology Data Exchange (ETDEWEB)

Godfrey, M. [Westinghouse Electric Company, Cranberry Township, Pennsy lvania (United States)

2012-07-01

Westinghouse is the only company solely focused on commercial nuclear technology. Westinghouse business is based on four product lines regionally divided: nuclear power plants, nuclear fuel, nuclear services and nuclear automation. The AP1000 is the technology of choice for more than half of the new plants identified in the US. Westinghouse has the only certified Generation III+ technology by the US Nuclear Regulatory Commission (NRC). The first Generation III+ plants are under construction in China and the US.

AP1000 - update on projects in US and China

International Nuclear Information System (INIS)

Godfrey, M.

2012-01-01

Westinghouse is the only company solely focused on commercial nuclear technology. Westinghouse business is based on four product lines regionally divided: nuclear power plants, nuclear fuel, nuclear services and nuclear automation. The AP1000 is the technology of choice for more than half of the new plants identified in the US. Westinghouse has the only certified Generation III+ technology by the US Nuclear Regulatory Commission (NRC). The first Generation III+ plants are under construction in China and the US.

Study of impact of the AP1000{sup Registered-Sign} reactor vessel upper internals design on fuel performance

Energy Technology Data Exchange (ETDEWEB)

Xu Yiban; Conner, Michael; Yuan Kun; Dzodzo, Milorad B.; Karoutas, Zeses; Beltz, Steven A.; Ray, Sumit; Bissett, Teresa A. [Westinghouse Electric Company, Cranberry Township, PA 16066 (United States); Chieng, Ching-Chang, E-mail: cchieng@ess.nthu.edu.tw [National Tsing Hua University, Hsinchu 30043, Taiwan (China); Kao, Min-Tsung; Wu, Chung-Yun [National Tsing Hua University, Hsinchu 30043, Taiwan (China)

2012-11-15

One aspect of the AP1000{sup Registered-Sign} reactor design is the reduction in the number of major components and simplification in manufacturing. One design change relative to current Westinghouse reactors of similar size is the reduction in the number of reactor vessel outlet nozzles/hot legs leaving the upper plenum from three to two. With regard to fuel performance, this design difference creates a different flow field in the AP1000 reactor vessel upper plenum (the region above the core). The flow exiting core and entering the upper plenum must turn 90 Degree-Sign , flow laterally through the upper plenum around support structures, and exit through one of the two outlet nozzles. While the flow in the top of the core is mostly axial, there is some lateral flow component as the core flow reacts to the flow field and pressure distribution in the upper plenum. The pressure distribution in the upper plenum varies laterally depending upon various factors including the proximity to the outlet nozzles. To determine how the lateral flow in the top of the AP1000 core compares to current Westinghouse reactors, a computational fluid dynamics (CFD) model of the flow in the upper portion of the AP1000 reactor vessel including the top region of the core, the upper plenum, the reactor vessel outlet nozzles, and a portion of the hot legs was created. Due to geometric symmetry, the computational domain was reduced to a quarter (from the top view) that includes Vulgar-Fraction-One-Quarter of the top of the core, Vulgar-Fraction-One-Quarter of the upper plenum, and Vulgar-Fraction-One-Half of an outlet nozzle. Results from this model include predicted velocity fields and pressure distributions throughout the model domain. The flow patterns inside and around guide tubes clearly demonstrate the influence of lateral flow due to the presence of the outlet nozzles. From these results, comparisons of AP1000 flow versus current Westinghouse plants were performed. Field performance

AP600 - an ALWR conceptual design

International Nuclear Information System (INIS)

Bruce, R.A.; Vijuk, R.P.

1988-01-01

The Electric Power Research Institute is spearheading an effort to develop utility requirements for the Advanced Light Water Reactor (ALWR) plants which will become the next generation nuclear power plants for the U.S. This EPRI ALWR Program involves utilities, the U.S. Department of Energy, the U.S. Nuclear Regulatory Commission, and various industry suppliers. The ALWR Program is aimed at ALWR plants which incorporate step improvements in safety, reliability, operability and power generation costs. As part of the ALWR efforts, a Westinghouse team is conducting conceptual design development of a PWR plant design called the AP600, reflecting advanced passive safety features and the chosen 600 MWe plant output. The AP600 conceptual design provides significant improvements while employing proven component technology. This paper describes the basic reactor and primary coolant system features, the passive safety system features, and plant arrangement/construction features of AP600

Reactor coolant system hydrostatic test and risk analysis for the first AP1000 unit

International Nuclear Information System (INIS)

Cao Hongjun; Yan Xiuping

2013-01-01

The cold hydrostatic test scheme of the primary coolant circuit, of the first AP1000 unit was described. Based on the up-stream design documents, standard specifications and design technical requirements, the select principle of test boundary was identified. The design requirements for water quality, pressure, temperature and temporary hydro-test pump were proposed. A reasonable argument for heating and pressurization rate, and cooling and depressurization rate was proposed. The possible problems and risks during the hydrostatic test were analyzed. This test scheme can provide guidance for the revisions and implementations of the follow-up test procedures. It is a good reference for hydrostatic tests of AP1000 units in the future in China. (authors)

An analysis of AP600 design features

International Nuclear Information System (INIS)

Park, Jong Kyoon; Jang, Moon Heui; Hwang, Yung Dong

1994-01-01

In the aspect of engineering, passive safety system concept has improved the safety degree of nuclear power plant. Therefore, the objective of this study is to check on the possibility of the capacity upgrade of nuclear power plant in the case of adopting the passive safety system concept of AP 600. The characteristics of AP 600 are the advanced functions in ECCS, heat removal of containment building and residual heat removal under the passive safety system concept. The result of this study will become the basic data of capacity upgrade of nuclear power plant and will be widely used in second year project. (Author)

An analysis of AP600 design features

Energy Technology Data Exchange (ETDEWEB)

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

1994-01-01

In the aspect of engineering, passive safety system concept has improved the safety degree of nuclear power plant. Therefore, the objective of this study is to check on the possibility of the capacity upgrade of nuclear power plant in the case of adopting the passive safety system concept of AP 600. The characteristics of AP 600 are the advanced functions in ECCS, heat removal of containment building and residual heat removal under the passive safety system concept. The result of this study will become the basic data of capacity upgrade of nuclear power plant and will be widely used in second year project. (Author).

Information on the Advanced Plant Experiment (APEX) Test Facility

International Nuclear Information System (INIS)

Smith, Curtis Lee

2015-01-01

The purpose of this report provides information related to the design of the Oregon State University Advanced Plant Experiment (APEX) test facility. Information provided in this report have been pulled from the following information sources: Reference 1: R. Nourgaliev and et.al, 'Summary Report on NGSAC (Next-Generation Safety Analysis Code) Development and Testing,' Idaho National Laboratory, 2011. Note that this is report has not been released as an external report. Reference 2: O. Stevens, Characterization of the Advanced Plant Experiment (APEX) Passive Residual Heat Removal System Heat Exchanger, Master Thesis, June 1996. Reference 3: J. Reyes, Jr., Q. Wu, and J. King, Jr., Scaling Assessment for the Design of the OSU APEX-1000 Test Facility, OSU-APEX-03001 (Rev. 0), May 2003. Reference 4: J. Reyes et al, Final Report of the NRC AP600 Research Conducted at Oregon State University, NUREG/CR-6641, July 1999. Reference 5: K. Welter et al, APEX-1000 Confirmatory Testing to Support AP1000 Design Certification (non-proprietary), NUREG-1826, August 2005.

Simulation of advanced accumulator and its application in CPR1000 LBLOCA analysis

International Nuclear Information System (INIS)

Hu, Hongwei; Shan, Jianqiang; Gou, Junli; Cao, Jianhua; Shen, Yonggang; Fu, Xiangang

2014-01-01

Highlights: • The analysis model was developed for advanced accumulator. • The sensitivity analysis of each key parameter was performed. • The LBLOCA was analyzed for the CPR1000 with advanced accumulator. • The analysis shows that advanced accumulator can improve CPR1000 safety performance. - Abstract: The advanced accumulator is designed to improve the safety and reliability of CPR1000 by providing a small injection flow to keep the reactor core in flooded condition. Thus, the core still stays in a cooling state without the intervention of low pressure safety injection and the startup grace time of the low pressure safety injection pump can be greatly extended. A new model for the advanced accumulator, which is based on the basic conservation equations, is developed and incorporated into RELAP5/MOD 3.3. The simulation of the advanced accumulator can be carried out and results show that the behavior of the advanced accumulator satisfied its primary design target. There is a large flow in the advanced accumulator at the initial stage. When the accumulator water level is lower than the stand pipe, a vortex appears in the damper, which results in a large pressure drop and a small flow. And then the sensitivity analysis is performed and the major factors which affected the flow rate of the advanced accumulator were obtained, including the damper diameter, the initial volume ratio of the water and the nitrogen and the diameter ratio of the standpipe and the small pipe. Additionally, the primary coolant loop cold leg double-ended guillotine break LBLOCA in CPR1000 with advanced accumulator is analyzed. The results show that the criterion for maximum cladding temperature limit (1477 K) (NRC, 1992) can be met ever with 200 s after the startup of the low pressure safety injection. From this point of view, passive advanced accumulator can strive a longer grace time for LPSI. Thus the reliability, safety and economy of the reactor system can be improved

Feasibility analysis of AP1000 wireless communication system and selection of technical solutions

International Nuclear Information System (INIS)

Zhao Xin

2012-01-01

This article expatiates the rationality and feasibility of AP1000 nuclear power plant adopts wireless communication system as the first choice in routine and emergency operations, compares and analysed. 5 major wireless communication technology solutions, and introduces the Wi-Fi based wireless communication system architecture. (author)

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

Science.gov (United States)

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

2018-02-01

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

Experimental research of liquid entrainment through ADS-4 in AP1000

International Nuclear Information System (INIS)

Meng, Zhaoming; Dong, Bo; Wang, Laishun; Fu, Xiaoliang; Tian, Wenxi; Yang, Yanhua; Su, Guanghui

2014-01-01

Highlights: • We performed experimental research of liquid entrainment through ADS-4 in AP1000. • Effect of various factors on entrainment at T-junction was conducted. • Visualization research was conducted to make entrainment mechanism clear. - Abstract: In this study, based on a T-junction that consists of Automatic Depressurization System Stage Four (ADS-4) and hot leg in an AP1000 plant, a small-scale experimental research on entrainment at a T-junction was performed. This study mainly focused on the effect of various factors on entrainment, such as the effect of branch size, branch shape and liquid crossflow. The flow pattern map was plotted from the experimental data, and the visualization research indicated that the entrainment phenomena through a large size branch were apparently different from that through a small branch. Three entrainment phenomena were observed in the studies, two entrainment mechanisms could be found in the stratified flow regime entrainment area, the existence of branch contributed to generating intermittent flow in the horizontal main pipe, and the backflow region was observable in the vicinity of a large size branch inlet. Also, experimental research showed that downstream of the branch of T-junction had an important effect on the onset entrainment, and liquid crossflow did not seem to affect the onset entrainment

AP: A Critical Examination of the Advanced Placement Program

Science.gov (United States)

Sadler, Philip M.; Sonnert, Gerhard; Tai, Robert; Klopfenstein, Kirstin

2016-01-01

The Advanced Placement (AP) program was created to enhance the experience of gifted students as they transition from high school to college. "AP: A Critical Examination of the Advanced Placement Program," edited by Philip M. Sadler, Gerhard Sonnert, Robert Tai, and Kirstin Klopfenstein (2010, Harvard Education Press), questions the…

Effects of the Application of the New Nuclear Data Library ENDF/B to the Criticality Analysis of AP1000

Science.gov (United States)

Kuntoro, Iman; Sembiring, T. M.; Susilo, Jati; Deswandri; Sunaryo, G. R.

2018-02-01

Calculations of criticality of the AP1000 core due to the use of new edition of nuclear data library namely ENDF/B-VII and ENDF/B-VII.1 have been done. This work is aimed to know the accuracy of ENDF/B-VII.1 compared to ENDF/B-VII and ENDF/B-VI.8. in determining the criticality parameter of AP1000. Analysis ws imposed to core at cold zero power (CZP) conditions. The calculations have been carried out by means of MCNP computer code for 3 dimension geometry. The results show that criticality parameter namely effective multiplication factor of the AP1000 core are higher than that ones resulted from ENDF/B-VI.8 with relative differences of 0.39% for application of ENDF/B-VII and of 0.34% for application of ENDF/B-VII.1.

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

International Nuclear Information System (INIS)

Perrotta, Jose Augusto

1999-01-01

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

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

International Nuclear Information System (INIS)

Forsberg, C.W.

1989-01-01

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

Operating experience review for the AP1000 plant

International Nuclear Information System (INIS)

Chaney, T. E.; Lipner, M. H.

2006-01-01

Westinghouse is performing an update to the Operating Experience Review (OER) Report for the AP1000 project to account for operating experience since December 1996. Significant Operating Experience Reports, Significant Event Reports, Significant Event Notifications, Operations and Maintenance Reminders, Topical Reports, Event Analysis Reports and Licensee Event Reports were researched for pertinent input to the update. As a part of the OER, Westinghouse has also conducted operator interviews and observations during simulated plant operations and after operating events. The main purpose of the OER is to identify Human Factors Engineering (HFE) related safety issues from existing operating plant experience and to ensure that these issues are addressed in the new design. The issues and lessons learned regarding operating experience provide a basis for improving the plant design. (authors)

Development of Fuel ROd Behavior Analysis code (FROBA) and its application to AP1000

International Nuclear Information System (INIS)

Yu, Hongxing; Tian, Wenxi; Yang, Zhen; SU, G.H.; Qiu, Suizheng

2012-01-01

Highlights: ► A Fuel ROd Behavior Analysis code (FROBA) has been developed. ► The effects irradiation and burnup has been considered in FROBA. ► The comparison with INL’s results shows a good agreement. ► The FROBA code was applied to AP1000. ► Peak fuel temperature, gap width, hoop strain, etc. were obtained. -- Abstract: The reliable prediction of nuclear fuel rod behavior is of great importance for safety evaluation of nuclear reactors. In the present study, a thermo-mechanical coupling code FROBA (Fuel ROd Behavior Analysis) has been independently developed with consideration of irradiation and burnup effects. The thermodynamic, geometrical and mechanical behaviors have been predicted and were compared with the results obtained by Idaho National Laboratory to validate the reliability and accuracy of the FROBA code. The validated code was applied to analyze the fuel behavior of AP1000 at different burnup levels. The thermal results show that the predicted peak fuel temperature experiences three stages in the fuel lifetime. The mechanical results indicate that hoop strain at high power is greater than that at low power, which means that gap closure phenomenon will occur earlier at high power rates. The maximum cladding stress meets the requirement of yield strength limitation in the entire fuel lifetime. All results show that there are enough safety margins for fuel rod behavior of AP1000 at rated operation conditions. The FROBA code is expected to be applied to deal with more complicated fuel rod scenarios after some modifications.

Grain size control method for the nozzles of AP1000 primary coolant pipes

Energy Technology Data Exchange (ETDEWEB)

Wang, Shenglong [State Key Laboratory for Advanced Metals and Materials, University of Science & Technology Beijing, Beijing 100083 (China); Sun, Yanhui [Collaborative Innovation Center of Steel Technology, University of Science & Technology Beijing, Beijing 100083 (China); Yang, Bin, E-mail: byang@ustb.edu.cn [State Key Laboratory for Advanced Metals and Materials, University of Science & Technology Beijing, Beijing 100083 (China); Collaborative Innovation Center of Steel Technology, University of Science & Technology Beijing, Beijing 100083 (China); Zhang, Mingxian [State Key Laboratory for Advanced Metals and Materials, University of Science & Technology Beijing, Beijing 100083 (China)

2017-04-01

Highlights: • Design a new forging technology for AP1000 primary coolant pipe. • Method combining FEM and scale-down experiments is adopted. • The grain size and distribution in simulation and experiment are consistent. • Get optimal forging parameters for production guiding. - Abstract: AP1000 primary coolant pipe is made of 316LN austenitic stainless steel. It is a large special-shaped pipe manufactured by integral forging technology. Owing to non-uniform temperature and deformation during forging, coarse grains often occur in the boss sections of the pipe especially in the nozzles’ parts. In the present study, a new forging technology was proposed to control the grain size. The finite element method was used to optimize the forging speed and friction coefficient, then the scale-down experiments were performed for comparison. The forging speed is suggested to be less than 20 mm/s, and effective lubricants should be used to decrease the friction coefficient. The errors of the grain size between the experiment and simulation are less than 20%.

Grain size control method for the nozzles of AP1000 primary coolant pipes

International Nuclear Information System (INIS)

Wang, Shenglong; Sun, Yanhui; Yang, Bin; Zhang, Mingxian

2017-01-01

Highlights: • Design a new forging technology for AP1000 primary coolant pipe. • Method combining FEM and scale-down experiments is adopted. • The grain size and distribution in simulation and experiment are consistent. • Get optimal forging parameters for production guiding. - Abstract: AP1000 primary coolant pipe is made of 316LN austenitic stainless steel. It is a large special-shaped pipe manufactured by integral forging technology. Owing to non-uniform temperature and deformation during forging, coarse grains often occur in the boss sections of the pipe especially in the nozzles’ parts. In the present study, a new forging technology was proposed to control the grain size. The finite element method was used to optimize the forging speed and friction coefficient, then the scale-down experiments were performed for comparison. The forging speed is suggested to be less than 20 mm/s, and effective lubricants should be used to decrease the friction coefficient. The errors of the grain size between the experiment and simulation are less than 20%.

Comparative study on aerosol removal by natural processes in containment in severe accident for AP1000 reactor

International Nuclear Information System (INIS)

Sun, Xiaohui; Cao, Xinrong; Shi, Xingwei; Yan, Jin

2017-01-01

Highlights: • Characteristics of aerosol distribution in containment are obtained. • Aerosol removal by natural processes is comparative studied by two methods. • Traditional rapid assessment method is conservative and can be applied in AP1000 reactor. - Abstract: Focusing on aerosol removal by naturally occurring processes in containment in severe accident for AP1000, integral severe accident code MELCOR and rapid assessment method mentioned in NUREG/CR-6189 are utilized to study aerosol removal by natural processes, respectively. Three typical severe accidents, induced by large break loss of coolant accident (LBLOCA), small break loss of coolant accident (SBLOCA) and steam generator tube rupture (SGTR), respectively, are selected for the study. The results obtained by two methods were further compared in the following several aspects: efficiency of aerosol removal by natural processes, peak time of aerosol suspended in containment atmosphere, peak amount of aerosol suspended in containment atmosphere, time when aerosol removal efficiency by natural processes is up to 99.9%. It was further concluded that results obtained by rapid assessment with shorter calculation process are more conservative. The analysis results provide reference to assessment method selection of severe accident source term for AP1000 nuclear emergency.

Natural circulation and stratification in the various passive safety systems of the SWR 1000

International Nuclear Information System (INIS)

Meseth, J.

2002-01-01

In some of the passive safety systems of Siemens' SWR 1000 boiling water reactor (i.e. the emergency condensers and containment cooling condensers), natural circulation is the main effect on both the primary and secondary sides by which optimum system efficiency is achieved. Other passive safety systems of the SWR 1000 require natural circulation on the secondary side only (condensation of steam discharged by the safety and relief valves; cooling of the Reactor Pressure Vessel (RPV) by flooding from the outside in case of core melt), while still other systems require stratification to be effective (i.e. the passive pressure pulse transmitters and steam-driven scram tanks). Complex natural circulation and stratification can take place simultaneously if fluids with different densities are enclosed in a single volume (in a core melt accident, for example, the nitrogen, steam and hydrogen in the containment). Related problems and the solutions thereto planned for the SWR 1000 are reported from the designer's viewpoint. (author)

An approach for assessing ALWR passive safety system reliability

International Nuclear Information System (INIS)

Hake, T.M.

1991-01-01

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

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

International Nuclear Information System (INIS)

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

2013-01-01

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

Study on radioactive release of gaseous and liquid effluents during normal operation of AP1000

International Nuclear Information System (INIS)

Gong Quan; Zhou Jing; Liu Yu

2014-01-01

The gaseous and liquid radioactive releases of pressurized water reactors plant during normal operation are an important content of environmental impact assessment and play a significant role in the design of nuclear power plant. According to the design characters of AP1OOO radioactive waste management system and the study on the calculation method and the release pathways, the calculation model of the gaseous and liquid radioactive releases during normal operation for AP1OOO are established. Base on the established calculation model and the design parameters of AP1000, the expected value of gaseous and liquid radioactive releases of AP1OOO is calculated. The results of calculation are compared with the limits in GB 6249-2011 and explain the adder that is included tu account for anticipated operational occurrences, providing a reference for environmental impact assessment of pressurized water reactor. (authors)

Use of phenomena identification and ranking (PIRT) process in research related to design certification of the AP600 advanced passive light water reactor (LWR)

International Nuclear Information System (INIS)

Wilson, G.E.; Fletcher, C.D.; Eltawila, F.

1996-01-01

The AP600 LWR is a new advanced passive design that has been submitted to the USNRC for design certification. Within the certification process the USNRC will perform selected system thermal hydraulic response audit studies to help confirm parts of the vendor's safety analysis submittal. Because of certain innovative design features of the safety systems, new experimental data and related advances in the system thermal hydraulic analysis computer code are being developed by the USNRC. The PIRT process is being used to focus the experimental and analytical work to obtain a sufficient and cost effective research effort. The objective of this paper is to describe the application and most significant results of the PIRT process, including several innovative features needed in the application to accommodate the short design certification schedule. The short design certification schedule has required that many aspects of the USNRC experimental and analytical research be performed in parallel, rather than in series as was normal for currently operating LWRS. This has required development and use of management techniques that focus and integrate the various diverse parts of the research. The original PIRTs were based on inexact knowledge of an evolving reactor design, and concentrated on the new passive features of the design. Subsequently, the PIRTs have evolved in two more stages as the design became more firm and experimental and analytical data became available. A fourth and final stage is planned and in progress to complete the PIRT development. The PIRTs existing at the end of each development stage have been used to guide the experimental program, scaling analyses and code development supporting the audit studies

Fuel burn-up distribution and transuranic nuclide contents produced at the first cycle operation of AP1000

International Nuclear Information System (INIS)

Jati Susilo; Jupiter Sitorus Pane

2016-01-01

AP1000 reactor core was designed with nominal power of 1154 MWe (3415 MWth), operated within life time of 60 years and cycle length of 18 months. For the first cycle, the AP1000 core uses three kinds of UO 2 enrichment, they are 2.35 w/o, 3.40 w/o and 4.45 w/o. Absorber materials such as ZrB 2 , Pyrex and Boron solution are used to compensate the excess reactivity at the beginning of cycle. In the core, U-235 fuels are burned by fission reaction and produce energy, fission products and new neutron. Because of the U-238 neutron absorption reaction, the high level radioactive waste of heavy nuclide transuranic such as Pu, Am, Cm and Np are also generated. They have a very long half life. The purpose of this study is to evaluate the result of fuel burn-up distribution and heavy nuclide transuranic contents produced by AP1000 at the end of first cycle operation (EOFC). Calculation of ¼ part of the AP1000 core in the 2 dimensional model has been done using SRAC2006 code with the module of COREBN/HIST. The input data called the table of macroscopic cross section, is calculated using module of PIJ. The result shows that the maximum fuel assembly (FA) burn-up is 27.04 GWD/MTU, that is still lower than allowed maximum burn-up of 62 GWD/MTU. Fuel loading position at the center/middle of the core will produce bigger burn-up and transuranic nuclide than one at the edges the of the core. The use of IFBA fuel just give a small effect to lessen the fuel burn-up and transuranic nuclide production. (author)

COMMIX analysis of AP-600 Passive Containment Cooling System

International Nuclear Information System (INIS)

Chang, J.F.C.; Chien, T.H.; Ding, J.; Sun, J.G.; Sha, W.T.

1992-01-01

COMMIX modeling and basic concepts that relate components, i.e., containment, water film cooling, and natural draft air flow systems. of the AP-600 Passive Containment Cooling System are discussed. The critical safety issues during a postulated accident have been identified as (1) maintaining the liquid film outside the steel containment vessel, (2) ensuring the natural convection in the air annulus. and (3) quantifying both heat and mass transfer accurately for the system. The lack of appropriate heat and mass transfer models in the present analysis is addressed. and additional assessment and validation of the proposed models is proposed

76 FR 73720 - Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Westinghouse AP1000...

Science.gov (United States)

2011-11-29

... NUCLEAR REGULATORY COMMISSION [NRC-2011-0272] Knowledge and Abilities Catalog for Nuclear Power...) is issuing for public comment a draft NUREG, NUREG-2103, Revision 0, ``Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Westinghouse AP1000 Pressurized-Water Reactors. DATES: Submit...

Advanced APS Impacts on Vehicle Payloads

Science.gov (United States)

Schneider, Steven J.; Reed, Brian D.

1989-01-01

Advanced auxiliary propulsion system (APS) technology has the potential to both, increase the payload capability of earth-to-orbit (ETO) vehicles by reducing APS propellant mass, and simplify ground operations and logistics by reducing the number of fluids on the vehicle and eliminating toxic, corrosive propellants. The impact of integrated cryogenic APS on vehicle payloads is addressed. In this system, launch propulsion system residuals are scavenged from integral launch propulsion tanks for use in the APS. Sufficient propellant is preloaded into the APS to return to earth with margin and noncomplete scavenging assumed. No propellant conditioning is required by the APS, but ambient heat soak is accommodated. High temperature rocket materials enable the use of the unconditioned hydrogen/oxygen in the APS and are estimated to give APS rockets specific impulse of up to about 444 sec. The payload benefits are quantified and compared with an uprated monomethyl hydrazine/nitrogen tetroxide system in a conservative fashion, by assuming a 25.5 percent weight growth for the hydrogen/oxygen system and a 0 percent weight growth for the uprated system. The combination and scavenging and high performance gives payload impacts which are highly mission specific. A payload benefit of 861 kg (1898 lbm) was estimated for a Space Station Freedom rendezvous mission and 2099 kg (4626 lbm) for a sortie mission, with payload impacts varying with the amount of launch propulsion residual propellants. Missions without liquid propellant scavenging were estimated to have payload penalties, however, operational benefits were still possible.

Review of the proposed materials of construction for the SBWR and AP600 advanced reactors

International Nuclear Information System (INIS)

Diercks, D.R.; Shack, W.J.; Chung, H.M.; Kassner, T.F.

1994-06-01

Two advanced light water reactor (LWR) concepts, namely the General Electric Simplified Boiling Water Reactor (SBWR) and the Westinghouse Advanced Passive 600 MWe Reactor (AP600), were reviewed in detail by Argonne National Laboratory. The objectives of these reviews were to (a) evaluate proposed advanced-reactor designs and the materials of construction for the safety systems, (b) identify all aging and environmentally related degradation mechanisms for the materials of construction, and (c) evaluate from the safety viewpoint the suitability of the proposed materials for the design application. Safety-related systems selected for review for these two LWRs included (a) reactor pressure vessel, (b) control rod drive system and reactor internals, (c) coolant pressure boundary, (d) engineered safety systems, (e) steam generators (AP600 only), (f) turbines, and (g) fuel storage and handling system. In addition, the use of cobalt-based alloys in these plants was reviewed. The selected materials for both reactors were generally sound, and no major selection errors were found. It was apparent that considerable thought had been given to the materials selection process, making use of lessons learned from previous LWR experience. The review resulted in the suggestion of alternate an possibly better materials choices in a number of cases, and several potential problem areas have been cited

PENGEMBANGAN MODEL UNTUK SIMULASI KESELAMATAN REAKTOR PWR 1000 MWe GENERASI III+ MENGGUNAKAN PROGRAM KOMPUTER RELAP5

Directory of Open Access Journals (Sweden)

Andi Sofrany Ekariansyah

2015-04-01

rinci. Kata kunci: pemodelan, Generasi III+, RELAP5.   Westinghouse’s AP1000 reactor design is the first Generation III+ nuclear power reactor to receive final design approval from the U.S. Nuclear Regulatory Commission (NRC. Currently, the China’s utilities are starting construction several units of AP1000 on two selected sites for scheduled operation in 2013–2015. The AP1000, based on proven technology of Westinghouse-designed PWR with enhancement on the passive safety system, could be considered to be built in Indonesia referring to the requirements of government regulation No. 43/2006 regarding the Nuclear Reactor Licensing. To be accepted by the regulation agency, the design needs to be verified by independent Technical Support Organization (TSO, which can be done using RELAP5 computer code as accident analyses. Currently, NPP safety accident analysis is performed for PWR 1000 MWe of generation II or conventional type. Considering that nowadays references about the technology of AP1000 that includes passive safety technology has been available and assessed, a modeling activity used for future accident analyzes is introduced. Method for developing the model refers to IAEA guide consisting of plant data collection, engineering data and input deck development, and verification and validation of input data. The model developed should be considered preliminary but has been generally representing the AP1000 systems as the basic model. The model has been verified and validated by comparing thermalhidraulic parameter responses with design data in references with ± 13% deviation except for core pressure drop with 13% lower than design. As a basic model, the input deck is ready for further development by integrating safety system, protection system and control system model specified for AP1000 for purposes of safety simulation in detailed way. Keywords: Modeling, Generation III+ , RELAP5.

RELAP5/SCDAPSIM model development for AP1000 and verification for large break LOCA

Energy Technology Data Exchange (ETDEWEB)

Trivedi, A.K. [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India); Allison, C. [Innovative Systems Software, Idaho Falls, ID 83406 (United States); Khanna, A., E-mail: akhanna@iitk.ac.in [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India); Munshi, P. [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India)

2016-08-15

Highlights: • RELAP5/SCDAPSIM model of AP1000 has been developed. • Analysis involves a LBLOCA (double ended guillotine break) study in cold leg. • Results are compared with those of WCOBRA–TRAC and TRACE. • Concluded that PCT does not violate the safety criteria of 1477 K. - Abstract: The AP1000 is a Westinghouse 2-loop pressurized water reactor (PWR) with all emergency core cooling systems based on natural circulation. Its core design is very similar to a 3-loop PWR with 157 fuel assemblies. Westinghouse has reported their results of the safety analysis in its design control document (DCD) for a large break loss of coolant accident (LOCA) using WCOBRA/TRAC and for a small break LOCA using NOTRUMP. The current study involves the development of a representative RELAP5/SCDASIM model for AP1000 based on publically available data and its verification for a double ended cold leg (DECL) break in one of the cold legs in the loop containing core makeup tanks (CMT). The calculated RELAP5/SCDAPSIM results have been compared to publically available WCOBRA–TRAC and TRACE results of DECL break in AP1000. The objective of this study is to benchmark thermal hydraulic model for later severe accident analyses using the 2D SCDAP fuel rod component in place of the RELAP5 heat structures which currently represent the fuel rods. Results from this comparison provides sufficient confidence in the model which will be used for further studies such as a station blackout. The primary circuit pumps, pressurizer and steam generators (including the necessary secondary side) are modeled using RELAP5 components following all the necessary recommendations for nodalization. The core has been divided into 6 radial rings and 10 axial nodes. For the RELAP5 thermal hydraulic calculation, the six groups of fuel assemblies have been modeled as pipe components with equivalent flow areas. The fuel including the gap and cladding is modeled as a 1d heat structure. The final input deck achieved

Screening key parameters related to passive system performance based on Analytic Hierarchy Process

International Nuclear Information System (INIS)

Ma, Guohang; Yu, Yu; Huang, Xiong; Peng, Yuan; Ma, Nan; Shan, Zuhua; Niu, Fenglei; Wang, Shengfei

2015-01-01

Highlights: • An improved AHP method is presented for screening key parameters used in passive system reliability analysis. • We take the special bottom parameters as criterion for calculation and the abrupt change of the results are verified. • Combination weights are also affected by uncertainty of input parameters. - Abstract: Passive safety system is widely used in the new generation nuclear power plant (NPP) designs such as AP1000 to improve the reactor safety benefitting from its simple construction and less request for human intervene. However, the functional failure induced by uncertainty in the system thermal–hydraulic (T–H) performance becomes one of the main contributors to system operational failure since the system operates based on natural circulation, which should be considered in the system reliability evaluation. In order to improve the calculation efficiency the key parameters which significantly affect the system T–H characteristics can be screened and then be analyzed in detail. The Analytical Hierarchy Process (AHP) is one of the efficient methods to analyze the influence of the parameters on a passive system based on the experts’ experience. The passive containment cooling system (PCCS) in AP1000 is one of the typical passive safety systems, nevertheless too many parameters need to be analyzed and the T–H model itself is more complicated, so the traditional AHP method should be mended to use for screening key parameters efficiently. In this paper, we adapt the improved method in hierarchy construction and experts’ opinions integration, some parameters at the bottom justly in the traditional hierarchy are studied as criterion layer in improved AHP, the rationality of the method and the effect of abrupt change with the data are verified. The passive containment cooling system (PCCS) in AP1000 is evaluated as an example, and four key parameters are selected from 49 inputs

Design and transient analyses of passive emergency feedwater system of CPR1000. Part 1. Air cooling condition

International Nuclear Information System (INIS)

Zhang Yapei; Qiu Suizheng; Su Guanghui; Tian Wenxi; Cao Jianhua; Lu Donghua; Fu Xiangang

2011-01-01

The steam generator secondary passive emergency feedwater system is a new design for traditional generation Ⅱ + reactor CPR1000. The passive emergency feedwater system is designed to supply water to the SG shell side and improve the safety and reliability of CPR1000 by completely or partially replacing traditional emergency water cooling system in the event of the feed line break (FLB) or loss of heat sink accident. The passive emergency feedwater system consists of steam generator (SG), heat exchanger (HX), air cooling tower, emergency makeup tank (EMT), and corresponding pipes and valves for air cooling condition. In order to improve the safety and reliability of CPR1000, the model of the primary loop system and the passive emergency feedwater system was developed to investigate residual heat removal capability of the passive emergency feedwater system and the transient characteristics of the primary loop system affected by the passive emergency feedwater system using RELAP5/MOD3.4. The transient characteristics of the primary loop system and the passive emergency feedwater system were calculated in the event of feed line break accident. Sensitivity studies of the passive emergency feedwater system were also conducted to investigate the response of the primary loop and the passive emergency feedwater system on the main parameters of the passive emergency feedwater system. The passive emergency feedwater system could supply water to the SG shell side from the EMT successfully. The calculation results showed that the passive emergency feedwater system could take away the decay heat from the primary loop effectively for air cooling condition, and that the single-phase and two-phase natural circulations were established in the primary loop and passive emergency feedwater system loop, respectively. (author)

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

International Nuclear Information System (INIS)

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

2003-01-01

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

AP1000 construction and operating costs

International Nuclear Information System (INIS)

Winters, J.W.; Corletti, M.M.; Thompson, M.

2001-01-01

Market analysis of the U.S. electricity generating market indicates that the generating cost of competitive new generating capacity must be less than $0.03/kw-hr. When such factors as an attractive return on investment and payback period are considered for a new nuclear electric generating facility, this results in the requirement to have an overnight capital cost of approximately $1000/kw. Industry executives indicate that any new nuclear plant must be able to compete in the de-regulated generation wholesale marketplace and provide a return to the shareholders. Against this standard, the costs of advanced nuclear power plants currently available are still too high. In the United States, the Utility Requirements Document for advanced light water reactor plants included a cost goal that was based on the cost of coal generated electricity at the time the document was written. Since that time, the cost of new generating capacity and the overall operating cost of generating electricity has gone down. This is a result of low natural gas prices, more efficient plants in general and the current record breaking reductions in outage times and operating costs for nuclear plants. The plant designs resulting from the United States advanced light water reactor plant programs received Design Certification from the United States Nuclear Regulatory Commission. Some are being deployed, with variations, in countries other than the United States. But they can not compete today with other sources of central station generation in the United States. (author)

AP1000 construction and operating costs

Energy Technology Data Exchange (ETDEWEB)

Winters, J.W.; Corletti, M.M. [Westinghouse Electric Corp., Pittsburgh, PA (United States); Thompson, M

2001-07-01

Market analysis of the U.S. electricity generating market indicates that the generating cost of competitive new generating capacity must be less than $0.03/kw-hr. When such factors as an attractive return on investment and payback period are considered for a new nuclear electric generating facility, this results in the requirement to have an overnight capital cost of approximately $1000/kw. Industry executives indicate that any new nuclear plant must be able to compete in the de-regulated generation wholesale marketplace and provide a return to the shareholders. Against this standard, the costs of advanced nuclear power plants currently available are still too high. In the United States, the Utility Requirements Document for advanced light water reactor plants included a cost goal that was based on the cost of coal generated electricity at the time the document was written. Since that time, the cost of new generating capacity and the overall operating cost of generating electricity has gone down. This is a result of low natural gas prices, more efficient plants in general and the current record breaking reductions in outage times and operating costs for nuclear plants. The plant designs resulting from the United States advanced light water reactor plant programs received Design Certification from the United States Nuclear Regulatory Commission. Some are being deployed, with variations, in countries other than the United States. But they can not compete today with other sources of central station generation in the United States. (author)

The Westinghouse AP600 an advanced nuclear option for small or medium electricity grids

International Nuclear Information System (INIS)

Bruschi, H. J.; Novak, V.

1996-01-01

During the early days of commercial nuclear power, many countries looking to add nuclear power to their energy mix required large plants to meet the energy needs of rapidly growing populations and large industrial complexes. The majority of plants worldwide are in the range of 100 megawatts and beyond. During the 1970s, it became apparent that a smaller nuclear plants would appeal to utilities looking to add additional power capacity to existing grids, or to utilities in smaller countries which were seeking efficient, new nuclear generation capacity for the first time. For instance, the Westinghouse-designed 600 megawatt Krsko plant in Slovenia began operation in 1980, providing electricity to inhabitants of relatively small, yet industrial populations of Slovenia and Croatia. This plant design incorporated the best, proven technology available at that time, based on 20 years of Westinghouse PWR pioneering experience. Beginning in the early 1980s, Westinghouse began to build further upon that experience - in part through the advanced light water reactor programs established by the Electric Power Research institute (EPRI) and the U.S. Department of Energy (DOE) - to design a simplified, advanced nuclear reactor in the 600 megawatt range. Originally, Westinghouse's development of its AP600 (advanced, passive 600-megawatt) plants was geared towards the needs of U.S. utilities which specified smaller, simplified nuclear options for the decades ahead. It soon became evident that the small and medium sized electricity grids of international markets could benefit from this new reactor. From the earliest days of Westinghouse's AP600 development, the corporation invited members of the international nuclear community to take part in the design, development and testing of the AP600 - with the goal of designing a reactor that would meet the diverse needs of an international industry composed of countries with similar, yet different, concerns. (author)

SWR 1000: An Advanced, Medium-Sized Boiling Water Reactor, Ready for Deployment

International Nuclear Information System (INIS)

Brettschuh, Werner

2006-01-01

The latest developments in nuclear power generation technology mainly concern large-capacity plants in the 1550 -1600 MW range, or very small plants (100 - 350 MW). The SWR 1000 boiling water reactor (BWR), by contrast, offers all of the advantages of an advanced plant design, with excellent safety performance and competitive power generation costs, in the medium-capacity range (1000 - 1250 MW). The SWR 1000 is particularly suitable for countries whose power systems are not designed for large-capacity generating facilities. The economic efficiency of this medium-sized plant in comparison with large-capacity designs is achieved by deploying very simple passive safety equipment, simplified systems for plant operation, and a very simple plant configuration in which systems engineering is optimized and dependence on electrical and instrumentation and control (I and C) systems is reduced. In addition, systems and components that require protection against natural and external man-made hazards are accommodated in such a way that as few buildings as possible have to be designed to withstand the loads from such events. The fuel assemblies to be deployed in the SWR 1000 core, meanwhile, have been enlarged from a 10 x 10 rod array to a 12 x 12 array. This reduces the total number of fuel assemblies in the core and thus also the number of control rods and control rod drives, as well as in-core neutron flux monitors. The design owes its competitiveness to the fact that investment costs, maintenance costs and fuel cycle costs are all lower. In addition, refueling outages are shorter, thanks to the reduced scope of outage activities. The larger fuel assemblies have been extensively and successfully tested, as have all of the other new components and systems incorporated into the plant design. As in existing plants, the forced coolant circulation method is deployed, ensuring problem-free startup, and enabling plant operators to adjust power rapidly in the high power range (70

Study and practice on NI design management of the first AP1000 plant

International Nuclear Information System (INIS)

Jiang Feizhou

2014-01-01

For difficulties on NI design management during the construction of the world's first AP1000 unit, Sanmen Nuclear Power Corporation has established the effective method to improve the management of the NI design and the project construction based on the practice and improvement. This paper introduces difficulties on NI design management, and analyzes the causes of problems, and put forward recommended measures, to provide help and reference for the construction of similar nuclear power plants. (authors)

The preliminary analysis of establishing the cost control system of AP1000 for the Haiyang nuclear power project

International Nuclear Information System (INIS)

Li Jing; Li Xiaobing

2012-01-01

The AP1000 technology has been first applied to Nuclear Power Plant construction in China. Haiyang Project is the second plant which applies the new technology, and it is the key to the success of the project, that how to control the cost. The cost control of AP1000 is to manage and monitor all the cost of the project, including the cost of project management, design, procurement, construction, and startup/commissioning. For the Haiyang Project, the cost control system should be established to ensure that the evaluation of the procurement order should be covered in the original budget, and all potential commitments are evaluated and approved within the confinement of cost control, and reduce the risk of the first reactor and get the most profit. (authors)

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.)

A neutronic feasibility study of the AP1000 design loaded with fully ceramic micro-encapsulated fuel

International Nuclear Information System (INIS)

Liang, C.; Ji, W.

2013-01-01

A neutronic feasibility study is performed to evaluate the utilization of fully ceramic microencapsulated (FCM) fuel in the AP1000 reactor design. The widely used Monte Carlo code MCNP is employed to perform the full core analysis at the beginning of cycle (BOC). Both the original AP1000 design and the modified design with the replacement of uranium dioxide fuel pellets with FCM fuel compacts are modeled and simulated for comparison. To retain the original excess reactivity, ranges of fuel particle packing fraction and fuel enrichment in the FCM fuel design are first determined. Within the determined ranges, the reactor control mechanism employed by the original design is directly used in the modified design and the utilization feasibility is evaluated. The worth of control of each type of fuel burnable absorber (discrete/integral fuel burnable absorbers and soluble boron in primary coolant) is calculated for each design and significant differences between the two designs are observed. Those differences are interpreted by the fundamental difference of the fuel form used in each design. Due to the usage of silicon carbide as the matrix material and the fuel particles fuel form in FCM fuel design, neutron slowing down capability is increased in the new design, leading to a much higher thermal spectrum than the original design. This results in different reactivity and fission power density distributions in each design. We conclude that a direct replacement of fuel pellets by the FCM fuel in the AP1000 cannot retain the original optimum reactor core performance. Necessary modifications of the core design should be done and the original control mechanism needs to be re-designed. (authors)

Status of the Advanced Photon Source (APS) linear accelerator

International Nuclear Information System (INIS)

White, M.; Berg, W.; Fuja, R.; Grelick, A.; Mavrogenes, G.; Nassiri, A.; Russell, T.; Wesolowski, W.

1993-01-01

A 2856-MHz S-band, 450-MeV electron/positron linear accelerator is the first part of the injector for the Advanced Photon Source (APS) 7-GeV storage ring. Construction of the APS linac is currently nearing completion, and commissioning will begin in July 1993. The linac and its current status are discussed in this paper

Westinghouse-GOTHIC comparisons to AP600 passive containment cooling tests

International Nuclear Information System (INIS)

Kennedy, M.D.; Woodcock, J.; Gresham, J.A.

1994-01-01

Westinghouse-GOTHIC is a thermal-hydraulics code well suited to analyzing passively cooled containments which depend on heat removal primarily through the containment shell. The code includes boundary layer heat and mass transfer correlations. A liquid film convective energy transport model has been added to the Westinghouse-GOTHIC code to account for the sensible heat change of the applied exterior water. The objective of this paper is to compare the code's predictions of the AP600 large scale test facility with and without the liquid film convective energy transport model. The predicted vessel pressure and integrated heat rate with and without the film convective energy transport model will be compared to the measured data. (author)

Knowledge-based software design for Defense-in-Depth risk monitor system and application for AP1000

International Nuclear Information System (INIS)

Ma Zhanguo; Yoshikawa, Hidekazu; Yang Ming; Nakagawa, Takashi

2017-01-01

As part of the new risk monitor system, the software for the plant Defense-in-Depth (DiD) risk monitor system was designed based on the state-transition and finite-state machine, and then the knowledge-based software was developed by object-oriented method utilizing the Unified Modeling Language (UML). Currently, there are mainly two functions in the developed plant DiD risk monitor software that are knowledge-base editor which is used to model the system in a hierarchical manner and the interaction simulator that simulates the interactions between the different actors in the model. In this paper, a model for playing its behavior is called an Actor which is modeled at the top level. The passive safety AP1000 power plant was studied and the small-break loss-of-coolant accident (SBLOCA) design basis accident transient is modeled using the plant DiD risk monitor software. Furthermore, the simulation result is shown for the interactions between the actors which are defined in the plant DiD risk monitor system as PLANT actor, OPERATOR actor, and SUPERVISOR actor. This paper shows that it is feasible to model the nuclear power plant knowledge base using the software modeling technique. The software can make the large knowledge base for the nuclear power plant with small effort. (author)

Neutronic and thermal-hydraulic calculations for the AP-1000 NPP with the MCNP6 and SERPENT codes

International Nuclear Information System (INIS)

Stefani, Giovanni Laranjo; Maiorino, Jose R.; Santos, Thiago A.

2015-01-01

The AP-1000 is an evolutionary PWR reactor designed as an evolution of the AP-600 project. The reactor is already pre-licensed by NRC, and is considered to have achieved high standards of safety, possible short construction time and good economic competitiveness. The core is a 17x17 typical assembly using Zirlo as cladding, 3 different enrichment regions, and is controlled by boron, control banks, and burnable poison. The expected fuel final burnup is 62 MWD/ton U and a cycle of 18 months. In this paper we present results for neutronic and thermal-hydraulic calculations for the AP-1000. We use the MCNP6 and SERPENT codes to calculate the first cycle of operation. The calculated parameters are K eff at BOL and EOL and its variation with burnup and neutron flux, and reactivity coefficients. The production of transuranic elements such as Pu-239 and Pu-241, and burning fuel are calculated over time. In the work a complete reactor was burned for 450 days with no control elements, boron or burnable poison were considered, these results were compared with data provided by the Westinghouse. The results are compared with those reported in the literature. A simple thermal hydraulic analysis allows verification of thermal limits such as fuel and cladding temperatures, and MDNB. (author)

Neutronic and thermal-hydraulic calculations for the AP-1000 NPP with the MCNP6 and SERPENT codes

Energy Technology Data Exchange (ETDEWEB)

Stefani, Giovanni Laranjo; Maiorino, Jose R.; Santos, Thiago A., E-mail: giovanni.laranjo@ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br, E-mail: thiago.santos@ufabc.edu.br [Universidade Federal do ABC (CECS/UFABC), Santo Andre, SP (Brazil). Centro de Engenharia, Modelagem e Ciencias Sociais; Rossi, Pedro R., E-mail: pedro.russorossi@gmail.com [FERMIUM - Tecnologia Nuclear, Sao Paulo, SP (Brazil)

2015-07-01

The AP-1000 is an evolutionary PWR reactor designed as an evolution of the AP-600 project. The reactor is already pre-licensed by NRC, and is considered to have achieved high standards of safety, possible short construction time and good economic competitiveness. The core is a 17x17 typical assembly using Zirlo as cladding, 3 different enrichment regions, and is controlled by boron, control banks, and burnable poison. The expected fuel final burnup is 62 MWD/ton U and a cycle of 18 months. In this paper we present results for neutronic and thermal-hydraulic calculations for the AP-1000. We use the MCNP6 and SERPENT codes to calculate the first cycle of operation. The calculated parameters are K{sub eff} at BOL and EOL and its variation with burnup and neutron flux, and reactivity coefficients. The production of transuranic elements such as Pu-239 and Pu-241, and burning fuel are calculated over time. In the work a complete reactor was burned for 450 days with no control elements, boron or burnable poison were considered, these results were compared with data provided by the Westinghouse. The results are compared with those reported in the literature. A simple thermal hydraulic analysis allows verification of thermal limits such as fuel and cladding temperatures, and MDNB. (author)

76 FR 44377 - Advisory Committee on Reactor Safeguards (ACRS), Meeting of the ACRS Subcommittee on AP1000...

Science.gov (United States)

2011-07-25

..., 2011-1 p.m. until 5 p.m. The Subcommittee will review technical updates in Revision 19 to the AP1000 Design Control Document (DCD). The Subcommittee will hear presentations by and hold discussions with the... statements can be obtained from the website cited above or by contacting the identified DFO. Moreover, in...

Application of fault tree methodology to modeling of the AP1000 plant digital reactor protection system

International Nuclear Information System (INIS)

Teolis, D.S.; Zarewczynski, S.A.; Detar, H.L.

2012-01-01

The reactor trip system (RTS) and engineered safety features actuation system (ESFAS) in nuclear power plants utilizes instrumentation and control (IC) to provide automatic protection against unsafe and improper reactor operation during steady-state and transient power operations. During normal operating conditions, various plant parameters are continuously monitored to assure that the plant is operating in a safe state. In response to deviations of these parameters from pre-determined set points, the protection system will initiate actions required to maintain the reactor in a safe state. These actions may include shutting down the reactor by opening the reactor trip breakers and actuation of safety equipment based on the situation. The RTS and ESFAS are represented in probabilistic risk assessments (PRAs) to reflect the impact of their contribution to core damage frequency (CDF). The reactor protection systems (RPS) in existing nuclear power plants are generally analog based and there is general consensus within the PRA community on fault tree modeling of these systems. In new plants, such as AP1000 plant, the RPS is based on digital technology. Digital systems are more complex combinations of hardware components and software. This combination of complex hardware and software can result in the presence of faults and failure modes unique to a digital RPS. The United States Nuclear Regulatory Commission (NRC) is currently performing research on the development of probabilistic models for digital systems for inclusion in PRAs; however, no consensus methodology exists at this time. Westinghouse is currently updating the AP1000 plant PRA to support initial operation of plants currently under construction in the United States. The digital RPS is modeled using fault tree methodology similar to that used for analog based systems. This paper presents high level descriptions of a typical analog based RPS and of the AP1000 plant digital RPS. Application of current fault

SWR 1000 related containment cooling system tests in PANDA

International Nuclear Information System (INIS)

Dreier, J.; Aubert, C.; Huggenberger, M.; Strassberger, H.J.; Yadigaroglu, G.

2000-01-01

Since 1991 the Paul Scherrer Institute has participated in the investigations of several of the new passive Advanced Light Water Reactor designs proposed world-wide. The current phase of the project, ALPHA-II, is focused on both the boiling water and the pressurized water reactor passive designs and consists of three projects under the sponsorship of the European Commission. The paper describes the performed PANDA transient system tests related to one of these projects, called 'BWR R and D Cluster for Innovative Passive Safety Systems (IPSS)', and details the PSI contribution to the experimental investigation of passive containment cooling by a Building Condenser system which is part of the advanced Boiling Water Reactor SWR 1000 designed by Siemens. First, a short description of the relevant systems of the SWR 1000 design and its simulation in the PANDA facility are presented. After the description of the experimental programme for the large-scale integral system test investigations in the PANDA facility, the main results of the performed tests are also given. Finally, the main conclusions, based on the to date available experimental results and their analysis, are summarised. (author)

Shadow management applied in the first AP1000 project under the islands contract condition

International Nuclear Information System (INIS)

Liu Xiao

2010-01-01

As the global first AP1000 nuclear project, Sanmen phase I nuclear project itself has many challenges from design, procurement to construction managements for non practical nuclear project and experience can be referenced. Islands contract pattern was adopted by this project and this contract pattern has its own strength and weakness. Considering the negative influence result from the first unit, this project has the great postpone risk. Shadow management here tries to reduce these risks and enhance the project surveillance and control by the owner to promote the final goal of this project. (authors)

Evaluation of the PRHRS Performance Degradation due to Non-Condensable Gas for the Small and Medium Reactor using MARS-KS code

International Nuclear Information System (INIS)

Kim, Sook Kwan; Sim, Suk Ku; Park, Ju Yeop; Seol, Kwang Won; Ryu, Yong Ho

2011-01-01

The effect of non-condensable gas on the performance of PRHRS (Passive Residual Heat Removal System) of the Small and Medium Reactor(SMR) was evaluated during a loss of flow event. Since the TMI accident in 1979, the passive systems have been considered in the advanced reactors as a feature of design improvement because the passive system simplifies the system and thus increases the reliability of the system. The Westinghouse received the design certification from the USNRC for the AP600 and AP1000 passive type pressurized water reactors. The APR+ under development by KEPCO considers the use of PAFS (Passive Auxiliary Feedwater System). And the PRHRS is adopted as a passive secondary heat removal system for the SMART (System-integrated Modular Advanced ReacTor)

Definition of parameters for a test section for the analysis of natural convection and coolant loss in the AP1000 nuclear reactor by similarity laws and fractional scaling analysis

Energy Technology Data Exchange (ETDEWEB)

Cadiz, Luis Felipe S.; Bezerra, Mario Augusto [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Departamento de Energia Nuclear; Lima, Fernando Roberto A., E-mail: falima@cnen.gov.br [Centro Regional de Ciências Nucleares do Nordeste (CRCN-NE/CNEN-PB), Recife, PB (Brazil)

2017-07-01

The present work develops and analyzes the main parameters of a test section for natural convection in case of a failure of the pumping system as much as the loss of coolant in refrigeration accidents. For this realization, a combination of laws of basic similarity and an innovative scale methodology, known as Fractional Scaling Analysis (FSA), was developed. The depressurizing is analyzed when a rupture occurs in one of the primary system piping of the AP1000 nuclear reactor. This reactor is developed by Westinghouse Electric Co., which is a PWR (Pressurized Water Reactor) with an electric power equal to 1000MW. Such a reactor is provided with a passive safety system that promotes considerable improvements in the safety, reliability, protection and reduction of costs of a nuclear power plant. The FSA is based on two concepts: fractional scale and hierarchy. It is used to provide experimental data that generate quantitative evaluation criteria as well as operational parameters in thermal and hydraulic processes of nuclear power plants. The results were analyzed with the use of computational codes. (author)

APS [Advanced Photon Source] interests in PEP

International Nuclear Information System (INIS)

Moncton, D.E.; Shenoy, G.K.; Mills, D.M.

1987-11-01

As one of the very few high-energy electron storage rings in the world, potentially available for synchrotron radiation studies, PEP represents an opportunity to accomplish certain preconstruction R and D tasks relevant to the successful construction and operation of dedicated user facilities such as the Advanced Photon Source (APS) at Argonne. Three topical areas are discussed: Accelerator R and D, Insertion Devices (ID) R and D, and Beam Line Instrumentation R and D

Issues affecting advanced passive light-water reactor safety analysis

International Nuclear Information System (INIS)

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

1992-01-01

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

Preliminary investigation on reliability assessment of passive safety system

International Nuclear Information System (INIS)

Huang Changfan; Kuang Bo

2012-01-01

The reliability evaluation of passive safety system plays an important part in probabilistic safety assessment (PSA) of nuclear power plant applying passive safety design, which depends quantitatively on reliabilities of passive safety system. According to the object of reliability assessment of passive safety system, relevant parameters are identified. Then passive system behavior during accident scenarios are studied. A practical example of this method is given for the case of reliability assessment of AP1000 passive heat removal system in loss of normal feedwater accident. Key and design parameters of PRHRS are identified and functional failure criteria are established. Parameter combinations acquired by Latin hyper~ cube sampling (LHS) in possible parametric ranges are input and calculations of uncertainty propagation through RELAP5/MOD3 code are carried out. Based on the calculations, sensitivity assessment on PRHRS functional criteria and reliability evaluation of the system are presented, which might provide further PSA with PRHR system reliability. (authors)

Passive cooling systems in power reactors

International Nuclear Information System (INIS)

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

1996-01-01

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

Application of status uncertainty analysis methods for AP1000 LBLOCA calculation

International Nuclear Information System (INIS)

Zhang Shunxiang; Liang Guoxing

2012-01-01

Parameter uncertainty analysis is developed by using the reasonable method to establish the response relations between input parameter uncertainties and output uncertainties. The application of the parameter uncertainty analysis makes the simulation of plant state more accuracy and improves the plant economy with reasonable security assurance. The AP1000 LBLOCA was analyzed in this paper and the results indicate that the random sampling statistical analysis method, sensitivity analysis numerical method and traditional error propagation analysis method can provide quite large peak cladding temperature (PCT) safety margin, which is much helpful for choosing suitable uncertainty analysis method to improve the plant economy. Additionally, the random sampling statistical analysis method applying mathematical statistics theory makes the largest safety margin due to the reducing of the conservation. Comparing with the traditional conservative bounding parameter analysis method, the random sampling method can provide the PCT margin of 100 K, while the other two methods can only provide 50-60 K. (authors)

Passive Nuclear Plants Program (UPDATE)

International Nuclear Information System (INIS)

Chimeno, M. A.

1998-01-01

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

Design of a decoupled AP1000 reactor core control system using digital proportional–integral–derivative (PID) control based on a quasi-diagonal recurrent neural network (QDRNN)

Energy Technology Data Exchange (ETDEWEB)

Wei, Xinyu, E-mail: xyuwei@mail.xjtu.edu.cn; Wang, Pengfei, E-mail: pengfeixiaoli@yahoo.cn; Zhao, Fuyu, E-mail: fuyuzhao_xj@163.com

2016-08-01

Highlights: • We establish a disperse dynamic model for AP1000 reactor core. • A digital PID control based on QDRNN is used to design a decoupling control system. • The decoupling performance is verified and discussed. • The decoupling control system is simulated under the load following operation. - Abstract: The control system of the AP1000 reactor core uses the mechanical shim (MSHIM) strategy, which includes a power control subsystem and an axial power distribution control subsystem. To address the strong coupling between the two subsystems, an interlock between the two subsystems is used, which can only alleviate but not eliminate the coupling. Therefore, sometimes the axial offset (AO) cannot be controlled tightly, and the flexibility of load-following operation is limited. Thus, the decoupling of the original AP1000 reactor core control system is the focus of this paper. First, a two-node disperse dynamic model is established for the AP1000 reactor core to use PID control. Then, a digital PID control system based on a quasi-diagonal recurrent neural network (QDRNN) is designed to decouple the original system. Finally, the decoupling of the control system is verified by the step signal and load-following condition. The results show that the designed control system can decouple the original system as expected and the AO can be controlled much more tightly. Moreover, the flexibility of the load following is increased.

Design of a decoupled AP1000 reactor core control system using digital proportional–integral–derivative (PID) control based on a quasi-diagonal recurrent neural network (QDRNN)

International Nuclear Information System (INIS)

Wei, Xinyu; Wang, Pengfei; Zhao, Fuyu

2016-01-01

Highlights: • We establish a disperse dynamic model for AP1000 reactor core. • A digital PID control based on QDRNN is used to design a decoupling control system. • The decoupling performance is verified and discussed. • The decoupling control system is simulated under the load following operation. - Abstract: The control system of the AP1000 reactor core uses the mechanical shim (MSHIM) strategy, which includes a power control subsystem and an axial power distribution control subsystem. To address the strong coupling between the two subsystems, an interlock between the two subsystems is used, which can only alleviate but not eliminate the coupling. Therefore, sometimes the axial offset (AO) cannot be controlled tightly, and the flexibility of load-following operation is limited. Thus, the decoupling of the original AP1000 reactor core control system is the focus of this paper. First, a two-node disperse dynamic model is established for the AP1000 reactor core to use PID control. Then, a digital PID control system based on a quasi-diagonal recurrent neural network (QDRNN) is designed to decouple the original system. Finally, the decoupling of the control system is verified by the step signal and load-following condition. The results show that the designed control system can decouple the original system as expected and the AO can be controlled much more tightly. Moreover, the flexibility of the load following is increased.

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

International Nuclear Information System (INIS)

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

2002-01-01

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

Refer to AP1000 for discussing the betterment of seismic design of internal nuclear power plant

International Nuclear Information System (INIS)

Gong Zhenbang; Zhang Renyan

2014-01-01

As a reference technique of AP1000, This paper discussed the betterment of seismic design of nuclear power plant in three ways. (1) Establish design criteria and guidelines for protection from seismic interaction; (2) Nuclear power plant seismic design of eliminating or weaken operation-basis earthquake; (3) Develop the seismic margin analysis (SMA) of the nuclear power plant. These three aspect are frontier technology in internal seismic design of internal nuclear power plant, and also these three technology are related intimately. (authors)

Passive safety and the advanced liquid metal reactors

International Nuclear Information System (INIS)

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

1988-01-01

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

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

International Nuclear Information System (INIS)

2014-09-01

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

AP600 passive containment cooling system phenomena identification and ranking table

International Nuclear Information System (INIS)

Spencer, D.R.; Woodcock, Joel

1999-01-01

This paper presents the Phenomena Identification and Ranking Table (PIRT) used in the containment Design Basis Analysis (DBA) for the AP600 nuclear power plant. The PIRT is a tool generally applied to best estimate thermal hydraulic analyses. In the conservative analytical modeling approach used for the AP600 DBA containment pressure response, the PIRT was a tool used to show completeness and relevance of the test database in accordance with the Code of Federal Regulations for advanced plant design. Additionally, the ranking of phenomena by relative importance in a PIRT allows appropriate focusing of resources during model development and licensing review. The focus of the paper is on the organization and structure of the PIRT to show level of detail and format accepted for the AP600, for potential application to other containment designs or accident scenarios. Conclusions of general interest are discussed regarding table organization and structure, the process for developing relative ranking and incorporating expert opinion, and the definition and usage of the relative ranking in support of the conservative evaluation model. The AP600 containment evaluation model approach, as influenced by the relative rankings, is briefly described to put into context this unique application of the PIRT to a conservative methodology. The bases for relative ranking of each phenomenon, which included expert opinion, and quantitative results of scaling and testing, was submitted to the NRC as part of AP600-specific evaluations. Since a PIRT supports the sufficiency of both a testing program and analytical modeling, the process followed to generate and confirm the PIRT, an important part of the licensing acceptance, was a focus of extensive NRC review. General descriptions of key phenomena are provided to aid in understanding the containment PIRT for more general applications for containment evaluations of other PWR designs or for other scenarios. (author)

A Metal Fuel Core Concept for 1000 MWt Advanced Burner Reactor

International Nuclear Information System (INIS)

Yang, W.S.; Kim, T.K.; Grandy, C.

2007-01-01

This paper describes the core design and performance characteristics of a metal fuel core concept for a 1000 MWt Advanced Burner Reactor. A ternary metal fuel form of U-TRU-Zr was assumed with weapons grade plutonium feed for the startup core and TRU recovered from LWR spent fuel for the recycled equilibrium core. A compact burner core was developed by trade-off between the burnup reactivity loss and TRU conversion ratio, with a fixed cycle length of one-year. In the startup core, the average TRU enrichment is 15.5%, the TRU conversion ratio is 0.81, and the burnup reactivity loss over a cycle is 3.6% Δk. The heavy metal and TRU inventories are 13.1 and 2.0 metric tons, respectively. The average discharge burnup is 93 MWd/kg, and the TRU consumption rate is 55.5 kg/year. For the recycled equilibrium core, the average TRU enrichment is 22.1 %, the TRU conversion ratio is 0.73, and the burnup reactivity loss is 2.2% Δk. The TRU inventory and consumption rate are 2.9 metric tons and 81.6 kg/year, respectively. The evaluated reactivity coefficients provide sufficient negative feedbacks. The control systems provide shutdown margins that are more than adequate. The integral reactivity parameters for quasi-static reactivity balance analysis indicate favorable passive safety features, although detailed safety analyses are required to verify passive safety behavior. (authors)

Overview of the advanced photon source (APS)

International Nuclear Information System (INIS)

White, M.M.

1994-01-01

The Advanced Photon Source (APS) is a state-of-the-art synchrotron light source facility dedicated to the production of extremely brilliant x-ray beams for research. Its super-intense x-ray beams will be used in many areas of research including industrial research, biological and medical research, defense-related research, and basic research. The APS x-ray beams will allow scientists to study smaller samples, more complex systems, faster reactions and processes, and gather data at a greater level of detail than has been possible to date. Creation of these beams begins with electron production by an electron gun with a thermionic cathode. The electrons are accelerated to 200 MeV by a linear accelerator (linac) and then impinge on a tungsten target, resulting in electron-positron pair production. The positrons are accelerated to 450 MeV in the remainder of the linac, then accumulated, damped, and transferred to a synchrotron that increases their energy to 7 GeV. The 7-GeV positrons are injected into a storage ring, where they pass through special magnets that cause them to emit x-rays of the desired quality. Construction at ANL is nearly complete at this time, and the APS will begin operating for users in 1996. The accelerator and experimental facilities are described in this paper, and a brief overview of some of the experimental programs is given

WASA-BOSS. Development and application of Severe Accident Codes. Evaluation and optimization of accident management measures. Subproject D. Study on water film cooling for PWR's passive containment cooling system. Final report

Energy Technology Data Exchange (ETDEWEB)

Huang, Xi

2016-07-15

In the present study, a new phenomenological model was developed, to describe the water film flow under conditions of a passive containment cooling system (PCCS). The new model takes two different flow regimes into consideration, i.e. continuous water film and rivulets. For water film flow, the traditional Nusselt's was modified, to consider orientation angle and surface sheer stress. The transition from water film to rivulet as well as the structure of the stable rivulet at its onset point was modeled by using the minimum energy principle (MEP) combined with conservation equations. In addition, two different contact angles, i.e. advancing angle and retreating angle, were applied to take the hysteresis effect into consideration. The models of individual processes were validated as far as possible based on experimental data selected from open literature and from collaboration partner as well. With the models a new program module was developed and implemented into the COCOSYS program. The extended COCOSYS program was applied to analyze the containment behavior of the European generic containment and the performance of the passive containment cooling system ofthe AP1000. The results indicate clearly the importance of the new model and provide information for the optimization of the PCCS of AP1000.

WASA-BOSS. Development and application of Severe Accident Codes. Evaluation and optimization of accident management measures. Subproject D. Study on water film cooling for PWR's passive containment cooling system. Final report

International Nuclear Information System (INIS)

Huang, Xi

2016-07-01

In the present study, a new phenomenological model was developed, to describe the water film flow under conditions of a passive containment cooling system (PCCS). The new model takes two different flow regimes into consideration, i.e. continuous water film and rivulets. For water film flow, the traditional Nusselt's was modified, to consider orientation angle and surface sheer stress. The transition from water film to rivulet as well as the structure of the stable rivulet at its onset point was modeled by using the minimum energy principle (MEP) combined with conservation equations. In addition, two different contact angles, i.e. advancing angle and retreating angle, were applied to take the hysteresis effect into consideration. The models of individual processes were validated as far as possible based on experimental data selected from open literature and from collaboration partner as well. With the models a new program module was developed and implemented into the COCOSYS program. The extended COCOSYS program was applied to analyze the containment behavior of the European generic containment and the performance of the passive containment cooling system ofthe AP1000. The results indicate clearly the importance of the new model and provide information for the optimization of the PCCS of AP1000.

CLEARING MAGNET DESIGN FOR APS-U

Energy Technology Data Exchange (ETDEWEB)

Abliz, M.; Grimmer, J.; Jaski, Y.; Westferro, F.; Ramanathan, M.

2017-06-25

The Advanced Photon Source is in the process of developing an upgrade (APS-U) of the storage ring. The upgrade will be converting the current double bend achromat (DBA) lattice to a multi-bend achromat (MBA) lattice. In addition, the storage ring will be operated at 6 GeV and 200 mA with regular swap-out injection to keep the stored beam current constant [1]. The swap-out injection will take place with beamline shutters open. For radiation safety to ensure that no electrons can exit the storage ring, a passive method of protecting the beamline and containing the electrons inside the storage ring is proposed. A clearing magnet will be located in all beamline front ends inside the storage ring tunnel. This article will discuss the features and design of the clearing magnet scheme for APS-U.

Experimental research progress on passive safety systems of Chinese advanced PWR

International Nuclear Information System (INIS)

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

2003-01-01

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

CMT scaling analysis and distortion evaluation in passive integral test facility

International Nuclear Information System (INIS)

Deng Chengcheng; Qin Benke; Wang Han; Chang Huajian

2013-01-01

Core makeup tank (CMT) is the crucial device of AP1000 passive core cooling system, and reasonable scaling analysis of CMT plays a key role in the design of passive integral test facilities. H2TS method was used to perform scaling analysis for both circulating mode and draining mode of CMT. And then, the similarity criteria for CMT important processes were applied in the CMT scaling design of the ACME (advanced core-cooling mechanism experiment) facility now being built in China. Furthermore, the scaling distortion results of CMT characteristic Ⅱ groups of ACME were calculated. At last, the reason of scaling distortion was analyzed and the distortion evaluation was conducted for ACME facility. The dominant processes of CMT circulating mode can be adequately simulated in the ACME facility, but the steam condensation process during CMT draining is not well preserved because the excessive CMT mass leads to more energy to be absorbed by cold metal. However, comprehensive analysis indicates that the ACME facility with high-pressure simulation scheme is able to properly represent CMT's important phenomena and processes of prototype nuclear plant. (authors)

SWR 1000 approaching bidding phase status

International Nuclear Information System (INIS)

Brettschuh, W.; Meseth, J.

2000-01-01

The SWR 1000 is an advanced version of the previous 69 and 72 lines by Siemens/KWU. Like its predecesor, the SWR 600, this development project is handled by Siemens on behalf of German utilities, and supported by various European partners contributing engineering and experimental work. With its electric power of approx. 1000 MW, the SWR 1000 differs from earlier KWU boiling water reactor designs especially in having several passive safety systems in addition to some of the active systems customarily used in earlier plants. In a postulated failure of the active safety systems, the passive safety systems can completely take over their function. The passive safety systems are characterized by not needing any external power supply after having been activated, and by not requiring initiation by the I and C system to be activated. In addition, plant operating systems were streamlined, always subject to the guiding principle to employ, as far as possible, technical components and systems with proven operating records. In late 1999, the basic design phase in the development of the SWR 1000 was completed in which all newly designed components and systems as well as those modified greatly over existing BWR plants, including electrical and I and C systems, were designed and the main buildings, such as the reactor building, turbine hall, auxiliary and utilities building, were planned. All newly designed components, systems, and processes were successfully tested for their functioning capability and capacity in large scale tests. The results of the basic design phase were compiled in a site independent safety report based on the rules and regulations applicable in Germany. (orig.) [de

FSI effects and seismic performance evaluation of water storage tank of AP1000 subjected to earthquake loading

Energy Technology Data Exchange (ETDEWEB)

Zhao, Chunfeng, E-mail: zhaowindy@126.com [Institute of Earthquake Engineering, Dalian University of Technology, Dalian 116024 (China); School of Civil Engineering, Hefei University of Technology, Anhui Province 230009 (China); Chen, Jianyun; Xu, Qiang [Institute of Earthquake Engineering, Dalian University of Technology, Dalian 116024 (China)

2014-12-15

Graphical abstract: - Highlights: • Water sloshing and oscillation of water tank under earthquake are simulated by FEM. • The influences of various water levels on seismic response are investigated. • ALE algorithm is applied to study the fluid–structure interaction effects. • The effects of different water levels in reducing seismic response are compared. • The optimal water level of water tank under seismic loading is obtained. - Abstract: The gravity water storage tank of AP1000 is designed to cool down the temperature of containment vessel by spray water when accident releases mass energy. However, the influence of fluid–structure interaction between water and water tank of AP1000 on dynamic behavior of shield building is still a hot research question. The main objective of the current study is to investigate how the fluid–structure interaction affects the dynamic behavior of water tank and whether the water sloshing and oscillation can reduce the seismic response of the shield building subjected to earthquake. For this purpose, a fluid–structure interaction algorithm of finite element technique is employed for the seismic analysis of water storage tank of AP1000. In the finite element model, 8 cases height of water, such as 10.8, 9.8, 8.8, 7.8, 6.8, 5.8, 4.8, and 3.8 m, are established and compared with the empty water tank in order to demonstrate the positive effect in mitigating the seismic response. An Arbitrary Lagrangian Eulerian (ALE) algorithm is used to simulate the fluid–structure interaction, fluid sloshing and oscillation of water tank under the El-Centro earthquake. The correlation between seismic response and parameters of water tank in terms of height of air (h{sub 1}), height of water (h{sub 2}), height ratio of water to tank (h{sub 2}/H{sub w}) and mass ratio of water to total structure (m{sub w}/m{sub t}) is also analyzed. The numerical results clearly show that the optimal h{sub 2}, h{sub 2}/H{sub w} and m{sub w}/m{sub t

An approach for assessing ALWR passive safety system reliability

International Nuclear Information System (INIS)

Hake, T.M.

1991-01-01

Many advanced light water reactor designs incorporate passive rather than active safety features for front-line accident response. A method for evaluating the reliability of these passive systems in the context of probabilistic risk assessment has been developed at Sandia National Laboratories. This method addresses both the component (e.g. valve) failure aspect of passive system failure, and uncertainties in system success criteria arising from uncertainties in the system's underlying physical processes. These processes provide the system's driving force; examples are natural circulation and gravity-induced injection. This paper describes the method, and provides some preliminary results of application of the approach to the Westinghouse AP600 design

Large LOCA accident analysis for AP1000 under earthquake

International Nuclear Information System (INIS)

Yu, Yu; Lv, Xuefeng; Niu, Fenglei

2015-01-01

Highlights: • Seismic failure event probability is induced by uncertainties in PGA and in Am. • Uncertainty in PGA is shared by all the components at the same place. • Relativity induced by sharing PGA value can be analyzed explicitly by MC method. • Multi components failures and accident sequences will occur under high PGA value. - Abstract: Seismic probabilistic safety assessment (PSA) is developed to give the insight of nuclear power plant risk under earthquake and the main contributors to the risk. However, component failure probability including the initial event frequency is the function of peak ground acceleration (PGA), and all the components especially the different kinds of components at same place will share the common ground shaking, which is one of the important factors to influence the result. In this paper, we propose an analysis method based on Monte Carlo (MC) simulation in which the effect of all components sharing the same PGA level can be expressed by explicit pattern. The Large LOCA accident in AP1000 is analyzed as an example, based on the seismic hazard curve used in this paper, the core damage frequency is almost equal to the initial event frequency, moreover the frequency of each accident sequence is close to and even equal to the initial event frequency, while the main contributors are seismic events since multi components and systems failures will happen simultaneously when a high value of PGA is sampled. The component failure probability is determined by uncertainties in PGA and in component seismic capacity, and the former is the crucial element to influence the result

Coming Soon: CADRE (Career Advancement and Development Resources and Education) website for all APS members

Science.gov (United States)

The Council of the American Phytopathological Society (APS) approved an initiative in February 2013 to create a web resource called CADRE (Career Advancement and Development Resources and Education). CADRE is to provide APS members an archive of articles, videos, and webinars about a variety of prof...

Overall design concepts for the APS storage ring machine protection system

International Nuclear Information System (INIS)

Lumpkin, A.; Fuja, R.; Votaw, A.; Wang, X.; Shu, D.; Stepp, J.; Arnold, N.; Nawrocki, G.; Decker, G.; Chung, Y.

1995-01-01

The basic design and status of the machine protection system for the Advanced Photon Source (APS) storage ring are discussed. The machine is passively safe to the bending magnet sources, but the high power of the insertion devices requires missteering conditions to be identified and the beam aborted in less than one millisecond. The basic aspects of waterflow, temperature, beam position, etc. monitoring are addressed. Initial commissioning of subsystems and sensors is statused

AP600 large-break loss-of-collant-accident developmental assessment plan for TRAC-PF1/MOD2

International Nuclear Information System (INIS)

Knight, T.D.

1996-07-01

The Westinghouse AP600 reactor is an advanced pressurized water reactor with passive safety systems to protect the plant against possible accidents and transients. The design has been submitted to the U.S. NRC for design certification. The NRC has selected the Transient Reactor Analysis Code (TRAC)-PF1/MOD2 for performing large break loss-of coolant-accident (LBLOCA) analysis to support the certification effort. This document defines the tests to be used in the current phase of developmental assessment related to AP600 LBLOCA

AP1000 construction schedule

International Nuclear Information System (INIS)

Winters, J.W.

2001-01-01

Westinghouse performed this study as part of EPRI interest in advancing the use of computer aided processes to reduce the cost of nuclear power plants. EPRI believed that if one could relate appropriate portions of an advanced light water reactor plant model to activities in its construction sequence, and this relationship could be portrayed visually, then optimization of the construction sequence could be developed as never before. By seeing a 3-D representation of the plant at any point in its construction sequence, more informed decisions can be made on the feasibility or attractiveness of follow on or parallel steps in the sequence. The 3-D representation of construction as a function of time (4-D) could also increase the confidence of potential investors concerning the viability of the schedule and the plant ultimate cost. This study performed by Westinghouse confirmed that it is useful to be able to visualize a plant construction in 3-D as a function of time in order to optimize the sequence of construction activities. (author)

Flow Accelerated Erosion-Corrosion (FAC) considerations for secondary side piping in the AP1000{sup R} nuclear power plant design

Energy Technology Data Exchange (ETDEWEB)

Vanderhoff, J. F.; Rao, G. V. [Westinghouse Electric Company LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States); Stein, A. [Shaw Power Nuclear, 1000 Technology Center Drive, Stoughton, MA 02072 (United States)

2012-07-01

The issue of Flow Accelerated Erosion-Corrosion (FAC) in power plant piping is a known phenomenon that has resulted in material replacements and plant accidents in operating power plants. Therefore, it is important for FAC resistance to be considered in the design of new nuclear power plants. This paper describes the design considerations related to FAC that were used to develop a safe and robust AP1000{sup R} plant secondary side piping design. The primary FAC influencing factors include: - Fluid Temperature - Pipe Geometry/layout - Fluid Chemistry - Fluid Velocity - Pipe Material Composition - Moisture Content (in steam lines) Due to the unknowns related to the relative impact of the influencing factors and the complexities of the interactions between these factors, it is difficult to accurately predict the expected wear rate in a given piping segment in a new plant. This paper provides: - a description of FAC and the factors that influence the FAC degradation rate, - an assessment of the level of FAC resistance of AP1000{sup R} secondary side system piping, - an explanation of options to increase FAC resistance and associated benefits/cost, - discussion of development of a tool for predicting FAC degradation rate in new nuclear power plants. (authors)

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

International Nuclear Information System (INIS)

Stosic, Z.V.

2006-01-01

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

Analysis for low and intermediate level radioactive waste disposal of AP1000 nuclear power in China

International Nuclear Information System (INIS)

Yang Bin; Ren Li; Hua Wei; Ma Xiaoqiang; Ma Ruoxia; Fang Xianghong

2014-01-01

AP1000 nuclear power is adopted in Sanmen Nuclear Power Plant, Zhejiang province and Haiyang Nuclear Power Plant, Shandong province. The filter and resin generated by the operation of nuclear power plants are handled by cement cured process and compression technology in Sanmen Nuclear Power Plant, while they are loaded in polyethylene HIC in Haiyang Nuclear Power Plant. At present, there were not engineering practice for dispose polyethylene HIC in China. The document discusses three disposal options of polyethylene HIC: one is that polyethylene HIC placed in over pack, another is that polyethylene HIC mixed with other drum, the third is that polyethylene HIC placed in disposal unit which is divided into small units. (authors)

7 CFR 1000.53 - Announcement of class prices, component prices, and advanced pricing factors.

Science.gov (United States)

2010-01-01

... advanced pricing factors. 1000.53 Section 1000.53 Agriculture Regulations of the Department of Agriculture..., component prices, and advanced pricing factors. (a) On or before the 5th day of the month, the market... administrator for each Federal milk marketing order shall announce the following prices and pricing factors for...

Heterogeneous LTE-Advanced Network Expansion for 1000x Capacity

DEFF Research Database (Denmark)

Hu, Liang; Sanchez, Maria Laura Luque; Maternia, Michal

2013-01-01

this paper studies LTE (Long-Term Evolution)-Advanced heterogeneous network expansion in a dense urban environment for a 1000 times capacity increase and a 10 times increase in minimum user data rate requirements. The radio network capacity enhancement via outdoor and indoor small cell densificat...

RELAP5/MOD3 AP600 problems

International Nuclear Information System (INIS)

Riemke, R.A.

1993-01-01

RELAP5/MOD3 is a reactor systems analysis code that has been developed jointly by the US Nuclear Regulatory Commission (USNRC) and a consortium consisting of several of the countries and domestic organizations that were members of the International Code Assessment and Applications Program (ICAP). The code is currently being used to simulate transients for the next generation of advanced light water reactors (ALWR's). One particular reactor design is the Westinghouse AP600 pressurized water reactor (PWR), which consists of two hot legs and four cold legs as well as passive emergency core cooling (ECC) systems. Initial calculations with RELAP5/MOD3 indicated that the code was not as robust as RELAP5/MOD2.5 with regard to AP600 calculations. Recent modifications in the areas of condensation wall heat transfer, interfacial heat transfer in the presence of noncondensibles, bubbly flow interfacial heat transfer, and time smoothing of both interfacial drag and interfacial heat transfer have improved the robustness, although more reliability is needed

A study on the economics enhancement of OPR1000 applied to advanced construction methods

International Nuclear Information System (INIS)

Park, Ki Jo; Yoon, Eun Sang

2007-01-01

OPR1000 (Optimized Power Reactor 1000MW) is a totally improved design model of Korea nuclear power plants and the latest 1,000MW nuclear power plant in the Republic of Korea. Shin Kori 1 and 2 and Shin Wolsong 1 and 2 and under construction and these are OPR1000 types. Although OPR1000 is up to data 1,000MW nuclear power plant, it is not enough to be much superior to other nuclear power plants. Under the WTO and FTA circumstance of domestic and stiff overseas competition for nuclear power plants, it is necessary to enhance the economics of OPR1000. And then, the enhanced economic alternatives are reviewed and the advanced construction methods are considered. Based on research and a comprehensive review of nuclear power plant construction experiences, an alternative application of advanced construction methods is developed and compared with existing OPR1000 for schedule and economics. In this paper, economic analyses of a construction cost and a levelized electricity generation cost are performed

Estimation of functional failure probability of passive systems based on adaptive importance sampling method

International Nuclear Information System (INIS)

Wang Baosheng; Wang Dongqing; Zhang Jianmin; Jiang Jing

2012-01-01

In order to estimate the functional failure probability of passive systems, an innovative adaptive importance sampling methodology is presented. In the proposed methodology, information of variables is extracted with some pre-sampling of points in the failure region. An important sampling density is then constructed from the sample distribution in the failure region. Taking the AP1000 passive residual heat removal system as an example, the uncertainties related to the model of a passive system and the numerical values of its input parameters are considered in this paper. And then the probability of functional failure is estimated with the combination of the response surface method and adaptive importance sampling method. The numerical results demonstrate the high computed efficiency and excellent computed accuracy of the methodology compared with traditional probability analysis methods. (authors)

Advanced configuration of hybrid passive filter for reactive power and harmonic compensation.

Science.gov (United States)

Kececioglu, O Fatih; Acikgoz, Hakan; Sekkeli, Mustafa

2016-01-01

Harmonics is one of the major power quality problems for power systems. The harmonics can be eliminated by power filters such as passive, active, and hybrid. In this study, a new passive filter configuration has been improved in addition to the existing passive filter configurations. Conventional hybrid passive filters are not successful to compensate rapidly changing reactive power demand. The proposed configure are capable of compensating both harmonics and reactive power at the same time. Simulation results show that performance of reactive power and harmonic compensation with advanced hybrid passive filter is better than conventional hybrid passive filters.

ACR-1000 design provisions for severe accidents

International Nuclear Information System (INIS)

Popov, N.K.; Santamaura, P.; Shapiro, H.; Snell, V.G.

2006-01-01

Atomic Energy of Canada Limited (AECL) developed the Advanced CANDU Reactor-700 (ACR-700) as an evolutionary advancement of the current CANDU 6 reactor. As a further advancement of the ACR design, AECL is currently developing the ACR-1000 for the Canadian and international market. The ACR-1000 is aimed at producing electrical power for a capital cost and a unit-energy cost significantly less than that of the current generation of operating nuclear plants, while achieving enhanced safety features, shorter construction schedule, high plant capacity factor, improved operations and maintenance, and increased operating life. The reference ACR-1000 plant design is based on an integrated two-unit plant, using enriched fuel and light-water coolant, with each unit having a nominal gross output of about 1200 MWe. The ACR-1000 design meets Canadian regulatory requirements and follows established international practice with respect to severe accident prevention and mitigation. This paper presents the ACR-1000 features that are designed to mitigate limited core damage and severe core damage states, including core retention within vessel, core damage termination, and containment integrity maintenance. While maintaining existing structures of CANDU reactors that provide inherent prevention and retention of core debris, the ACR-1000 design includes additional features for prevention and mitigation of severe accidents. Core retention within vessel in CANDU-type reactors includes both retention within fuel channels, and retention within the calandria vessel. The ACR-1000 calandria vessel design permits for passive rejection of decay heat from the moderator to the shield water. Also, the calandria vessel is designed for debris retention by minimizing penetrations at the bottom periphery and by accommodating thermal and weight loads of the core debris. The ACR-1000 containment is required to withstand external events such as earthquakes, tornados, floods and aircraft crashes

The link between the use of advanced planning and scheduling (APS) modules and factory context

DEFF Research Database (Denmark)

Kristensen, Jesper; Asmussen, Jesper Normann; Wæhrens, Brian Vejrum

2017-01-01

at factories characterized by low planning maturity, but lower for factories with medium planning maturity. For low planning maturity, the APS module is used for improving the configuration of the manufacturing system, whereas high planning maturity is required to capture performance benefits from optimization......Through a study of four embedded action research cases within a global OEM, it is investigated how the frequency of use and contribution of an Advanced Planning and Scheduling (APS) module are affected by factory context. The performance contribution of the APS module is found to be high...... and scenario planning. Further, it is found that planning complexity at the factory increases both the frequency of use and the contribution of using APS modules. On the basis of the findings, three propositions are formulated on the link between factory context and the use of APS module....

Thermal-hydraulic modeling needs for passive reactors

Energy Technology Data Exchange (ETDEWEB)

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

1997-07-01

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

Thermal-hydraulic modeling needs for passive reactors

International Nuclear Information System (INIS)

Kelly, J.M.

1997-01-01

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

Status and topics of thermal-hydraulic analysis for next-generation LWRs with passive safety systems

International Nuclear Information System (INIS)

Aritomi, Masanori; Ohnuki, Akira; Arai, Kenji; Kikuta, Michitaka; Yonomoto, Taisuke; Araya, Fumimasa; Akimoto, Hajime

1999-01-01

For increasing of electric power demand and reducing of carbon dioxide exhaust in the 21st century, studies of the next-generation light water reactor (LWR) with passive safety systems are developing in the world: AP-600 (by Westing House Co.); SBWR (by General Electric Co.); SWR1000 (by Siemens Co.); NP21 (by Mitsubishi Heavy Industry Co., et al.); JPSR (by JAERI). The passive equipment using natural circulation and natural convection are installed in the passive safety system, instead of active safety equipment, such as pumps, etc. It remains still as a important issue, however, to verify the reliability on the functions of the passive equipment, since that the driving forces of the passive equipment are small at comparison with the active safety equipment. The various subjects of thermal-hydraulic analysis for the next-generation light water reactors, such as temperature stratification in the passive safety systems, vapor condensation in the mixture of non-condensable gases and the interactions of the passive safety system with the primary cooling system, are illustrated and discussed in the paper. (M. Suetake)

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

International Nuclear Information System (INIS)

1991-03-01

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

Reactivity Coefficient Calculation for AP1000 Reactor Using the NODAL3 Code

Science.gov (United States)

Pinem, Surian; Malem Sembiring, Tagor; Tukiran; Deswandri; Sunaryo, Geni Rina

2018-02-01

The reactivity coefficient is a very important parameter for inherent safety and stability of nuclear reactors operation. To provide the safety analysis of the reactor, the calculation of changes in reactivity caused by temperature is necessary because it is related to the reactor operation. In this paper, the temperature reactivity coefficients of fuel and moderator of the AP1000 core are calculated, as well as the moderator density and boron concentration. All of these coefficients are calculated at the hot full power condition (HFP). All neutron diffusion constant as a function of temperature, water density and boron concentration were generated by the SRAC2006 code. The core calculations for determination of the reactivity coefficient parameter are done by using NODAL3 code. The calculation results show that the fuel temperature, moderator temperature and boron reactivity coefficients are in the range between -2.613 pcm/°C to -4.657pcm/°C, -1.00518 pcm/°C to 1.00649 pcm/°C and -9.11361 pcm/ppm to -8.0751 pcm/ppm, respectively. For the water density reactivity coefficients, the positive reactivity occurs at the water temperature less than 190 °C. The calculation results show that the reactivity coefficients are accurate because the results have a very good agreement with the design value.

76 FR 82079 - AP1000 Design Certification Amendment

Science.gov (United States)

2011-12-30

... generators, the refueling water storage tank, and various equipment for power generation, refueling, and... in settling of any resultant debris) to facilitate heat transfer to the containment vessel and for...). The source of water for the evaporative cooling is the passive containment cooling water storage tank...

Numerical simulation and experimental verification of microstructure evolution in large forged pipe used for AP1000 nuclear power plants

International Nuclear Information System (INIS)

Wang, Shenglong; Yang, Bin; Zhang, Mingxian; Wu, Huanchun; Peng, Jintao; Gao, Yang

2016-01-01

Highlights: • Establish systematically the database of 316LN stainless steel for Deform-3D. • Simulate the microstructure evolution during forging of AP1000 primary coolant pipe. • Carry out full-scale forging experiment for verification in engineering practice. • Get desirable grain size in simulation and experiment. • The variation trends of grain sizes in simulation and experiment are consistent. - Abstract: AP1000 primary coolant pipe is a large special-shaped forged pipe made of 316LN stainless steel. Due to the non-uniform temperature and deformation during its forging, coarse and fine grains usually coexist in the forged pipe, resulting in the heterogeneous microstructure and anisotropic performance. To investigate the microstructure evolution during the entire forging process, in the present research, the database of the 316LN stainless steel was established and a numerical simulation was performed. The results indicate that the middle body section of the forged pipe has an extremely uniform average grain size with the value smaller than 30 μm. The grain sizes in the ends of body sections were ranged from 30 μm to 60 μm. Boss sections have relatively homogeneous microstructure with the average grain size 30 μm to 44 μm. Furthermore, a full-scale hot forging was carried out for verification. Comparison of theoretical and experimental results showed good agreement and hence demonstrated the capabilities of the numerical simulation presented here. It is noteworthy that all grains in the workpiece were confirmed less than 180 μm, which meets the designer’s demands.

Analysis method for the design of a hydrogen mitigation system with passive autocatalytic recombiners in OPR-1000

Energy Technology Data Exchange (ETDEWEB)

Kim, C-H.; Sung, J-J.; Ha, S-J. [Korea Hydro and Nuclear Power Co. Ltd., Central Research Inst., Daejeon (Korea, Republic of); Yeo, I-S. [KEPCO Engineering and Construction Co. Ltd, Gyeonggi-do (Korea, Republic of)

2014-07-01

The importance of hydrogen safety in nuclear power plants has been emphasized especially after the Fukushima accident in Japan. A passive autocatalytic recombiner (PAR) is considered as a viable option for the mitigation of hydrogen risk because of its passive operation for hydrogen removal. This paper presents a licensed hydrogen analysis method of OPR-1000, a 1,000MWe Korea standardized pressurized water reactor with a large dry containment, to determine the capacity and locations of PARs for the design of a hydrogen mitigation system with PAR. Various accident scenarios have been adopted considering important event sequences from a combination of probabilistic methods, deterministic methods and sound engineering judgment. A MAAP 4.0.6+ with a multi-compartment model is used as an analysis tool with conservative hydrogen generation and removal models. The detailed analyses are performed for selected severe accident scenarios including sensitivity analysis with/without operations of various safety systems. The possibility of global flame acceleration (FA) and deflagration-to-detonation transient (DDT) are assessed with sigma (flame acceleration potential) and 7-lambda (DDT potential) criterion. It is concluded that the newly designed hydrogen mitigation system with twenty-four (24) PARs can effectively remove hydrogen in the containment atmosphere and prevent global FA and DDT. (author)

Advanced Reactor Passive System Reliability Demonstration Analysis for an External Event

Directory of Open Access Journals (Sweden)

Matthew Bucknor

2017-03-01

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

Advanced reactor passive system reliability demonstration analysis for an external event

Energy Technology Data Exchange (ETDEWEB)

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

2017-03-15

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

Advanced reactor passive system reliability demonstration analysis for an external event

International Nuclear Information System (INIS)

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

2017-01-01

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

APS SCIENCE 2016

Energy Technology Data Exchange (ETDEWEB)

Fenner, Richard B. [Argonne National Lab. (ANL), Argonne, IL (United States). Advanced Photon Source (APS)

2017-05-01

The Advanced Photon Source (APS) occupies an 80-acre site on the Argonne national laboratory campus, about 25 miles from downtown chicago, illinois. it shares the site with the center for nanoscale materials and the Advanced Protein characterization facility. for directions to Argonne, see http://www.anl.gov/directions-and-visitor-information. The APS, a national synchrotron radiation research facility operated by Argonne for the u.S. department of energy (doe) office of Science, provides this nation’s brightest high-energy x-ray beams for science. research by APS users extends from the center of the earth to outer space, from new information on combustion engines and microcircuits to new drugs and nanotechnologies whose scale is measured in billionths of a meter. The APS helps researchers illuminate answers to the challenges of our high-tech world, from developing new forms of energy, to sustaining our nation’s technological and economic competitiveness, to pushing back against the ravages of disease. research at the APS promises to have far-reaching

AP Music Theory Applied

Science.gov (United States)

Spieker, Matthew H.

2016-01-01

Some American high schools include Advanced Placement (AP) Music Theory within their course offerings. Students who pass the AP exam can receive college credit either as a music or humanities credit. An AP class, however, offers music students more than future college credit; it ultimately improves musicianship skills and promotes deeper…

Analysis of large scale tests for AP-600 passive containment cooling system

International Nuclear Information System (INIS)

Sha, W.T.; Chien, T.H.; Sun, J.G.; Chao, B.T.

1997-01-01

One unique feature of the AP-600 is its passive containment cooling system (PCCS), which is designed to maintain containment pressure below the design limit for 72 hours without action by the reactor operator. During a design-basis accident, i.e., either a loss-of-coolant or a main steam-line break accident, steam escapes and comes in contact with the much cooler containment vessel wall. Heat is transferred to the inside surface of the steel containment wall by convection and condensation of steam and through the containment steel wall by conduction. Heat is then transferred from the outside of the containment surface by heating and evaporation of a thin liquid film that is formed by applying water at the top of the containment vessel dome. Air in the annual space is heated by both convection and injection of steam from the evaporating liquid film. The heated air and vapor rise as a result of natural circulation and exit the shield building through the outlets above the containment shell. All of the analytical models that are developed for and used in the COMMIX-ID code for predicting performance of the PCCS will be described. These models cover governing conservation equations for multicomponents single phase flow, transport equations for the κ-ε two-equation turbulence model, auxiliary equations, liquid-film tracking model for both inside (condensate) and outside (evaporating liquid film) surfaces of the containment vessel wall, thermal coupling between flow domains inside and outside the containment vessel, and heat and mass transfer models. Various key parameters of the COMMIX-ID results and corresponding AP-600 PCCS experimental data are compared and the agreement is good. Significant findings from this study are summarized

ACR-1000: Enhanced response to severe accidents

International Nuclear Information System (INIS)

Popov, N.K.; Santamaura, P.; Shapiro, H.; Snell, V.G.

2006-01-01

Full text: Atomic Energy of Canada Limited (AECL) developed the Advanced CANDU Reactor-TM700 (ACR-700TM) as an evolutionary advancement of the current CANDU 6R reactor. As further advancement of the ACR design, AECL is currently developing the ACR-1000TM for the Canadian and international market. The ACR-1000 is aimed at producing electrical power for a capital cost and a unit-energy cost significantly less than that of the current generation of operating nuclear plants, while achieving shorter construction schedule, high plant capacity factor, improved operations and maintenance, increased operating life. and enhanced safety features. The reference ACR-1000 plant design is based on an integrated two-unit plant, using enriched fuel and light-water coolant, with each unit having a nominal gross output of about 1200 MWe. This paper presents the ACR-1000 features that are designed to mitigate limited core damage and severe core damage states, including core retention within vessel, core damage termination, and containment integrity maintenance. Core retention within vessel in CANDU-type reactors includes both retention within fuel channels, and retention within the calandria vessel. The moderator heavy water in the ACR-1000 calandria vessel, as in any other CANDU-type reactor, provides ample heat removal capacity in severe accidents. The ACR-1000 calandria vessel design permits for passive rejection of decay heat from the moderator to the shield water. Also, the calandria vessel will be designed for debris retention. Core damage termination is achieved by flooding of the core components with water and keeping them flooded thereafter. Successful termination can be achieved in the fuel channels, calandria vessel or calandria vault by water supply by the Long Term Cooling (LTC) pumps and by gravity feed from the Reserve Water System. The ACR-1000 containment is required to withstand external events such as earthquakes, tornados, floods and aircraft crashes. Containment

AP statistics crash course

CERN Document Server

D'Alessio, Michael

2012-01-01

AP Statistics Crash Course - Gets You a Higher Advanced Placement Score in Less Time Crash Course is perfect for the time-crunched student, the last-minute studier, or anyone who wants a refresher on the subject. AP Statistics Crash Course gives you: Targeted, Focused Review - Study Only What You Need to Know Crash Course is based on an in-depth analysis of the AP Statistics course description outline and actual Advanced Placement test questions. It covers only the information tested on the exam, so you can make the most of your valuable study time. Our easy-to-read format covers: exploring da

Availability analysis of the AP600 passive core cooling system

Energy Technology Data Exchange (ETDEWEB)

Syarip, M [National Atomic Energy Research Agency, Yogyakarta (Indonesia); Subki, I R [BATAN Head Office, Jakarta (Indonesia); Canton, M H [Westinghouse Electric Corp. (United States)

1996-12-01

The reliability analysis of the AP600 Passive Core Cooling System (PXS) has been done. The fault tree analysis method was used for the quantitative analysis. The PXS can be grouped to several sub-systems i.e.: Reactor Coolant System (RCS) Injection Subsystem, Emergency Core Decay Heat Removal Subsystem, and Containment Sump pH Control Subsystem. The quantitative analysis results indicates that the system unavailability is highly dependent on the valves configuration of the Automatic Depressurization System (ADS). If the ADS valves is arranged in Option-1, the system unavailability is 2.347E-03, this means that the yearly contribution to plant down time can be estimated to be about 20.56 hours per year. Whereas, by using Option-2 of fourth stage ADS valves, the system unavailability is reduced to be 9.877E-04 or 8.65 hours per year and this value is consistent with the allocated goal value (8.0 hours per year). The ADS contributes 66.89% to the system unavailability if it is arranged in Option-1, and will reduced to be about 21.21% if its fourth stages are arranged in Option-2. If the ADS is not included as a subsystem of the PXS (relocate to RCS as a subsystem of RCS), then the PXS unavailability will be reduced to about 7.784E-04 or 6.82 hours per year; this is less then the allocated goal value. The major contributors to the system unavailability are mostly dominated by Stage-4 ADS valves (air piston operated valves and squib valves), inservice testing valves of ADS (solenoid operated valves), solenoid valves of Nitrogen Supply to Accumulator, and Passive Residual Heat Removal actuation valves (air operated valves). It is recommended that those valves be analyzed more detail to gain the improvement in its reliability. It is also recommended that the fourth stage of ADS valves should be arranged according to Option-2, i.e. one 10-inch normally open motor operated gate valve in series with one 10-inch normally closed squib valve. (author). 13 refs, 3 figs, 3 tabs.

Improvement of operational performance and increase of safety of WWER-1000/V-392

International Nuclear Information System (INIS)

Kurakov, Y.A.; Dragunov, Y.G.; Podshibiakin, A.K.; Fil, N.S.; Krushelnitsky, V.N.; Berkovich, V.M.

2001-01-01

The national programme of nuclear power development approved by the Russian Federation Government in 1998 considers the design of WWER-1000/V-392 power unit as a priority project of the new generation NPP with improved operational performances and increased safety. The pilot unit of this design (NVAES-2) is licensed for construction at the Novovoronezh NPP site. The NVAES-2 design is developed on the basis of standard power unit with reactor plant V-320. Twenty units of this type are in operation at the nuclear power plants in Russia, Ukraine and Bulgaria having totally about 270 reactor-years of operation. Two more V-320 units are being commissioned this year at Rostov NPP and Temelin NPP. So, the WWER-1000/V-392 design is as a whole an evolutionary development of the operating standard unit WWER-1000/V-320. Many technical solutions aimed at increase of safety and improvement of operational performance of the plant are implemented in the NVAES-2 design, such as advanced reactor WWER-1000, passive system of residual power removal, passive system of the core flooding under loss-of-coolant accidents, and others. NVAES-2 design refers to a class of advanced light water reactors and corresponds to the international requirements imposed to the nuclear power plants to be put into operation after the year 2000. New V-392 power unit has a good perspective from the view point of extensive implementation in the framework of the nuclear electricity production in Russia. Design decisions on NVAES-2 power unit with WWER-1000/V-392 reactor plant which assure significantly higher safety level and improve economical performance as compared to the operating WWER-1000 units are briefly considered in the present paper. (author)

Preliminary analyses of AP600 using RELAP5

International Nuclear Information System (INIS)

Modro, S.M.; Beelman, R.J.; Fisher, J.E.

1991-01-01

This paper presents results of preliminary analyses of the proposed Westinghouse Electric Corporation AP600 design. AP600 is a two loop, 600 MW (e) pressurized water reactor (PWR) arranged in a two hot leg, four cold leg nuclear steam supply system (NSSS) configuration. In contrast to the present generation of PWRs it is equipped with passive emergency core coolant (ECC) systems. Also, the containment and the safety systems of the AP600 interact with the reactor coolant system and each other in a more integral fashion than present day PWRs. The containment in this design is the ultimate heat sink for removal of decay heat to the environment. Idaho National Engineering Laboratory (INEL) has studied applicability of the RELAP5 code to AP600 safety analysis and has developed a model of the AP600 for the Nuclear Regulatory Commission. The model incorporates integral modeling of the containment, NSSS and passive safety systems. Best available preliminary design data were used. Nodalization sensitivity studies were conducted to gain experience in modeling of systems and conditions which are beyond the applicability of previously established RELAP5 modeling guidelines or experience. Exploratory analyses were then undertaken to investigate AP600 system response during postulated accident conditions. Four small break LOCA calculations and two large break LOCA calculations were conducted

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

International Nuclear Information System (INIS)

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

2014-01-01

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

Charged-particle beam diagnostics for the Advanced Photon Source (APS)

International Nuclear Information System (INIS)

Lumpkin, A.H.; Decker, G.; Kahana, E.; Patterson, D.; Sellyey, W.; Wang, X.; Chung, Y.

1992-01-01

Plans, prototypes, and initial test results for the charged-particle beam (e - , e + ) diagnostic systems on the injector rings, their transport lines, and the storage ring for the Advanced Photon Source (APS) are presented. The APS will be a synchrotron radiation user facility with one of the world's brightest x-ray sources in the 10-keV to 100-keV regime. Its 200-MeV electron linac, 450-MeV positron linac, positron accumulator ring, 7-GeV booster synchrotron, 7-GeV storage ring, and undulator test lines will also demand the development and demonstration of key particle-beam characterization techniques over a wide range of parameter space. Some of these parameter values overlap or approach those projected for fourth generation light sources (linac-driven FELs and high brightness storage rings) as described at a recent workshop. Initial results from the diagnostics prototypes on the linac test stand operating at 45-MeV include current monitor data, beam loss monitor data, and video digitization using VME architecture

Estimation of functional failure probability of passive systems based on subset simulation method

International Nuclear Information System (INIS)

Wang Dongqing; Wang Baosheng; Zhang Jianmin; Jiang Jing

2012-01-01

In order to solve the problem of multi-dimensional epistemic uncertainties and small functional failure probability of passive systems, an innovative reliability analysis algorithm called subset simulation based on Markov chain Monte Carlo was presented. The method is found on the idea that a small failure probability can be expressed as a product of larger conditional failure probabilities by introducing a proper choice of intermediate failure events. Markov chain Monte Carlo simulation was implemented to efficiently generate conditional samples for estimating the conditional failure probabilities. Taking the AP1000 passive residual heat removal system, for example, the uncertainties related to the model of a passive system and the numerical values of its input parameters were considered in this paper. And then the probability of functional failure was estimated with subset simulation method. The numerical results demonstrate that subset simulation method has the high computing efficiency and excellent computing accuracy compared with traditional probability analysis methods. (authors)

Local panels and maintainability human factors assessment for AP1000 nuclear power plant

International Nuclear Information System (INIS)

Li, Zhonghai; Reed, Julie I.

2011-01-01

A document entitled 'AP1000 Local Panels and Maintainability Human Factors Design Guidelines' was produced to aid the designers to specifically include human factors (HF) considerations in the design, operation, and maintenance of local control stations and plant equipment. To ensure that the applicable HF design guidelines are appropriately applied to the design of local panels and maintenance activities, and identify any HF improvement opportunities that can readily be implemented at the design stage, a HF assessment of maintenance activities and local plant operations is underway. This assessment gives priority to local control stations and equipment which have been identified as having a potential impact on safety. This includes risk-significant systems, structures and components (SSCs) identified through the probabilistic risk assessment (PRA), and local operator actions as required by the Emergency Operating Procedures (EOPs). Local actions, maintenance activities and associated operator interfaces are reviewed against the relevant HF guidelines. The results of the assessment include a description of the component, associated local actions and/or required maintenance activities, good design features and/or potential issues, and recommendations for change or improvement. These results are communicated to responsible design engineers who evaluate the impact to plant design and implement design changes, if deemed necessary. (author)

Engineering reliability in design phase: An application to AP-600 reactor passive safety system

International Nuclear Information System (INIS)

Majumdr, D.; Siahpush, A.S.; Hills, S.W.

1992-01-01

A computerized reliability enhancement methodology is described that can be used at the engineering design phase to help the designer achieve a desired reliability of the system. It can take into account the limitation imposed by a constraint such as budget, space, or weight. If the desired reliability of the system is known, it can determine the minimum reliabilities of the components, or how many redundant components are needed to achieve the desired reliability. This methodology is applied to examine the Automatic Depressurization System (ADS) of the new passively safe AP-600 reactor. The safety goal of a nuclear reactor dictates a certain reliability level of its components. It is found that a series parallel valve configuration instead of the parallel-series configuration of the four valves in one stage would improve the reliability of the ADS. Other valve characteristics and arrangements are explored to examine different reliability options for the system

Research on the improvement design for the attachment of supports to AP1000 module wall

International Nuclear Information System (INIS)

Li Cheng; Liu Jianwei; Shan Ying

2013-01-01

Background: Modularization is one of the main characteristics for AP1000 nuclear power plant building. The steel-concrete-steel module wall is used instead of reinforced concrete structure wall. Usually, lots of Overlay Plate Embedments will be installed on the module wall to connect and fasten other structures, such as pipes, equipment and operation platforms. As for many supports taking less design loads, the safety margin is too big when using OLP embedment. Purpose: An improvement design will make sense that the supports with less design loads can be welded directly to the module wall instead of embedments. Methods: A finite element analysis based on nuclear-related concrete code is carried out. Results: Through analysis, the equations for the allowable design loads of supports to be welded directly to module wall are provided in this paper. Conclusions: The improvement design is proved feasible. In this way, the strength for steel face plate and studs will be utilized fully and this method will facilitate and simplify the design and construction with considerable engineering application value. (authors)

Seismic dynamic analysis of Heat Exchangers inside of the Auxiliary Buildings in AP1000TM NPP

International Nuclear Information System (INIS)

Di Fonzo, M.; Aragon, J.; Moraleda, F.; Palazuelos, M.; San Vicente, J. L.

2011-01-01

Seismic dynamic analysis was carried out for the Heat Exchangers (RNS-HR) located inside of the Auxiliary Building in AP 1000 T M NPP. The main function of the RNS-HX is to provide shutdown reactor cooling. These equipment's are safety-related. So the seismic analysis was done using the methodology for Seismic Category I (SCI) structures. The most important topic is that the RNS-HX shall withstand the effects of the Safe Shutdown Earthquake (SSE) and maintain the specified design functions. for the analysis, two finite element models (FEM) were built in order to investigate the structural response of the couple system of building and equipment. The response spectra method was used. The floor response spectra (FRS) at the slab-wall connection were used as input Lateral seismic restrain was necessary to added in order to achieve the natural frequency of 33 Hz. The global structural response was obtained by means of the modal combination method indicated in the Regulatory Guide 1.92.

SPES-2, the full-height, full-pressure, test facility simulating the AP600 plant: Main results from the experimental campaign

International Nuclear Information System (INIS)

Medich, C.; Rigamonti, M.; Martinelli, R.; Tarantini, M.; Conway, L.

1995-01-01

The SPES-2 is a full height, full pressure experimental test facility reproducing the Westinghouse AP600 reactor with a scaling factor of 1/395. The experimental plant, designed and operated by SIET in Piacenza, consists of a full simulation of the AP600 primary core cooling system including all the passive and active safety systems. In 1992, Westinghouse, in cooperation with ENEL, ENEA, SIET and ANSALDO developed an experimental program to test the integrated behavior of the AP600 passive safety systems. The SPES-2 test matrix, concluded in November 1994, has examined the AP600 passive safety system response for a range of small break LOCAs at different locations on the primary system and on the passive system lines; single steam generator tube ruptures with both passive and active non-safety systems, and a main steam line break transient to demonstrate the capability of passive safety systems for rapid cooldown. Each of the tests has provided detailed experimental results for verification of the capability of the analysis methods to predict the integrated passive safety system behavior

Commercializing the next generation: the AP600 advanced simplified nuclear power plant

International Nuclear Information System (INIS)

Bruschi, H.J.

1994-01-01

Today, government and industry are working together on advanced nuclear power plant designs that take advantage of valuable lessons learned from the experience to date and promise to reconcile the demands of economic expansion with the laws of environmental protection. In the U.S., the Department of Energy (DOE) and the Electric Power Research Institute (EPRI) initiated a design certification program in 1989 to develop and commercialize advanced light water reactors (ALWRs) for the next round of power plant construction. Advanced, simplified technology is one approach under development to end the industry's search for a simpler, more forgiving, and less costly reactor. As part of this program, Westinghouse is developing the AP600, a new standard 600 MWe advanced, simplified plant. The design strikes a balance between the use of proven technology and new approaches. The result is a greatly streamlined plant that can meet safety regulations and reliability requirements, be economically competitive, and promote broader public confidence in nuclear energy. 1 fig

25 CFR 1000.52 - What criteria will the Director use to award advance planning grants?

Science.gov (United States)

2010-04-01

... 25 Indians 2 2010-04-01 2010-04-01 false What criteria will the Director use to award advance... Planning Grant Funding § 1000.52 What criteria will the Director use to award advance planning grants? Advance planning grants are discretionary and based on need. The Director will use the following criteria...

A Quantitative Feasibility Study on Potential Safety Improvement Effects of Advanced Safety Features in APR-1400 when Applied to OPR-1000

Energy Technology Data Exchange (ETDEWEB)

Ualikhan Zhiyenbayev [KAIST, Daejeon (Korea, Republic of); Chung, Dae Wook [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2015-10-15

This study aims to test the feasibility of the applications using Probabilistic Safety Assessment (PSA). Particularly, three of those advanced safety features are selected as follows: 1. Providing an additional Emergency Diesel Generator (EDG); 2. Increasing the capacity of Class 1E batteries; 3. Placing a Refueling Water Storage Tank (RWST) inside containment, i.e., change from RWST to IRWST. The Advanced Power Reactor 1400 (APR-1400) adopts several advanced safety features compared to its predecessor, the Optimized Power Reactor 1000 (OPR-1000), which includes an additional Emergency Diesel Generator, increase in battery capacity, in-containment refueling water storage tank (IRWST), and so on. Considering the remarkable advantages of these safety features in safety improvement and the design similarities between APR-1400 and OPR-1000, it is feasible to apply key advanced safety features of APR-1400 to OPR-1000 to enhance the safety. The selected safety features are incorporated into OPR-1000 PSA model using the Advanced Information Management System (AIMS) for PSA and CDFs are re-evaluated for each application and combination of three applications. Based on current results, it is concluded that three of key advanced safety features of APR-1400 can be effectively applied to OPR-1000, resulting in considerable safety improvement. In aggregate, three advanced safety features, which are an additional EDG, increased battery capacity and IRWST, can reduce the CDF of OPR-1000 by more than 15% when applied altogether.

A Quantitative Feasibility Study on Potential Safety Improvement Effects of Advanced Safety Features in APR-1400 when Applied to OPR-1000

International Nuclear Information System (INIS)

Ualikhan Zhiyenbayev; Chung, Dae Wook

2015-01-01

This study aims to test the feasibility of the applications using Probabilistic Safety Assessment (PSA). Particularly, three of those advanced safety features are selected as follows: 1. Providing an additional Emergency Diesel Generator (EDG); 2. Increasing the capacity of Class 1E batteries; 3. Placing a Refueling Water Storage Tank (RWST) inside containment, i.e., change from RWST to IRWST. The Advanced Power Reactor 1400 (APR-1400) adopts several advanced safety features compared to its predecessor, the Optimized Power Reactor 1000 (OPR-1000), which includes an additional Emergency Diesel Generator, increase in battery capacity, in-containment refueling water storage tank (IRWST), and so on. Considering the remarkable advantages of these safety features in safety improvement and the design similarities between APR-1400 and OPR-1000, it is feasible to apply key advanced safety features of APR-1400 to OPR-1000 to enhance the safety. The selected safety features are incorporated into OPR-1000 PSA model using the Advanced Information Management System (AIMS) for PSA and CDFs are re-evaluated for each application and combination of three applications. Based on current results, it is concluded that three of key advanced safety features of APR-1400 can be effectively applied to OPR-1000, resulting in considerable safety improvement. In aggregate, three advanced safety features, which are an additional EDG, increased battery capacity and IRWST, can reduce the CDF of OPR-1000 by more than 15% when applied altogether

Design and transient analyses of emergency passive residual heat removal system of CPR1000

International Nuclear Information System (INIS)

Zhang, Y.P.; Qiu, S.Z.; Su, G.H.; Tian, W.X.

2012-01-01

Highlights: ► Designing an EPRHRs for CPR1000. ► Developing a RELAP model of the EPRHRs. ► The EPRHRs could take away the decay heat effectively. - Abstract: The steam generator secondary emergency passive residual heat removal system (EPRHRs) is a new design for traditional generation II + reactor CPR1000. The EPRHRs is designed to improve the safety and reliability of CPR1000 by completely or partially replacing traditional emergency water cooling system in the event of the station blackout or loss of heat sink accident. The EPRHRs consists of steam generator (SG), heat exchanger (HX), emergency makeup tank (EMT), cooling water tank (CWT), and corresponding pipes and valves. In order to improve the safety and reliability of CPR1000, the model of the primary loop and the EPRHRs was developed to investigate residual heat removal capability of the EPRHRs and the transient characteristics of the primary loop affected by the EPRHRs using RELAP5/MOD3.4. The transient characteristics of the primary loop and the EPRHRs were calculated in the event of station blackout accident. Sensitivity studies of the EPRHRs were also conducted to investigate the response of the primary loop and the EPRHRs on the main parameters of the EPRHRs. The EPRHRs could supply water to the SG shell side from the EMT successfully. The calculation results showed that the EPRHRs could take away the decay heat from the primary loop effectively, and that the single-phase and two-phase natural circulations were established in the primary loop and EPRHRs loop, respectively. The results also indicated that the effect of isolation valve open time on the transient characteristics of the primary loop was little. However, the effect of isolation valve open time on the EPRHRs condensate flow was relatively greater. The isolation valves should not be opened too rapidly during the isolation valve opening process, and the isolation valve opening time should be greater than 10 s, which could avoid the

APS Science 2006

International Nuclear Information System (INIS)

Gibson, J.M.; Fenner, R.B.; Long, G.; Borland, M.; Decker, G.

2007-01-01

In my five years as the Director of the Advanced Photon Source (APS), I have been fortunate to see major growth in the scientific impact from the APS. This year I am particularly enthusiastic about prospects for our longer-term future. Every scientific instrument must remain at the cutting edge to flourish. Our plans for the next generation of APS--an APS upgrade--got seriously in gear this year with strong encouragement from our users and sponsors. The most promising avenue that has emerged is the energy-recovery linac (ERL) (see article on page xx), for which we are beginning serious R and D. The ERL(at)APS would offer revolutionary performance, especially for x-ray imaging and ultrafast science, while not seriously disrupting the existing user base. I am very proud of our accelerator physics and engineering staff, who not only keep the current APS at the forefront, but were able to greatly impress our international Machine Advisory Committee with the quality of their work on the possible upgrade option (see page xx). As we prepare for long-term major upgrades, our plans to develop and optimize all the sectors at APS in the near future are advancing. Several new beamlines saw first light this year, including a dedicated powder diffraction beamline (11-BM), two instruments for inelastic x-ray scattering at sector 30, and the Center for Nanoscale Materials (CNM) Nanoprobe beamline at sector 26. Our partnership in the first x-ray free-electron laser (LCLS) to be built at Stanford contributes to revolutionary growth in ultrafast science (see page xx), and we are developing a pulse chirping scheme to get ps pulses at sector 7 of the APS within a year or so. In this report, you will find selected highlights of scientific research at the APS from calendar year 2006. The highlighted work covers diverse disciplines, from fundamental to applied science. In the article on page xx you can see the direct impact of APS research on technology. Several new products have emerged

APS Science 2006.

Energy Technology Data Exchange (ETDEWEB)

Gibson, J. M.; Fenner, R. B.; Long, G.; Borland, M.; Decker, G.

2007-05-24

In my five years as the Director of the Advanced Photon Source (APS), I have been fortunate to see major growth in the scientific impact from the APS. This year I am particularly enthusiastic about prospects for our longer-term future. Every scientific instrument must remain at the cutting edge to flourish. Our plans for the next generation of APS--an APS upgrade--got seriously in gear this year with strong encouragement from our users and sponsors. The most promising avenue that has emerged is the energy-recovery linac (ERL) (see article on page xx), for which we are beginning serious R&D. The ERL{at}APS would offer revolutionary performance, especially for x-ray imaging and ultrafast science, while not seriously disrupting the existing user base. I am very proud of our accelerator physics and engineering staff, who not only keep the current APS at the forefront, but were able to greatly impress our international Machine Advisory Committee with the quality of their work on the possible upgrade option (see page xx). As we prepare for long-term major upgrades, our plans to develop and optimize all the sectors at APS in the near future are advancing. Several new beamlines saw first light this year, including a dedicated powder diffraction beamline (11-BM), two instruments for inelastic x-ray scattering at sector 30, and the Center for Nanoscale Materials (CNM) Nanoprobe beamline at sector 26. Our partnership in the first x-ray free-electron laser (LCLS) to be built at Stanford contributes to revolutionary growth in ultrafast science (see page xx), and we are developing a pulse chirping scheme to get ps pulses at sector 7 of the APS within a year or so. In this report, you will find selected highlights of scientific research at the APS from calendar year 2006. The highlighted work covers diverse disciplines, from fundamental to applied science. In the article on page xx you can see the direct impact of APS research on technology. Several new products have emerged from

Advanced Photon Source experimental beamline Safety Assessment Document: Addendum to the Advanced Photon Source Accelerator Systems Safety Assessment Document (APS-3.2.2.1.0)

International Nuclear Information System (INIS)

1995-01-01

This Safety Assessment Document (SAD) addresses commissioning and operation of the experimental beamlines at the Advanced Photon Source (APS). Purpose of this document is to identify and describe the hazards associated with commissioning and operation of these beamlines and to document the measures taken to minimize these hazards and mitigate the hazard consequences. The potential hazards associated with the commissioning and operation of the APS facility have been identified and analyzed. Physical and administrative controls mitigate identified hazards. No hazard exists in this facility that has not been previously encountered and successfully mitigated in other accelerator and synchrotron radiation research facilities. This document is an updated version of the APS Preliminary Safety Analysis Report (PSAR). During the review of the PSAR in February 1990, the APS was determined to be a Low Hazard Facility. On June 14, 1993, the Acting Director of the Office of Energy Research endorsed the designation of the APS as a Low Hazard Facility, and this Safety Assessment Document supports that designation

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

International Nuclear Information System (INIS)

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

2014-01-01

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

Simulation of fission products behavior in severe accidents for advanced passive PWR

International Nuclear Information System (INIS)

Tong, L.L.; Huang, G.F.; Cao, X.W.

2015-01-01

Highlights: • A fission product analysis model based on thermal hydraulic module is developed. • An assessment method for fission product release and transport is constructed. • Fission products behavior during three modes of containment response is investigated. • Source term results for the three modes of containment response are obtained. - Abstract: Fission product behavior for common Pressurized Water Reactor (PWR) has been studied for many years, and some analytical tools have developed. However, studies specifically on the behavior of fission products related to advanced passive PWR is scarce. In the current study, design characteristics of advanced passive PWR influencing fission product behavior are investigated. An integrated fission products analysis model based on a thermal hydraulic module is developed, and the assessment method for fission products release and transport for advanced passive PWR is constructed. Three modes of containment response are simulated, including intact containment, containment bypass and containment overpressure failure. Fission products release from the core and corium, fission products transport and deposition in the Reactor Coolant System (RCS), fission products transport and deposition in the containment considering fission products retention in the in-containment refueling water storage tank (IRWST) and in the secondary side of steam generators (SGs) are simulated. Source term results of intact containment, containment bypass and containment overpressure failure are obtained, which can be utilized to evaluate the radiological consequences

AP English language & composition crash course

CERN Document Server

Hogue, Dawn

2012-01-01

AP English Language & Composition Crash Course - Gets You a Higher Advanced Placement Score in Less Time Crash Course is perfect for the time-crunched student, the last-minute studier, or anyone who wants a refresher on the subject. AP English Language & Composition Crash Course gives you: Targeted, Focused Review - Study Only What You Need to Know Crash Course is based on an in-depth analysis of the AP English Language & Composition course description outline and actual Advanced Placement test questions. It covers only the information tested on the exam, so you can make the most of your valua

Multipacting study of the RF window at the Advanced Photon Source (APS)

International Nuclear Information System (INIS)

Song, J. J.

1999-01-01

Multipacting current can cause breakdowns in high power rf components such as input couplers, waveguide windows, and higher-order mode (HOM) dampers. To understand and prevent the loss of a ceramic window or an input coupler in the Advanced Photon Source (APS) storage ring rf cavity, the multipacting phenomenon is being investigated experimentally. This paper begins with a description of simple model, presents a hardware design, and concludes with measurement of multipacting. Multipacting is explored in conjunction with conditioning the cavities and interaction with the stored beam

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

International Nuclear Information System (INIS)

Atkinson, S.A.

1989-01-01

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

Advanced light-water reactors

International Nuclear Information System (INIS)

Golay, M.W.; Todreas, N.E.

1990-01-01

Environmental concerns, economics and the earth's finite store of fossil fuels argue for a resuscitation of nuclear power. The authors think improved light-water reactors incorporating passive safety features can be both safe and profitable, but only if attention is paid to economics, effective management and rigorous training methods. The experience of nearly four decades has winnowed out designs for four basic types of reactor: the heavy-water reactor (HWR), the gas-cooled rector (GCR), the liquid-metal-cooled reactor (LMR) and the light-water reactor (LWR). Each design is briefly described before the paper discusses the passive safety features of the AP-600 rector, so-called because it employs an advanced pressurized water design and generates 600 MW of power

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

International Nuclear Information System (INIS)

Huang Xueqing; Zhang Senru

1999-01-01

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

NRC confirmatory safety system testing in support of AP600 design review

International Nuclear Information System (INIS)

Rhee, G.S.; Bessette, D.E.; Shotkin, L.M.

1994-01-01

Westinghouse Electric Corporation has submitted the Advanced Passive 600 MWe (AP600) nuclear power plant design to the NRC for design certification. The Office of Nuclear Regulatory Research is proceeding to conduct confirmatory testing to help the NRC staff evaluate the AP600 safety system design. For confirmatory testing, it was determined that the cost-effective route was to modify an existing full-height, full-pressure test facility rather than build a new one. Thus, all the existing integral effects test facilities, both in the US and abroad, were screened to select the best candidate. As a result, the ROSA-V (Rig of Safety Assessment-V) test facility located in the Japan Atomic Energy Research Institute (JAERI) was chosen. However, because of some differences in design between the existing ROSA-V facility and the AP600, the ROSA-V is being modified to conform to the AP600 safety system design. The modification work will be completed by the end of this year. A series of facility characterization tests will then be performed in January 1994 for the modified part of the facility before the main test series is initiated in February 1994. A total of 12 tests will be performed in 1994 under Phase I of this cooperative program with JAERI. Phase II testing is being considered to be conducted in 1995 mainly for beyond-design-basis accident evaluation

Analysis of neutronic parameters of AP1000 core for 18 month and 16/20 month cycle schemes using CASMO4E and SIMULATE-3 codes

International Nuclear Information System (INIS)

Nawaz Amjad; Yoshikawa, Hidekazu; Ming Yang

2015-01-01

AP1000 reactor is designed for 18 month of operating cycle. The core can also be used for 16/20 months of operating cycle. This study is performed to analyze and compare the neutronic parameters of typical AP1000 reactor core for 18 month and 16/20 month alternate cycle lengths. CASMO4E and SIMULATE-3 code package is used for the analysis of initial and equilibrium cores. The key reactor physics safety parameters were analyzed including power peaking factors, core radial and axial power distribution and core reactivity feedback coefficients. Moreover, the analysis of fuel depletion, fission product buildup and burnable poison behaviour with burnup is also analyzed. Full 2-D fuel assembly model in CASMO4E and full 3-D core model in SIMULATE-3 is employed to examine core performance and safety parameters. In order to evaluate the equilibrium core neutronic parameters, the equilibrium core model is attained by performing burnup analysis from initial to equilibrium cycle, where optimized transition core design is obtained so that the power peaking factors remain within designed limits. The MTC for higher concentration of critical boron concentrations is slightly positive at lower moderator temperatures. However, it remains negative at operating temperature ranges. The radial core relative power distribution indicates that low leakage capability of initial and equilibrium cores is reduced at EOC. (author)

Loss-of-normal-feedwater sensitivity studies for AP600 behavior characterization

International Nuclear Information System (INIS)

Saiu, G.

1996-01-01

Activity concerning the development of a RELAP5/MOD3 model to simulate the Westinghouse Electric Corporation AP600 is summarized. The aim is to gain initial insight into the capability of RELAP5 to simulate the behavior of AP600 safety features. A-loss-of-normal-feedwater event is studied. Of the transients that must be investigated, this transient has been chosen to be one of the most relevant because the response of the AP600 to a loss-of-normal-feedwater event differs significantly from that of current pressurized water reactors in the extensive use of passive safety features peculiar to the AP600. Also, strong interactions among the AP600 safety systems, which should be further analyzed to permit full optimization of the system actuation logic and operation, are shown. Finally, a loss of normal feedwater without reactor scram, performed to investigate short-term plant behavior, shows that the pressure peak is affected by critical discharge flow coefficients applied to the pressurizer safety valves, while a relatively small reduction of the pressure peak is observed when both heat exchangers of the passive heat removal system are operating as opposed to the case in which only one is available. The data used for this study are derived from the Standard Safety Analysis Report configuration of the Westinghouse AP600 as of 1992

APS Science 2009.

Energy Technology Data Exchange (ETDEWEB)

Gibson, J. M; Mills, D. M.; Gerig, R.

2010-05-01

It is my pleasure to introduce the 2009 annual report of the Advanced Photon Source. This was a very good year for us. We operated with high reliability and availability, despite growing problems with obsolete systems, and our users produced a record output of publications. The number of user experiments increased by 14% from 2008 to more than 3600. We congratulate the recipients of the 2009 Nobel Prize in Chemistry-Venkatraman Ramakrishnan (Cambridge Institute for Medical Research), Thomas Steitz (Yale University), and Ada Yonath (Weizmann Institute) - who did a substantial amount of this work at APS beamlines. Thanks to the efforts of our users and staff, and the ongoing counsel of the APS Scientific Advisory Committee, we made major progress in advancing our planning for the upgrade of the APS (APS-U), producing a proposal that was positively reviewed. We hope to get formal approval in 2010 to begin the upgrade. With advocacy from our users and the support of our sponsor, the Office of Basic Energy Sciences in the Department of Energy (DOE) Office of Science, our operating budgets have grown to the level needed to more adequately staff our beamlines. We were also extremely fortunate to have received $7.9 M in American Recovery and Reinvestment Act ('stimulus') funding to acquire new detectors and improve several of our beamlines. The success of the new Linac Coherent Light Source at Stanford, the world's first x-ray free-electron laser, made us particularly proud since the undulators were designed and built by the APS. Among other highlights, we note that more than one-quarter of the 46 Energy Frontier Research Centers, funded competitively across the U.S. in 2009 by the DOE, included the Advanced Photon Source in their proposed work, which shows that synchrotron radiation, and the APS in particular, are central to energy research. While APS research covers everything from fundamental to applied science (reflected by the highlights in this report

Passive Solar still: Recent advancement in design and related Performance.

Science.gov (United States)

Awasthi, Anuradha; Kumari, Kanchan; Panchal, Hitesh; Sathyamurthy, Ravishankar

2018-05-31

Present review paper mainly focuses on different varieties of solar stills and highlights mostly the passive solar still with advanced modifications in the design and development of material, single and multi-effect solar still with augmentation of different materials, energy absorbing, insulators, mechanisms of heat and mass transfer to improve the loss of heat and enhance the productivity of solar still. The cost-benefit analysis along with the progressive advancement for solar stills is the major highlights of this review. To increase the output of solar still nowadays, applications of advance modifications is one of the promising tools, and it is anticipated that shortly more vigor will be added in this area with the modifications in designs of solar stills.

Experimental research on passive residual heat remove system for advanced PWR

International Nuclear Information System (INIS)

Huang Yanping; Zhuo Wenbin; Yang Zumao; Xiao Zejun; Chen Bingde

2003-01-01

The experimental and qualified results of MISAP in the research of passive residual heat remove system of advanced PWR performed in the Bubble physics and natural circulation laboratory in Nuclear Power Institute of China in the past ten years is overviewed. Further researches for engineering research and design are also suggested

Seismic dynamic analysis of Heat Exchangers inside of the Auxiliary Buildings in AP1000{sup T}M NPP

Energy Technology Data Exchange (ETDEWEB)

Di Fonzo, M.; Aragon, J.; Moraleda, F.; Palazuelos, M.; San vicente, J. L.

2011-07-01

Seismic dynamic analysis was carried out for the Heat Exchangers (RNS-HR) located inside of the Auxiliary Building in AP 1000{sup T}M NPP. The main function of the RNS-HX is to provide shutdown reactor cooling. These equipment's are safety-related. So the seismic analysis was done using the methodology for Seismic Category I (SCI) structures. The most important topic is that the RNS-HX shall withstand the effects of the Safe Shutdown Earthquake (SSE) and maintain the specified design functions. for the analysis, two finite element models (FEM) were built in order to investigate the structural response of the couple system of building and equipment. The response spectra method was used. The floor response spectra (FRS) at the slab-wall connection were used as input Lateral seismic restrain was necessary to added in order to achieve the natural frequency of 33 Hz. The global structural response was obtained by means of the modal combination method indicated in the Regulatory Guide 1.92.

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-10-15

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

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

International Nuclear Information System (INIS)

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

2014-01-01

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

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

International Nuclear Information System (INIS)

Wolff, U.; Sliter, G.

2004-01-01

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

AP600 containment purge radiological analysis

Energy Technology Data Exchange (ETDEWEB)

O`Connor, M.; Schulz, J.; Tan, C. [Bechtel Power Corporation (United States)] [and others

1995-02-01

The AP600 Project is a passive pressurized water reactor power plant which is part of the Design Certification and First-of-a-Kind Engineering effort under the Advanced Light Water Reactor program. Included in this process is the design of the containment air filtration system which will be the subject of this paper. We will compare the practice used by previous plants with the AP600 approach to meet the goals of industry standards in sizing the containment air filtration system. The radiological aspects of design are of primary significance and will be the focus of this paper. The AP600 Project optimized the design to combine the functions of the high volumetric flow rate, low volumetric flow rate, and containment cleanup and other filtration systems into one multi-functional system. This achieves a more simplified, standardized, and lower cost design. Studies were performed to determine the possible concentrations of radioactive material in the containment atmosphere and the effectiveness of the purge system to keep concentrations within 10CFR20 limits and within offsite dose objectives. The concentrations were determined for various reactor coolant system leakage rates and containment purge modes of operation. The resultant concentrations were used to determine the containment accessibility during various stages of normal plant operation including refueling. The results of the parametric studies indicate that a dual train purge system with a capacity of 4,000 cfm per train is more than adequate to control the airborne radioactivity levels inside containment during normal plant operation and refueling, and satisfies the goals of ANSI/ANS-56.6-1986 and limits the amount of radioactive material released to the environment per ANSI/ANS 59.2-1985 to provide a safe environment for plant personnel and offsite residents.

Quality surveillance for steel forgings of SA508 Gr.3 used on the main NI equipment of AP1000 nuclear island

International Nuclear Information System (INIS)

Liu Lizhao

2011-01-01

Being a type of steel with ideal weldability, outstanding ability of anti-neutron irradiation embitterment and good property of fracture toughness and impact toughness, the steel of ASME SA508-3 was used widely for the nuclear island equipment of PWR Nuclear Power Plant. For the 3rd generation nuclear power plant AP1000, all large forgings and some critical components of the SG, RV and PRZ adopt this steel. Through analysis on the critical technical points during manufacturing of the SA508-3 forgings, this article try to identify the key points should be paid attention during the quality surveillance for this type of forgings, and to put forward the supervision method and focus during quality surveillance activities. (author)

Advances in passive-remote and extractive Fourier transform infrared spectroscopic systems

International Nuclear Information System (INIS)

Demirgian, J.C.; Hammer, C.; Hwang, E.; Mao, Zhuoxiong.

1993-01-01

The Clean Air Act of 1990 requires the monitoring of air toxics including those from incinerator emissions. Continuous emission monitors (CEM) would demonstrate the safety of incinerators and address public concern about emissions of hazardous organic compounds. Fourier transform infrared (FTIR) spectroscopy can provide the technology for continuous emission monitoring of stacks. Stack effluent can be extracted and analyzed in less than one minute with conventional FTIR spectrometers. Passive-remote FTIR spectrometers can detect certain emission gases over 1 km away from a stack. The authors discuss advances in both extractive and passive-remote FTIR technology. Extractive systems are being tested with EPA protocols, which will soon replace periodic testing methods. Standard operating procedures for extractive systems are being developed and tested. Passive-remote FTIR spectrometers have the advantage of not requiring an extracted sample; however, they have less sensitivity. We have evaluated the ability of commercially available systems to detect fugitive plumes and to monitor carbon monoxide at a coal-fired power plant

Development of a superconducting undulator for the APS

International Nuclear Information System (INIS)

Ivanyushenkov, Y; Abliz, M; Doose, C; Fuerst, J; Hasse, Q; Kasa, M; Trakhtenberg, E; Vasserman, I; Gluskin, E; Lev, V; Mezentsev, N; Syrovatin, V; Tsukanov, V

2013-01-01

As the western hemisphere's premier x-ray synchrotron radiation source, the Advanced Photon Source (APS) continues to advance the state of the art in insertion device technology in order to maintain record high brightness, especially in the hard x-ray wavelength region. Due to the unique bunch pattern used for normal APS operations and its ultimate capabilities, the APS has chosen superconducting technology for its future hard x-ray undulator sources. In the last several years, the APS in collaboration with the Budker Institute of Nuclear Physics has being developing the technology for planar, small-period superconducting undulators (SCUs). These developments include the design and construction of several prototypes and the construction of the necessary mechanical, vacuum, and cryogenic infrastructure at the APS site. Several prototypes of the SCU magnetic structure have been built and tested. The first SCU is assembled and will be installed in the APS storage ring at the end of 2012. Expected SCU performance in terms of x-ray brightness should noticeably exceed that of existing APS undulators. Immediately after commissioning, the SCU will be used at APS Sector 6 as the radiation source for high-energy x-ray studies.

SWR 1000 severe accident control through in-vessel melt retention by external RPV cooling

Energy Technology Data Exchange (ETDEWEB)

Kolev, N.I. [Framatome Advanced Nuclear Power, NDSI, Erlangen (Germany)

2001-07-01

Framatome Advanced Nuclear Power is being designing a new generation NPP with boiling water reactor SWR1000. Besides of various of modern passive and active safety features the system is also designed for controlling of a postulated severe accident with extreme low probability of occurrence. This work presents the rationales behind the decision to select the external cooling as a safety management strategy during severe accident. Bounding scenery are analyzed regarding the core melting, melt-water interaction during relocation of the melt from the core region into the lower head and the external coolability of the lower head. The conclusion is reached that the external cooling for the SWR1000 is a valuable strategy for accident management during postulated severe accidents. (authors)

SWR 1000 severe accident control through in-vessel melt retention by external RPV cooling

International Nuclear Information System (INIS)

Kolev, N.I.

2001-01-01

Framatome Advanced Nuclear Power is being designing a new generation NPP with boiling water reactor SWR1000. Besides of various of modern passive and active safety features the system is also designed for controlling of a postulated severe accident with extreme low probability of occurrence. This work presents the rationales behind the decision to select the external cooling as a safety management strategy during severe accident. Bounding scenery are analyzed regarding the core melting, melt-water interaction during relocation of the melt from the core region into the lower head and the external coolability of the lower head. The conclusion is reached that the external cooling for the SWR1000 is a valuable strategy for accident management during postulated severe accidents. (authors)

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.)

Application of the LBB regulatory approach to the steamlines of advanced WWER 1000 reactor

Energy Technology Data Exchange (ETDEWEB)

Kiselyov, V.A.; Sokov, L.M.

1997-04-01

The LBB regulatory approach adopted in Russia in 1993 as an extra safety barrier is described for advanced WWER 1000 reactor steamline. The application of LBB concept requires the following additional protections. First, the steamline should be a highly qualified piping, performed in accordance with the applicable regulations and guidelines, carefully screened to verify that it is not subjected to any disqualifying failure mechanism. Second, a deterministic fracture mechanics analysis and leak rate evaluation have been performed to demonstrate that postulated through-wall crack that yields 95 1/min at normal operation conditions is stable even under seismic loads. Finally, it has been verified that the leak detection systems are sufficiently reliable, diverse and sensitive, and that adequate margins exist to detect a through wall crack smaller than the critical size. The obtained results are encouraging and show the possibility of the application of the LBB case to the steamline of advanced WWER 1000 reactor.

Application of the LBB regulatory approach to the steamlines of advanced WWER 1000 reactor

International Nuclear Information System (INIS)

Kiselyov, V.A.; Sokov, L.M.

1997-01-01

The LBB regulatory approach adopted in Russia in 1993 as an extra safety barrier is described for advanced WWER 1000 reactor steamline. The application of LBB concept requires the following additional protections. First, the steamline should be a highly qualified piping, performed in accordance with the applicable regulations and guidelines, carefully screened to verify that it is not subjected to any disqualifying failure mechanism. Second, a deterministic fracture mechanics analysis and leak rate evaluation have been performed to demonstrate that postulated through-wall crack that yields 95 1/min at normal operation conditions is stable even under seismic loads. Finally, it has been verified that the leak detection systems are sufficiently reliable, diverse and sensitive, and that adequate margins exist to detect a through wall crack smaller than the critical size. The obtained results are encouraging and show the possibility of the application of the LBB case to the steamline of advanced WWER 1000 reactor

Advanced nuclear reactor safety design technology research in NPIC

International Nuclear Information System (INIS)

Yu, H.

2014-01-01

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

Advanced passivation techniques for Si solar cells with high-κ dielectric materials

International Nuclear Information System (INIS)

Geng, Huijuan; Lin, Tingjui; Letha, Ayra Jagadhamma; Hwang, Huey-Liang; Kyznetsov, Fedor A.; Smirnova, Tamara P.; Saraev, Andrey A.; Kaichev, Vasily V.

2014-01-01

Electronic recombination losses at the wafer surface significantly reduce the efficiency of Si solar cells. Surface passivation using a suitable thin dielectric layer can minimize the recombination losses. Herein, advanced passivation using simple materials (Al 2 O 3 , HfO 2 ) and their compounds H (Hf) A (Al) O deposited by atomic layer deposition (ALD) was investigated. The chemical composition of Hf and Al oxide films were determined by X-ray photoelectron spectroscopy (XPS). The XPS depth profiles exhibit continuous uniform dense layers. The ALD-Al 2 O 3 film has been found to provide negative fixed charge (−6.4 × 10 11  cm −2 ), whereas HfO 2 film provides positive fixed charge (3.2 × 10 12  cm −2 ). The effective lifetimes can be improved after oxygen gas annealing for 1 min. I-V characteristics of Si solar cells with high-κ dielectric materials as passivation layers indicate that the performance is significantly improved, and ALD-HfO 2 film would provide better passivation properties than that of the ALD-Al 2 O 3 film in this research work.

AP calculus AB & BC crash course

CERN Document Server

Rosebush, J

2012-01-01

AP Calculus AB & BC Crash Course - Gets You a Higher Advanced Placement Score in Less Time Crash Course is perfect for the time-crunched student, the last-minute studier, or anyone who wants a refresher on the subject. AP Calculus AB & BC Crash Course gives you: Targeted, Focused Review - Study Only What You Need to Know Crash Course is based on an in-depth analysis of the AP Calculus AB & BC course description outline and actual AP test questions. It covers only the information tested on the exams, so you can make the most of your valuable study time. Written by experienced math teachers, our

Modeling of Wi-Fi IEEE 802.11ac Offloading Performance For 1000x Capacity Expansion of LTE-Advanced

DEFF Research Database (Denmark)

Hu, Liang; Sanchez, Maria Laura Luque; Maternia, Michael

2013-01-01

This paper studies indoor Wi-Fi IEEE 802.11ac deployment as a capacity expansion solution of LTE-A (Long Term Evolution-Advanced) network to achieve 1000 times higher capacity. Besides increasing the traffic volume by a factor of x1000, we also increase the minimum target user data rate to 10Mbit...

Results of the AP600 advanced plant probabilistic risk assessment

International Nuclear Information System (INIS)

Bueter, T.; Sancaktar, S.; Freeland, J.

1997-01-01

The AP600 Probabilistic Risk Assessment (PRA) includes detailed models of the plant systems, including the containment and containment systems that would be used to mitigate the consequences of a severe accident. The AP600 PRA includes a level 1 analysis (core damage frequency), and a level 2 analysis (environmental consequences), an assessment of the plant vulnerability to accidents caused by fire or floods, and a seismic margins analysis. Numerous sensitivities are included in the AP600 PRA including one that assumes no credit for non-safety plant systems. The core damage frequency for the AP600 of 1.7E-07/year is small compared with other PRAs performed in the nuclear industry. The AP600 large release frequency of 1.8E-08/year is also small and shows the ability of the containment systems to prevent a large release should a severe accident occur. Analyses of potential consequences to the environment from a severe accident show that a release would be small, and that containment still provides significant protection 24 hours after an assumed accident. Sensitivity analyses show that plant risk (as measured by core damage frequency and large release frequency) is not sensitive to the reliability of operator actions. 6 refs., 1 fig., 1 tab

25 CFR 1000.53 - Can Tribes/Consortia that receive advance planning grants also apply for a negotiation grant?

Science.gov (United States)

2010-04-01

... 25 Indians 2 2010-04-01 2010-04-01 false Can Tribes/Consortia that receive advance planning grants also apply for a negotiation grant? 1000.53 Section 1000.53 Indians OFFICE OF THE ASSISTANT SECRETARY, INDIAN AFFAIRS, DEPARTMENT OF THE INTERIOR ANNUAL FUNDING AGREEMENTS UNDER THE TRIBAL SELF-GOVERNMENT ACT AMENDMENTS TO THE INDIAN...

Status of the Short-Pulse X-ray Project (SPX) at the Advanced Photon Source (APS)

International Nuclear Information System (INIS)

Nassiri, R.; Arnold, N.D.; Berenc, G.; Borland, M.; Bromberek, D.J.; Chae, Y.-C.; Decker, G.; Emery, L.; Fuerst, J.D.; Grelick, A.E.; Horan, D.; Lenkszus, F.; Lill, R.M.; Sajaev, V.; Smith, T.L.; Waldschmidt, G.J.; Wu, G.; Yang, B.X.; Zholents, A.; Byrd, J.M.; Doolittle, L.R.; Huang, G.; Cheng, G.; Ciovati, G.; Henry, J.; Kneisel, P.; Mammosser, J.D.; Rimmer, R.A.; Turlington, L.; Wang, H.

2011-01-01

The Advanced Photon Source Upgrade project (APS-U) at Argonne includes implementation of Zholents deflecting cavity scheme for production of short x-ray pulses. This is a joint project between Argonne National Laboratory, Thomas Jefferson National Laboratory, and Lawrence Berkeley National Laboratory. This paper describes performance characteristics of the proposed source and technical issues related to its realization. Ensuring stable APS storage ring operation requires reducing quality factors of these modes by many orders of magnitude. These challenges reduce to those of the design of a single-cell SC cavity that can achieve the desired operating deflecting fields while providing needed damping of all these modes. The project team is currently prototyping and testing several promising designs for single-cell cavities with the goal of deciding on a winning design in the near future. Here we describe the approach undertaken and report the preliminary results. The concept of using transverse superconducting rf deflecting cavities to produce high-repetition-rate picoseconds x-rays with the APS has been previously described. Briefly, two cavities are required: the first cavity to impose a chirp on the electron beam and a second cavity to cancel the effects on the electron beam of the first cavity. The cavities must have a deflecting mode frequency that is a harmonic h of the APS storage ring rf frequency, 352 MHz A workable choice of h=8 corresponds to a deflecting cavity frequency of 2815 MHz. R and D activities include design and prototyping of superconducting deflecting cavities and components, cryomodule, low-level rf, particle/optical beam diagnostics, and timing/synchronization.

SWR 1000: A Next-Generation Boiling Water Reactor Ready for Deployment

International Nuclear Information System (INIS)

Brettschuh, W.

2006-01-01

The latest developments in nuclear power generation technology mainly concern large-capacity plants in the 1550 -1600 MW range, or very small plants (100 - 350 MW). The SWR 1000 boiling water reactor (BWR), by contrast, offers all of the advantages of an advanced plant design, with excellent safety performance and competitive power generation costs, in the medium-capacity range (1000 - 1250 MW). The SWR 1000 design is particularly suitable for countries whose power systems do not include any large power plants. The economic efficiency of this medium-sized plant in comparison with large-capacity designs is achieved by deploying very simple passive safety equipment, simplified systems for plant operation, and a very simple plant configuration in which systems engineering is optimized and dependence on electrical and instrumentation and control (IandC) systems is reduced. In addition, systems and components that require protection against natural and external man-made hazards are accommodated in such a way that as few buildings as possible have to be designed to withstand the loads from such events. The fuel assemblies to be deployed in the SWR 1000 core, meanwhile, have been enlarged from a 10x10 rod array to a 12x12 array. This reduces the total number of fuel assemblies in the core and thus also the number of control rods and control rod drives, as well as in-core neutron flux monitors. The design owes its competitiveness to the fact that investment costs, maintenance costs and fuel cycle costs are all lower. In addition, refueling outages are shorter, thanks to the reduced scope of outage activities. The larger fuel assemblies have been extensively and successfully tested, as have all of the other new components and systems incorporated into the plant design. As in existing plants, the forced coolant circulation method is deployed, ensuring problem-free start-up, and enabling plant operators to adjust power rapidly in the high power range (70%-100%) without moving

Operation of the APS rf gun

International Nuclear Information System (INIS)

Lewellen, J. W.

1998-01-01

The Advanced Photon Source (APS) has a thermionic-cathode rf gun system capable of providing beam to the APS linac. The gun system consists of a 1.6-cell thermionic-cathode rf gun, a fast kicker for beam current control, and an alpha magnet for bunch compression and injection into the APS linac line. This system is intended for use both as an injector for positron creation, and as a first beam source for the Low-Energy Undulator Test Line (LEUTL) project [1]. The first measured performance characteristics of the gun are presented.

Feasibility to convert an advanced PWR from UO2 to a mixed U/ThO2 core – Part I: Parametric studies

International Nuclear Information System (INIS)

Maiorino, Jose R.; Stefani, Giovanni Laranjo; Moreira, João M.L.; Rossi, Pedro C.R.; Santos, Thiago A.

2017-01-01

Highlights: • Neutronics calculation using SERPENT code. • Conversion of an advanced PWR from a UO 2 to (U-Th)O 2 core. • AP 1000-advanced PWR. • Parametric studies to define a converted core. • Demonstration of the feasibility to convert the AP 1000 by using mixed uranium thorium oxide fuel with advantages. - Abstract: This work presents the neutronics and thermal hydraulics feasibility to convert the UO 2 core of the Westinghouse AP1000 in a (U-Th)O 2 core by performing a parametric study varying the type of geometry of the pins in fuel elements, using the heterogeneous seed blanket concept and the homogeneous concept. In the parametric study, all geometry and materials for the burnable poison were kept the same as the AP 1000, and the only variable was the fuel pin material, in which we use several mass proportion of uranium and thorium but keeping the enrichment in 235 U, as LEU (20 w/o). The neutronics calculations were made by SERPENT code, and to validate the thermal limits we used a homemade code. The optimization criteria were to maximize the 233 U, and conversion factor, and minimize the plutonium production. The results obtained showed that the homogeneous concept with three different mass proportion zones, the first containing (32% UO 2 -68%ThO 2 ); the second with (24% UO 2 -76% ThO 2 ), and the third with (20% UO 2 -80% ThO 2 ), using 235 U LEU (20 w/o), and corresponding with the 3 enrichment zones of the AP 1000 (4.45 w/o; 3.40 w/o; 2.35 w/o), satisfies the optimization criteria as well as attending all thermal constrain. The concept showed advantages compared with the original UO 2 core, such a lower power density, and keeping the same 18 months of cycle a reduction of B-10 concentration at the soluble poison as well as eliminating in the integral boron poison coated (IFBA).

Management and integration of engineering and construction activities: Lessons learned from the AP1000R nuclear power plant China project

International Nuclear Information System (INIS)

McCullough, M. C.; Ebeling-Koning, D.; Evans, M. C.

2012-01-01

The lessons learned during the early phase of design engineering and construction activities for the AP1000 China Project can be applied to any project involving multiple disciplines and multiple organizations. Implementation of a first-of-a-kind design to directly support construction activities utilizing resources assigned to design development and design delivery creates challenges with prioritization of activities, successful closure of issues, and communication between site organizations and the home office. To ensure successful implementation, teams were assigned and developed to directly support construction activities including prioritization of activities, site communication and ensuring closure of site emergent issues. By developing these teams, the organization is better suited to meet the demands of the construction schedule while continuing with design evolution of a standard plant and engineering delivery for multiple projects. For a successful project, proper resource utilization and prioritization are key for overcoming obstacles and ensuring success of the engineering organization. (authors)

Market analysis of APS/SCM applications and technologies

OpenAIRE

Prášil, Zdeněk

2009-01-01

The bachelor thesis aims to describe a current worldwide market of APS/SCM technologies and applications , i.e. advanced planned scheduling /supply chain management. In the first part, the APS/SCM is described in theory. The APS/SCM is defined and its benefits and impacts on company are discussed. The next part of the work is focused on the market with APS/SCM and the distribution of forces in the market. The demand and supply of this market segment is analyzed. In the last part, solutions of...

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-06-28

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

Feasibility to convert an advanced PWR from UO2 to a mixed (U,Th)O2 core

International Nuclear Information System (INIS)

Stefani, Giovanni Laranjo de; Maiorino, José Rubens; Moreira, João Manoel de Losada; Santos, Thiago Augusto dos; Rossi, Pedro Carlos Russo

2017-01-01

This work presents the neutronics and thermal hydraulics feasibility to convert the UO2 core of the Westinghouse AP1000 in a (U-Th)O 2 core, rather than the traditional uranium dioxide, for the purpose of reducing long-lived actinides, especially plutonium, and generates a stock pile of 233 U, which could in the future be used in advanced fuel cycles, in a more sustainable process and taking advantage of the large stock of thorium available on the planet and especially in Brazil. The reactor chosen as reference was the AP1000, which is considered to be one of the most reliable and modern reactor of the current Generation III, and its similarity to the reactors already consolidated and used in Brazil for electric power generation. The results show the feasibility and potentiality of the concept, without the necessity of changes in the core of the AP1000, and even with advantages over this. The neutron calculations were made by the SERPENT code. The results provided a maximum linear power density lower than the AP1000, favoring safety. In addition, the delayed neutron fraction and the reactivity coefficients proved to be adequate to ensure the safety of the concept. The results show that a production of about 260 Kg of 233 U per cycle is possible, with a minimum production of fissile plutonium that favors the use of the concept in U-Th cycles. (author)

Defence in Depth by Design for the Advanced GIII NPP in China

Energy Technology Data Exchange (ETDEWEB)

Liu, S.; Zhang, Y.; Zhang, X., E-mail: liusongtao.npic@gmail.com [Science and Technology on Reactor System Design Technology Laboratory Chengdu, Sichuan (China)

2014-10-15

This paper describes the design of the advanced nuclear power plant ACP1000 in China that keeps the principle of defence in depth. To enhance the safety of the new generation NPPs, passive and active engineering safety features are used. The reactor will be kept safe under design basis accidents by using active engineering safety features, such as the medium and low pressure safety injection systems, and the emergency feedwater system. Under beyond DBAs, the passive safety systems will be actuated to keep removing residual heat for more than 72 hours, and to keep the core melt retained and cooled in the vessel. After the Fukushima nuclear accident, there are six main design enhancements in ACP1000 to meet the demands of the China authorities. (author)

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

International Nuclear Information System (INIS)

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

1996-01-01

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

National nuclear power planning of China and advanced reactor

International Nuclear Information System (INIS)

Qian Jihui

1990-01-01

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

SWR 1000: The new boiling water reactor power plant concept

International Nuclear Information System (INIS)

Brettschuh, W.

1999-01-01

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

Management and integration of engineering and construction activities: Lessons learned from the AP1000{sup R} nuclear power plant China project

Energy Technology Data Exchange (ETDEWEB)

McCullough, M. C.; Ebeling-Koning, D.; Evans, M. C. [Westinghouse Electric Company LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

2012-07-01

The lessons learned during the early phase of design engineering and construction activities for the AP1000 China Project can be applied to any project involving multiple disciplines and multiple organizations. Implementation of a first-of-a-kind design to directly support construction activities utilizing resources assigned to design development and design delivery creates challenges with prioritization of activities, successful closure of issues, and communication between site organizations and the home office. To ensure successful implementation, teams were assigned and developed to directly support construction activities including prioritization of activities, site communication and ensuring closure of site emergent issues. By developing these teams, the organization is better suited to meet the demands of the construction schedule while continuing with design evolution of a standard plant and engineering delivery for multiple projects. For a successful project, proper resource utilization and prioritization are key for overcoming obstacles and ensuring success of the engineering organization. (authors)

The Role of AP and the Composition Program.

Science.gov (United States)

Mahala, Daniel; Vivion, Michael

1993-01-01

Suggests that most programs have not based their acceptance of advanced placement credit on reasoned endorsement of the views of language, literature, and rhetoric that AP exams present. Criticizes the views implicit in the AP program and shows how they conflict with the goals of one particular college composition program. (RS)

Feasibility to convert an advanced PWR from UO{sub 2} to a mixed (U,Th)O{sub 2} core

Energy Technology Data Exchange (ETDEWEB)

Stefani, Giovanni Laranjo de; Maiorino, José Rubens; Moreira, João Manoel de Losada; Santos, Thiago Augusto dos, E-mail: giovanni_laranjo@yahoo.com.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil); Rossi, Pedro Carlos Russo [Department of Energy, System, Territory, and Construction Engineering (DESTEC), Pisa (Italy)

2017-07-01

This work presents the neutronics and thermal hydraulics feasibility to convert the UO2 core of the Westinghouse AP1000 in a (U-Th)O{sub 2} core, rather than the traditional uranium dioxide, for the purpose of reducing long-lived actinides, especially plutonium, and generates a stock pile of {sup 233}U, which could in the future be used in advanced fuel cycles, in a more sustainable process and taking advantage of the large stock of thorium available on the planet and especially in Brazil. The reactor chosen as reference was the AP1000, which is considered to be one of the most reliable and modern reactor of the current Generation III, and its similarity to the reactors already consolidated and used in Brazil for electric power generation. The results show the feasibility and potentiality of the concept, without the necessity of changes in the core of the AP1000, and even with advantages over this. The neutron calculations were made by the SERPENT code. The results provided a maximum linear power density lower than the AP1000, favoring safety. In addition, the delayed neutron fraction and the reactivity coefficients proved to be adequate to ensure the safety of the concept. The results show that a production of about 260 Kg of {sup 233}U per cycle is possible, with a minimum production of fissile plutonium that favors the use of the concept in U-Th cycles. (author)

Geochemistry and genesis of APS minerals in advanced argillic alteration zone, northwest of Shir-Kuh, Yazd

Directory of Open Access Journals (Sweden)

Batool Taghipour

2010-11-01

Full Text Available Northwest of Shir-Kuh batholith, a number of leucocratic granitic and granodioritic plutons have intruded sedimentary hosts including shale-sandstone (Triassic-Jurassic and sandstone-conglomerate (Lower Cretaceous. Contact metamorphism and hydrothermal alterations are widespread. Late alteration assemblage mainly occurs in arkosic sandstones of Sangestan Formation and includes propylitic, quartz-sericitic, advanced argillic and silicific zones. Quartz-sericite zone is the most widespread. Advanced argillic alteration is characterized by the following assemblage: jarosite, alunite, turquoise, from the Al-Phosphate-Sulfate group (APS. Considering this mineral assemblage and probable interactions taking place between the minerals, a geochemical environment with high fO2 and low pH is thought to be prevailing at the time of alteration and formation of alunite, jarosite and turquoise.

Passive cooling in modern nuclear reactors

International Nuclear Information System (INIS)

Rouai, N. M.

1998-01-01

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

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.)

New reactor programs from passive to pebble bed

Energy Technology Data Exchange (ETDEWEB)

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

2002-07-01

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

New reactor programs from passive to pebble bed

International Nuclear Information System (INIS)

Bruschi, H.J.

2002-01-01

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

Teaching Materials and Strategies for the AP Music Theory Exam

Science.gov (United States)

Lively, Michael T.

2017-01-01

Each year, many students take the Advanced Placement (AP) Music Theory Exam, and the majority of these students enroll in specialized AP music theory classes as part of the preparation process. For the teachers of these AP music theory classes, a number of challenges are presented by the difficulty and complexity of the exam subject material as…

Approaches to passive safety in advanced thermal reactors

International Nuclear Information System (INIS)

Moses, D.L.

1986-01-01

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

Analysis of AP1000{sup TM} reactor vessel cavity and support cooling

Energy Technology Data Exchange (ETDEWEB)

Craig, K.J. [Westinghouse Electric South Africa, 32 Park Avenue North, Highway Business Park, Centurion, 0157 (SOUTH AFRICA); Harkness, A.W. [Nuclear Power Plants, Westinghouse Electric Company, LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States); Kritzinger, H.P.; Hoffmann, J.E. [Pebble Bed Modular Reactor (Pty) Ltd, 1279 Mike Crawford Avenue, Centurion (South Africa)

2010-07-01

The paper investigates a Computational Fluid Dynamic (CFD) analysis of the air cooling of the Reactor Vessel (RV) cavity and RV supports. All the Heating, Ventilation and Air Conditioning (HVAC) flow of the RV cavity has to pass through the four RV supports supporting the four cold legs (cold inlets from the two steam generators) of the AP1000{sup TM} reactor. The RV support has a complex flow path leading to significant pressure drops to provide the necessary cooling. The insulation surrounding the RV has a specification on the amount of heat that may be transferred (lost) from the RV in order to maximize the heat transfer to the coolant driving the steam generators. This heat loss is applied as a boundary condition to the solution domain. Another heat source that is considered is that due to nuclear heating. Due to the fact that the heat source is nuclear in nature, gamma and neutron heating have to be considered for the surrounding structures. These include the carbon steel structural module that encapsulates the RV cavity, as well as the concrete poured around this module. The space in the gap between the RV insulation and the structural module steel shell is not only obstructed by the insulation supports, but also by wells or tubes within which power and intermediate ex-core detectors are located. Source-range ex-core detectors are embedded in the concrete surrounding the structural module. All these detectors have a limited operating temperature range, and together with limits on concrete temperatures for safety considerations, necessitate the need for CFD simulations to determine the range of operational temperatures seen by these components. The CFD simulations also provide an estimate of the pressure drop through the cavity between the RV insulation and structural module, as well as that through the four RV supports. Results presented include ANSYS{sup R} FLUENT{sup R} simulations describing the modelling procedure that was followed, namely to combine

Advancing the Use of Passive Sampling in Risk Assessment and Management of Sediments Contaminated with Hydrophobic Organic Chemicals: Results of an International Ex Situ Passive Sampling Interlaboratory Comparison.

Science.gov (United States)

Jonker, Michiel T O; van der Heijden, Stephan A; Adelman, Dave; Apell, Jennifer N; Burgess, Robert M; Choi, Yongju; Fernandez, Loretta A; Flavetta, Geanna M; Ghosh, Upal; Gschwend, Philip M; Hale, Sarah E; Jalalizadeh, Mehregan; Khairy, Mohammed; Lampi, Mark A; Lao, Wenjian; Lohmann, Rainer; Lydy, Michael J; Maruya, Keith A; Nutile, Samuel A; Oen, Amy M P; Rakowska, Magdalena I; Reible, Danny; Rusina, Tatsiana P; Smedes, Foppe; Wu, Yanwen

2018-03-20

This work presents the results of an international interlaboratory comparison on ex situ passive sampling in sediments. The main objectives were to map the state of the science in passively sampling sediments, identify sources of variability, provide recommendations and practical guidance for standardized passive sampling, and advance the use of passive sampling in regulatory decision making by increasing confidence in the use of the technique. The study was performed by a consortium of 11 laboratories and included experiments with 14 passive sampling formats on 3 sediments for 25 target chemicals (PAHs and PCBs). The resulting overall interlaboratory variability was large (a factor of ∼10), but standardization of methods halved this variability. The remaining variability was primarily due to factors not related to passive sampling itself, i.e., sediment heterogeneity and analytical chemistry. Excluding the latter source of variability, by performing all analyses in one laboratory, showed that passive sampling results can have a high precision and a very low intermethod variability (sampling, irrespective of the specific method used, is fit for implementation in risk assessment and management of contaminated sediments, provided that method setup and performance, as well as chemical analyses are quality-controlled.

Analysis of the AP600 core makeup tank experiments using the NOTRUMP code

International Nuclear Information System (INIS)

Cunningham, J.C.; Haberstroh, R.C.; Hochreiter, L.E.; Jaroszewicz, J.

1995-01-01

The AP600 design utilizes passive methods to perform core and containment cooling functions for a postulated loss of coolant. The core makeup tank (CMT) is an important feature of the AP600 passive safety system. The NOTRUMP code has been compared to the 300-series core makeup tank experiments. It has been observed that the code will capture the correct thermal-hydraulic behavior observed in the experiments. The correlations used for wall film condensation and convective heat transfer to the heated CMT liquid appear to be appropriate for these applications. The code will predict the rapid condensation and mixing thermal-hydraulic behavior observed in the 300-series tests. The NOTRUMP predictions can be noding-dependent since the condensation is extremely dependent on the amount of cold CMT liquid that mixes with the incoming steam flow

ONLINE MINIMIZATION OF VERTICAL BEAM SIZES AT APS

Energy Technology Data Exchange (ETDEWEB)

Sun, Yipeng

2017-06-25

In this paper, online minimization of vertical beam sizes along the APS (Advanced Photon Source) storage ring is presented. A genetic algorithm (GA) was developed and employed for the online optimization in the APS storage ring. A total of 59 families of skew quadrupole magnets were employed as knobs to adjust the coupling and the vertical dispersion in the APS storage ring. Starting from initially zero current skew quadrupoles, small vertical beam sizes along the APS storage ring were achieved in a short optimization time of one hour. The optimization results from this method are briefly compared with the one from LOCO (Linear Optics from Closed Orbits) response matrix correction.

The APS SASE FEL: modeling and code comparison

International Nuclear Information System (INIS)

Biedron, S. G.

1999-01-01

A self-amplified spontaneous emission (SASE) free-electron laser (FEL) is under construction at the Advanced Photon Source (APS). Five FEL simulation codes were used in the design phase: GENESIS, GINGER, MEDUSA, RON, and TDA3D. Initial comparisons between each of these independent formulations show good agreement for the parameters of the APS SASE FEL

Analysis on working pressure selection of ACME integral test facility

International Nuclear Information System (INIS)

Chen Lian; Chang Huajian; Li Yuquan; Ye Zishen; Qin Benke

2011-01-01

An integral effects test facility, advanced core cooling mechanism experiment facility (ACME) was designed to verify the performance of the passive safety system and validate its safety analysis codes of a pressurized water reactor power plant. Three test facilities for AP1000 design were introduced and review was given. The problems resulted from the different working pressures of its test facilities were analyzed. Then a detailed description was presented on the working pressure selection of ACME facility as well as its characteristics. And the approach of establishing desired testing initial condition was discussed. The selected 9.3 MPa working pressure covered almost all important passive safety system enables the ACME to simulate the LOCAs with the same pressure and property similitude as the prototype. It's expected that the ACME design would be an advanced core cooling integral test facility design. (authors)

Summary test results of the particle-beam diagnostics for the Advanced Photon Source (APS) subsystems

International Nuclear Information System (INIS)

Lumpkin, A.; Wang, X.; Sellyey, W.; Patterson, D.; Kahana, E.

1994-01-01

During the first half of 1994, a number of the diagnostic systems for measurement of the charged-particle beam parameters throughout the subsystems of the Advanced Photon Source (APS) have been installed and tested. The particle beams eventually will involve 450-MeV to 7-GeV positrons and with different pulse formats. The first test and commissionin results for beam profiles, beam position monitors, loss rate monitors, current monitors, and synchrotron radiation photon monitors hve been obtained using 200- to 350-MeV electron beams injected into the subsystems. Data presented are principally from the transport lines and the positron accumulator ring

SWR-1000 concept on control of severe accidents

International Nuclear Information System (INIS)

Meyer, P.J.

1998-01-01

It is essential for the SWR-1000 probabilistic safety concept to consider the results from experiments and reliability system failure within the probabilistic safety analyses for passive systems. Active and passive safety features together reduce the probability of the occurrence of beyond design basis accidents in order to limit their consequences in accordance with the German law. As a reference case we analyzed the most probable core melt accident sequence with a very conservative assumption. An initial event, stuck open of safety and relief valves without the probability of active and passive feeding systems of the pressure vessel, was considered. Other sequences of the loss of coolant accidents lead to lower probability

SPES-2, AP600 intergral system test S01007 2 inch CL to core make-up tank pressure balance line break

Energy Technology Data Exchange (ETDEWEB)

Bacchiani, M.; Medich, C.; Rigamonti, M. [SIET S.p.A. Piacenza (Italy)] [and others

1995-09-01

The SPES-2 is a full height, full pressure experimental test facility reproducing the Westinghouse AP600 reactor with a scaling factor of 1/395. The experimental plant, designed and operated by SIET in Piacenza, consists of a full simulation of the AP600 primary core cooling system including all the passive and active safety systems. In 1992, Westinghouse, in cooperation with ENEL (Ente Nazionale per l` Energia Elettrica), ENEA (Enter per le numove Technlogie, l` Energia e l` Ambient), Siet (Societa Informazioni Esperienze Termoidraulich) and ANSALDO developed an experimental program to test the integrated behaviour of the AP600 passive safety systems. The SPES-2 test matrix, concluded in November 1994, has examined the AP600 passive safety system response for a range of small break LOCAs at different locations on the primary system and on the passive system lines; single steam generator tube ruptures with passive and active safety systems and a main steam line break transient to demonstrate the boration capability of passive safety systems for rapid cooldown. Each of the tests has provided detailed experimental results for verification of the capability of the analysis methods to predict the integrated passive safety system behaviour. Cold and hot shakedown tests have been performed on the facility to check the characteristics of the plant before starting the experimental campaign. The paper first presents a description of the SPES-2 test facility then the main results of S01007 test {open_quotes}2{close_quotes} Cold Leg (CL) to Core Make-up Tank (CMT) pressure balance line break{close_quotes} are reported and compared with predictions performed using RELAP5/mod3/80 obtained by ANSALDO through agreement with U.S.N.R.C. (U.S. Nuclear Regulatory Commission). The SPES-2 nodalization and all the calculations here presented were performed by ANSALDO and sponsored by ENEL as a part of pre-test predictions for SPES-2.

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Oncology Advanced Practitioners Bring Advanced Community Oncology Care.

Science.gov (United States)

Vogel, Wendy H

2016-01-01

Oncology care is becoming increasingly complex. The interprofessional team concept of care is necessary to meet projected oncology professional shortages, as well as to provide superior oncology care. The oncology advanced practitioner (AP) is a licensed health care professional who has completed advanced training in nursing or pharmacy or has completed training as a physician assistant. Oncology APs increase practice productivity and efficiency. Proven to be cost effective, APs may perform varied roles in an oncology practice. Integrating an AP into an oncology practice requires forethought given to the type of collaborative model desired, role expectations, scheduling, training, and mentoring.

Genome-Wide Identification of the Target Genes of AP2-O, a Plasmodium AP2-Family Transcription Factor.

Directory of Open Access Journals (Sweden)

Izumi Kaneko

2015-05-01

Full Text Available Stage-specific transcription is a fundamental biological process in the life cycle of the Plasmodium parasite. Proteins containing the AP2 DNA-binding domain are responsible for stage-specific transcriptional regulation and belong to the only known family of transcription factors in Plasmodium parasites. Comprehensive identification of their target genes will advance our understanding of the molecular basis of stage-specific transcriptional regulation and stage-specific parasite development. AP2-O is an AP2 family transcription factor that is expressed in the mosquito midgut-invading stage, called the ookinete, and is essential for normal morphogenesis of this stage. In this study, we identified the genome-wide target genes of AP2-O by chromatin immunoprecipitation-sequencing and elucidate how this AP2 family transcription factor contributes to the formation of this motile stage. The analysis revealed that AP2-O binds specifically to the upstream genomic regions of more than 500 genes, suggesting that approximately 10% of the parasite genome is directly regulated by AP2-O. These genes are involved in distinct biological processes such as morphogenesis, locomotion, midgut penetration, protection against mosquito immunity and preparation for subsequent oocyst development. This direct and global regulation by AP2-O provides a model for gene regulation in Plasmodium parasites and may explain how these parasites manage to control their complex life cycle using a small number of sequence-specific AP2 transcription factors.

A Closer Examination of the Academic Benefits of AP

Science.gov (United States)

McKillip, Mary E. M.; Rawls, Anita

2013-01-01

The authors sought to better understand the relationship between students participating in the Advanced Placement (AP) program and subsequent performance on the Scholastic Aptitude Test (SAT). Focusing on students graduating from U.S. public high schools in 2010, the authors used propensity scores to match junior year AP examinees in 3 subjects to…

Advanced fuel cycles for WWER-1000 reactors

International Nuclear Information System (INIS)

Semchenkov, Y. M.; Pavlovichev, A. M.; Pavlov, V. I.; Spirkin, E. I.; Styrin, Y. A.; Kosourov, E. K.

2007-01-01

Main stages of Russian uranium fuel development regarding improvement of safety and economics of fuel load operation are presented. Intervals of possible changes in fuel cycle duration have been demonstrated for the use of current and perspective fuel. Examples of equilibrium fuel load patterns have been demonstrated and main core neutronics parameters have been presented. Problems on the use of axial blankets with reduced enrichment in WWER-1000 fuel assemblies are considered. Some results are presented regarding core neutronic characteristics of WWER-1000 at the use of regenerated uranium and uranium-plutonium fuel. Examples of equilibrium fuel cycles for the core partially loaded with MOX fuel from weapon-grade plutonium are also considered (Authors)

Advanced Performance Modeling with Combined Passive and Active Monitoring

Energy Technology Data Exchange (ETDEWEB)

Dovrolis, Constantine [Georgia Inst. of Technology, Atlanta, GA (United States); Sim, Alex [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)

2015-04-15

To improve the efficiency of resource utilization and scheduling of scientific data transfers on high-speed networks, the "Advanced Performance Modeling with combined passive and active monitoring" (APM) project investigates and models a general-purpose, reusable and expandable network performance estimation framework. The predictive estimation model and the framework will be helpful in optimizing the performance and utilization of networks as well as sharing resources with predictable performance for scientific collaborations, especially in data intensive applications. Our prediction model utilizes historical network performance information from various network activity logs as well as live streaming measurements from network peering devices. Historical network performance information is used without putting extra load on the resources by active measurement collection. Performance measurements collected by active probing is used judiciously for improving the accuracy of predictions.

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

International Nuclear Information System (INIS)

Karameldin, A.; Samer S. Mekhemar

2004-01-01

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

A hot-spare injector for the APS linac

International Nuclear Information System (INIS)

Lewellen, J. W.

1999-01-01

Last year a second-generation SSRL-type thermionic cathode rf gun was installed in the Advanced Photon Source (APS) linac. This gun (referred to as ''gun2'') has been successfully commissioned and now serves as the main injector for the APS linac, essentially replacing the Koontz-type DC gun. To help ensure injector availability, particularly with the advent of top-up mode operation at the APS, a second thermionic-cathode rf gun will be installed in the APS linac to act as a hot-spare beam source. The hot-spare installation includes several unique design features, including a deep-orbit Panofsky-style alpha magnet. Details of the hot-spare beamline design and projected performance are presented, along with some plans for future performance upgrades

Materials for passively safe reactors

International Nuclear Information System (INIS)

Simnad, T.

1993-01-01

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

Task Force on Catastrophic Antiphospholipid Syndrome (APS) and Non-criteria APS Manifestations (I): catastrophic APS, APS nephropathy and heart valve lesions.

Science.gov (United States)

Cervera, R; Tektonidou, M G; Espinosa, G; Cabral, A R; González, E B; Erkan, D; Vadya, S; Adrogué, H E; Solomon, M; Zandman-Goddard, G; Shoenfeld, Y

2011-02-01

The objectives of the 'Task Force on Catastrophic Antiphospholipid Syndrome (APS) and Non-criteria APS Manifestations' were to assess the clinical utility of the international consensus statement on classification criteria and treatment guidelines for the catastrophic APS, to identify and grade the studies that analyse the relationship between the antiphospholipid antibodies and the non-criteria APS manifestations and to present the current evidence regarding the accuracy of these non-criteria APS manifestations for the detection of patients with APS. This article summarizes the studies analysed on the catastrophic APS, APS nephropathy and heart valve lesions, and presents the recommendations elaborated by the Task Force after this analysis.

SWR 1000: Efficient design for operational excellence

International Nuclear Information System (INIS)

Brettschuh, W.

2008-01-01

The SWR 1000 boiling water reactor (BWR) offers all of the advantages associated with an advanced plant design, i.e. excellent safety performance and competitive power generation costs, in the medium capacity range (1000-1250 MW). The economic efficiency of this medium-sized plant in comparison with large-capacity designs is achieved by using very simple passive safety equipment, simplified plant operating systems, and a very simple plant configuration. Systems engineering is thus optimized, reducing dependence on electrical and instrumentation and control (I and C) systems. The fuel assemblies deployed in the SWR 1000 core are enlarged from a 10 x 10 to a 12 x 12 rod array. This cuts down the total number of fuel assemblies in the core and hence also the number of control rods and control rod drives, as well as in-core neutron flux monitors. The design owes its competitiveness to the fact that investment, maintenance and fuel cycle costs are all lower. In addition, refueling outages are shorter, thanks to the reduced scope for outage activities. For example, there are no bolted reactor internal joints, and the SWR 1000 has a permanently installed reactor vessel-to-drywell seal. Replacement of in-core detectors is carried out from below, and does not affect the critical path of the outage. Furthermore all in-service -inspections (ISIs) on the reactor pressure vessel (RPV) and its nozzles can be undertaken from the outside, so that no extended ISI outages are necessary. As in existing plants, forced coolant circulation is utilized, ensuring problem-free startup and enabling plant operators to adjust power rapidly in the high power range (70%-100%). This is achieved without moving the control rods, and allows both spectral-shift and stretch-out operation. The considerable gains provided by forced coolant circulation, in terms of operational flexibility and fuel utilization, mean that the investment and maintenance costs of the pumps are covered within just a few

A new gap separation mechanism for APS insertion devices

International Nuclear Information System (INIS)

Trakhtenberg, E. M.; Tcheskidov, V.; Den Hartog, P. K.; Deriy, B.; Erdmann, M.; Makarov, O.; Moog, E. R.

1999-01-01

A new gap separation mechanism for use with the standard Advanced Photon Source (APS) 3.3-cm-period undulator magnetic structures has been designed and built and the first system has been installed in the APS storage ring. The system allows a minimum magnetic gap of 10 mm for use with the APS 8-mm insertion device vacuum chambers. The mechanism is a bolted steel frame structure with a simple 4-motor mechanical drive train. The control system uses servomotors with incremental rotary encoders and virtual absolute linear encoders

AP600 design certification thermal hydraulics testing and analysis

Energy Technology Data Exchange (ETDEWEB)

Hochreiter, L.E.; Piplica, E.J.

1995-09-01

Westinghouse Electric Corporation, in conjunction with the Department of Energy and the Electric Power Research Institute, have been developing an advanced light water reactor design; the AP600. The AP600 is a 1940 Mwt, 600Mwe unit which is similar to a Westinghouse two-loop Pressurized Water Reactor. The accumulated knowledge on reactor design to reduce the capital costs, construction time, and the operational and maintenance cost of the unit once it begins to generate electrical power. The AP600 design goal is to maintain an overall cost advantage over fossil generated electrical power.

Reactor coolant pump type RUV for Westinghouse Electric Company LLC reactor AP1000 TM

International Nuclear Information System (INIS)

Baumgarten, S.; Brecht, B.; Bruhns, U.; Fehring, P.

2010-01-01

The RUV is a reactor coolant pump, specially designed for the Westinghouse Electric Company LLC AP1000 TM reactor. It is a hermetically sealed, wet winding motor pump. The RUV is a very compact, vertical pump/motor unit, designed to fit into the compartment next to the reactor pressure vessel. Each of the two steam generators has two pump casings welded to the channel head by the suction nozzle. The pump/motor unit consists of a pump part, where a semi-axial impeller/diffuser combination is mounted in a one-piece pump casing. Computational Fluid Dynamics methods combined with various hydraulic tests in a 1:2 scale hydraulic test assure full compliance with the specific customer requirements. A short and rigid shaft, supported by a radial bearing, connects the impeller with the high inertia flywheel. This flywheel consists of a one-piece forged stainless steel cylinder, with an option for several smaller heavy metal cylinders inside. The flywheel is located inside the thermal barrier, which forms part of the pressure boundary. A specific arrangement of cooling water circuits guarantees a homogeneous temperature distribution in and around the flywheel, minimizes the friction losses of the flywheel and protects the motor from hot coolant. The driving torque is transmitted by the motor shaft, which itself is supported by two radial bearings. A three-phase, high-voltage squirrel-cage induction motor generates the driving torque. Due to the wet winding concept it is possible to achieve positive effects regarding motor lifetime. The cooling water is forced through the stator windings and the gap between rotor and stator by an auxiliary impeller. Furthermore, this wet winding motor concept has higher efficiency as compared to a canned motor since there are no eddy current losses. As part of the design process and in addition to the hydraulic scale model, a complete half scale model pump was built. It was used to verify the calculations performed like coast

25 CFR 1000.54 - How will a Tribe/Consortium know whether or not it has been selected to receive an advance...

Science.gov (United States)

2010-04-01

...) Planning and Negotiation Grants Advance Planning Grant Funding § 1000.54 How will a Tribe/Consortium know... Director will notify the Tribe/Consortium by letter whether it has been selected to receive an advance... 25 Indians 2 2010-04-01 2010-04-01 false How will a Tribe/Consortium know whether or not it has...

State-of-the-Art developments in accelerator controls at the APS

International Nuclear Information System (INIS)

Lenkszus, F.

1999-01-01

The performance requirements of the Advanced Photon Source (APS) challenge the control system in a number of areas. This paper will review a few applications of advanced technology in the control and monitoring of the APS. The application of digital signal processors (DSPs) and techniques will be discussed, both from the perspective of a large distributed multiprocessor system and from that of embedded systems. In particular, two embedded applications will be highlighted, a beam position monitor processor and a DSP-based power supply controller. Fast data distribution is often a requirement. The application of a high-speed network based on reflective memory will also be discussed in the context of the APS global orbit feedback system. Timing systems provide opportunities to apply technologies such as high-speed logic and fiber optics. Examples of the use of these technologies will also be included. Finally, every modern accelerator control system of any size requires networking. Features of the APS accelerator controls network will be discussed

AP Geography, Environmental Science Thrive

Science.gov (United States)

Robelen, Erik W.

2012-01-01

Geography may not be particularly known as a hot topic among today's students--even some advocates suggest it suffers from an image problem--but by at least one measure, the subject is starting to come into its own. Across more than 30 topics covered in the Advanced Placement (AP) program, participation in geography is rising faster than any…

The APS control system network upgrade

International Nuclear Information System (INIS)

Sidorowicz, K. v.; Leibfritz, D.; McDowell, W. P.

1999-01-01

When it was installed,the Advanced Photon Source (APS) control system network was at the state-of-the-art. Different aspects of the system have been reported at previous meetings [1,2]. As loads on the controls network have increased due to newer and faster workstations and front-end computers, we have found performance of the system declining and have implemented an upgraded network. There have been dramatic advances in networking hardware in the last several years. The upgraded APS controls network replaces the original FDDI backbone and shared Ethernet hubs with redundant gigabit uplinks and fully switched 10/100 Ethernet switches with backplane fabrics in excess of 20 Gbits/s (Gbps). The central collapsed backbone FDDI concentrator has been replaced with a Gigabit Ethernet switch with greater than 30 Gbps backplane fabric. Full redundancy of the system has been maintained. This paper will discuss this upgrade and include performance data and performance comparisons with the original network

APS Science 2007.

Energy Technology Data Exchange (ETDEWEB)

2008-05-30

This report provides research highlights from the Advanced Photon Source (APS). Although these highlights represent less than 10% of the published work from the APS in 2007, they give a flavor of the diversity and impact of user research at the facility. In the strategic planning the aim is to foster the growth of existing user communities and foresee new areas of research. This coming year finds the APS engaged in putting together, along with the users, a blueprint for the next five years, and making the case for a set of prioritized investments in beamlines, the accelerator, and infrastructure, each of which will be transformational in terms of scientific impact. As this is written plans are being formulated for an important user workshop on October 20-21, 2008, to prioritize strategic plans. The fruit from past investments can be seen in this report. Examples include the creation of a dedicated beamline for x-ray photon correlation spectroscopy at Sector 8, the evolution of dedicated high-energy x-ray scattering beamlines at sectors 1 and 11, a dedicated imaging beamline at Sector 32, and new beamlines for inelastic scattering and powder diffraction. A single-pulse facility has been built in collaboration with Sector 14 (BioCARS) and Phil Anfinrud at the National Institutes of Health, which will offer exceptionally high flux for single-pulse diffraction. The nanoprobe at Sector 26, built and operated jointly by the Argonne Center for Nanoscale Materials and the X-ray Operations and Research (XOR) section of the APS X-ray Science Division, has come on line to define the state of the art in nanoscience.

APS Science 2007

International Nuclear Information System (INIS)

2008-01-01

This report provides research highlights from the Advanced Photon Source (APS). Although these highlights represent less than 10% of the published work from the APS in 2007, they give a flavor of the diversity and impact of user research at the facility. In the strategic planning the aim is to foster the growth of existing user communities and foresee new areas of research. This coming year finds the APS engaged in putting together, along with the users, a blueprint for the next five years, and making the case for a set of prioritized investments in beamlines, the accelerator, and infrastructure, each of which will be transformational in terms of scientific impact. As this is written plans are being formulated for an important user workshop on October 20-21, 2008, to prioritize strategic plans. The fruit from past investments can be seen in this report. Examples include the creation of a dedicated beamline for x-ray photon correlation spectroscopy at Sector 8, the evolution of dedicated high-energy x-ray scattering beamlines at sectors 1 and 11, a dedicated imaging beamline at Sector 32, and new beamlines for inelastic scattering and powder diffraction. A single-pulse facility has been built in collaboration with Sector 14 (BioCARS) and Phil Anfinrud at the National Institutes of Health, which will offer exceptionally high flux for single-pulse diffraction. The nanoprobe at Sector 26, built and operated jointly by the Argonne Center for Nanoscale Materials and the X-ray Operations and Research (XOR) section of the APS X-ray Science Division, has come on line to define the state of the art in nanoscience

The development of beam current monitors in the APS

International Nuclear Information System (INIS)

Wang, X.; Lenkszus, F.; Rotela, E.

1995-01-01

The Advanced Photon Source (APS) is a third-generation 7-GeV synchrotron radiation source. The precision measurement of beam current is a challenging task in high energy accelerators, such as the APS, with a wide range of beam parameters and complicated noise, radiation, and thermal environments. The beam pulses in the APS injector and storage ring have charge ranging from 50pC to 25nC with pulse durations varying from 30ps to 30ns. A total of nine non- intercepting beam current monitors have been installed in the APS facility (excluding those in the linac) for general current measurement. In addition, several independent current monitors with specially designed redundant interlock electronics are installed for personnel safety and machine protection. This paper documents the design and development of current monitors in the APS,. discusses the commissioning experience in the past year, and presents the results of recent operations

CGM ApS Årsberetning til DANAK

DEFF Research Database (Denmark)

De Chiffre, Leonardo

Denne årsberetning omfatter CGM ApS' akkrediterede virksomhed i kalenderåret 2003. Årsberetningen er udarbejdet til DANAK (Dansk Akkreditering, ErhvervsfremmeStyrelsen), som led i opfyldelsen af laboratoriets informationspligt i henhold til gældende regler.......Denne årsberetning omfatter CGM ApS' akkrediterede virksomhed i kalenderåret 2003. Årsberetningen er udarbejdet til DANAK (Dansk Akkreditering, ErhvervsfremmeStyrelsen), som led i opfyldelsen af laboratoriets informationspligt i henhold til gældende regler....

Understanding and Using the New Guided-Inquiry AP Chemistry Laboratory Manual

Science.gov (United States)

Cacciatore, Kristen L.

2014-01-01

To support teaching and learning in the advanced placement (AP) chemistry laboratory, the College Board published a laboratory manual, "AP Chemistry Guided-Inquiry Experiments: Applying the Science Practices," in 2013 as part of the redesigned course. This article provides a discussion of the rationale for the existence of the manual as…

Renal involvement in the antiphospholipid syndrome (APS)-APS nephropathy.

Science.gov (United States)

Tektonidou, Maria G

2009-06-01

Although the kidney represents a major target organ in antiphospholipid syndrome (APS), renal involvement in APS was poorly recognized until recently. The most well-recognized renal manifestations of APS are the renal artery thrombosis/stenosis, renal infarction, hypertension, renal vein thrombosis, end-stage renal disease, increased allograft vascular thrombosis, some types of glomerular disease, and a small-vessel vaso-occlusive nephropathy, recently defined as APS nephropathy. APS nephropathy was first described in primary APS patients, characterized by acute thrombotic lesions in glomeruli and/or arterioles (thrombotic microangiopathy) and chronic vascular lesions such as fibrous intimal hyperplasia of arterioles and interlobular arteries, organized thrombi with or without recanalization, and fibrous arterial and arteriolar occlusions or focal cortical atrophy. APS nephropathy was also detected in further studies including patients with systemic lupus erythematosus (SLE)-related APS and SLE/non-APS patients with positive antiphospholipid antibodies, independently of lupus nephritis. The same histologic lesions, especially thrombotic mictroangiopathy, were also observed in patients with catastrophic APS. The most frequent clinical and laboratory characteristics of APS nephropathy in all the above groups of patients are hypertension (often severe), proteinuria (ranging from mild to nephrotic range), hematuria, and acute or chronic renal insufficiency.

Advanced operation strategy for feed-and-bleed operation in an OPR1000

International Nuclear Information System (INIS)

Kim, Bo Gyung; Yoon, Ho Joon; Kim, Jaewhan; Kang, Hyun Gook

2016-01-01

study is expected to provide a systematic operation strategy to initiate F&B operation under various plant situations. An OPR1000 is used in this study as an example plant, with the resulting advanced operating strategy able to be applied to most PWRs which have F&B operation capability.

Using a Classroom Response System to Improve Multiple-Choice Performance in AP[R] Physics

Science.gov (United States)

Bertrand, Peggy

2009-01-01

Participation in rigorous high school courses such as Advanced Placement (AP[R]) Physics increases the likelihood of college success, especially for students who are traditionally underserved. Tackling difficult multiple-choice exams should be part of any AP program because well-constructed multiple-choice questions, such as those on AP exams and…

Development of a simulation platform for dynamic simulation and control studies of AP1000 nuclear steam supply system

International Nuclear Information System (INIS)

Wan, Jiashuang; Song, Hongbing; Yan, Shoujun; Sun, Jian; Zhao, Fuyu

2015-01-01

Highlights: • A fast-running simulation platform named NCAP was developed on a personal computer using MATLAB/Simulink. • Three types of typical operations, namely 10% step load change, 5%/min ramp load change and load follow were simulated. • NCAP predictions were compared with those obtained by CENTS for the load regulation transients. - Abstract: This paper presents the development, application and performance assessment of a fast-running NCAP (NSSS Control & Analysis Platform) in MATLAB/Simulink environment. First, a nodal core model, a lumped parameter dynamic steam generator model with moving boundary, a non-equilibrium two-regions-three-volumes pressurizer model, and the relevant pipe and plenum models were proposed based on the fundamental conservation of mass, energy and momentum. Then, these first order nonlinear models and the NSSS control systems were implemented in the Simulink by the predefined library blocks. Based on the developed NCAP, three types of typical operational transients, namely the 10% step load change, the 5%/min ramp load change and the daily load follow were simulated to study the dynamic behavior and control characteristics of the AP1000 NSSS. It has been demonstrated that the dynamic responses of the selected key parameters agree well with the general physical rules. In addition, the comparison of load regulation simulation results obtained by NCAP and CENTS shows a good agreement in terms of the changing trends. With the adoption of modular programming techniques, the NCAP facilitates easy modification and runs quickly, which easily allows the control system designer to test and compare various ideas efficiently

APS beamline standard components handbook

International Nuclear Information System (INIS)

Kuzay, T.M.

1992-01-01

It is clear that most Advanced Photon Source (APS) Collaborative Access Team (CAT) members would like to concentrate on designing specialized equipment related to their scientific programs rather than on routine or standard beamline components. Thus, an effort is in progress at the APS to identify standard and modular components of APS beamlines. Identifying standard components is a nontrivial task because these components should support diverse beamline objectives. To assist with this effort, the APS has obtained advice and help from a Beamline Standardization and Modularization Committee consisting of experts in beamline design, construction, and operation. The staff of the Experimental Facilities Division identified various components thought to be standard items for beamlines, regardless of the specific scientific objective of a particular beamline. A generic beamline layout formed the basis for this identification. This layout is based on a double-crystal monochromator as the first optical element, with the possibility of other elements to follow. Pre-engineering designs were then made of the identified standard components. The Beamline Standardization and Modularization Committee has reviewed these designs and provided very useful input regarding the specifications of these components. We realize that there will be other configurations that may require special or modified components. This Handbook in its current version (1.1) contains descriptions, specifications, and pre-engineering design drawings of these standard components. In the future, the APS plans to add engineering drawings of identified standard beamline components. Use of standard components should result in major cost reductions for CATs in the areas of beamline design and construction

MULTI-OBJECTIVE ONLINE OPTIMIZATION OF BEAM LIFETIME AT APS

Energy Technology Data Exchange (ETDEWEB)

Sun, Yipeng

2017-06-25

In this paper, online optimization of beam lifetime at the APS (Advanced Photon Source) storage ring is presented. A general genetic algorithm (GA) is developed and employed for some online optimizations in the APS storage ring. Sextupole magnets in 40 sectors of the APS storage ring are employed as variables for the online nonlinear beam dynamics optimization. The algorithm employs several optimization objectives and is designed to run with topup mode or beam current decay mode. Up to 50\\% improvement of beam lifetime is demonstrated, without affecting the transverse beam sizes and other relevant parameters. In some cases, the top-up injection efficiency is also improved.

Passive heat transport in advanced CANDU containment

International Nuclear Information System (INIS)

Krause, M.; Mathew, P.M.

1994-01-01

A passive CANDU containment design has been proposed to provide the necessary heat removal following a postulated accident to maintain containment integrity. To study its feasibility and to optimize the design, multi-dimensional containment modelling may be required. This paper presents a comparison of two CFD codes, GOTHIC and PHOENICS, for multi-dimensional containment analysis and gives pressure transient predictions from a lumped-parameter and a three-dimensional GOTHIC model for a modified CANDU-3 containment. GOTHIC proved suitable for multidimensional post-accident containment analysis, as shown by the good agreement with pressure transient predictions from PHOENICS. GOTHIC is, therefore, recommended for passive CANDU containment modelling. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-04-26

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

Simulation and Analysis of Small Break LOCA for AP1000 Using RELAP5-MV and Its Comparison with NOTRUMP Code

Directory of Open Access Journals (Sweden)

Eltayeb Yousif

2017-01-01

Full Text Available Many reactor safety simulation codes for nuclear power plants (NPPs have been developed. However, it is very important to evaluate these codes by testing different accident scenarios in actual plant conditions. In reactor analysis, small break loss of coolant accident (SBLOCA is an important safety issue. RELAP5-MV Visualized Modularization software is recognized as one of the best estimate transient simulation programs of light water reactors (LWR. RELAP5-MV has new options for improved modeling methods and interactive graphics display. Though the same models incorporated in RELAP5/MOD 4.0 are in RELAP5-MV, the significant difference of the latter is the interface for preparing the input deck. In this paper, RELAP5-MV is applied for the transient analysis of the primary system variation of thermal hydraulics parameters in primary loop under SBLOCA in AP1000 NPP. The upper limit of SBLOCA (10 inches is simulated in the cold leg of the reactor and the calculations performed up to a transient time of 450,000.0 s. The results obtained from RELAP5-MV are in good agreement with those of NOTRUMP code obtained by Westinghouse when compared under the same conditions. It can be easily inferred that RELAP5-MV, in a similar manner to RELAP5/MOD4.0, is suitable for simulating a SBLOCA scenario.

Westinghouse Small Modular Reactor passive safety system response to postulated events

International Nuclear Information System (INIS)

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

2012-01-01

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

Status of magnet power supply development for the APS [Advanced Photon Source] storage ring

International Nuclear Information System (INIS)

McGhee, D.

1989-01-01

To simplify installation and speed testing of the Advanced Photon Source (APS) storage ring magnets, vacuum chambers and magnet power supplies, a modular approach was developed. All but the dipole magnets are independently controlled. Pulse width modulated dc-to-dc converters are used to power the individual magnets, with 12-pulse power supplies providing the raw dc to the converters. A magnet support base is the heart of a module and may hold as many as 7 magnets with 8 individually powered coils. The dc-to-dc converters are part of each magnet base module. This paper will show the modular approach which is used for the storage ring magnet systems and will give the test results of the prototype topology for the dc-to-dc converters that are being built and tested to power 680 quadrupole and sextupole magnets. 4 refs., 11 figs., 1 tab

ACR-1000: Operator - based development

International Nuclear Information System (INIS)

Shalaby, B.; Alizadeh, A.

2007-01-01

Atomic Energy of Canada Limited (AECL) has adapted the successful features of CANDU * reactors to establish Generation III + Advanced CANDU Reactor T M (ACR T M) technology. The ACR-1000 T M nuclear power plant is an evolutionary product, starting with the strong base of CANDU reactor technology, coupled with thoroughly-demonstrated innovative features to enhance economics, safety, operability and maintainability. The ACR-1000 benefits from AECL's continuous-improvement approach to design, that enabled the traditional CANDU 6 product to compile an exceptional track record of on-time, on budget product delivery, and also reliable, high capacity-factor operation. The ACR-1000 engineering program has completed the basic plant design and has entered detailed pre-project engineering and formal safety analysis to prepare the preliminary (non-project-specific) safety case. The engineering program is strongly operator-based, and encompasses much more than traditional pre-project design elements. A team of utility-experienced operations and maintenance experts is embedded in the engineering team, to ensure that all design decisions, at the system and the component level, are taken with the owner-operator interest in mind. The design program emphasizes formal review of operating feedback, along with extensive operator participation in program management and execution. Design attention is paid to layout and access of equipment, to component and material selection, and to ensuring maximum ability for on-line maintenance. This enables the ACR-1000 to offer a three-year interval between scheduled maintenance outages, with a standard 21-day outage duration. SMART CANDU T M technology allows on-line monitoring and diagnostics to further enhance plant operation. Modules of the Advanced CANDU SMART technologies are already being back-fitted to current CANDU plants. As well as reviewing the ACR-1000 design features and their supporting background, the paper describes the status of

Comparison of a CCD and an APS for soft x-ray diffraction

OpenAIRE

Stewart, G.; Bates, R.; Blue, A.; Clark, A.; Dhesi, S.S.; Maneuski, D.; Marchal, J.; Steadman, P.; Tartoni, N.; Turchetta, R.

2011-01-01

We compare a new CMOS Active Pixel Sensor (APS) to a Princeton Instruments PIXIS-XO: 2048B Charge Coupled Device (CCD) with soft X-rays tested in a synchrotron beam line at the Diamond Light Source (DLS). Despite CCDs being established in the field of scientific imaging, APS are an innovative technology that offers advantages over CCDs. These include faster readout, higher operational temperature, in-pixel electronics for advanced image processing and reduced manufacturing cost.\\ud \\ud The AP...

Structuring the AP Art History Course

Science.gov (United States)

Herscher, Walter R.

2013-01-01

While AP (Advanced Placement) Art History may be taught within the art department in many schools, social studies teachers are equally capable of teaching the course well. They have the historical background to discuss the reasons for changes in art styles. A teacher's preparation is similar to teaching a course stressing political history,…

New Frontiers in Passive and Active Nanoantennas

DEFF Research Database (Denmark)

Arslanagic, Samel; Ziolkowski, Richard W.

2017-01-01

The articles included in this special section focus on several recent advances in the field of passive and active nanoantennas that employ not only traditional based realizations but also their new frontiers.......The articles included in this special section focus on several recent advances in the field of passive and active nanoantennas that employ not only traditional based realizations but also their new frontiers....

The impact of innovation and organizational factors on APS adoption: Evidence from the Dutch discrete parts industry

NARCIS (Netherlands)

A.P. van Hezewijk (Bart); M.F. van Assen (Marcel); S.L. van de Velde (Steef)

2003-01-01

textabstractAdvanced Planning and Scheduling (APS) systems have gained renewed interest from academics and practitioners. However, literature on APS adoption is scant. This study explores the impact of organizational and innovation related factors on the adoption of APS systems from a factors

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)

Characterization techniques for the high-brightness particle beams of the Advanced Photon Source (APS)

International Nuclear Information System (INIS)

Lumpkin, A.H.

1993-01-01

The Advanced Photon Source (APS) will be a third-generation synchrotron radiation (SR) user facility in the hard x-ray regime (10--100 keV). The design objectives for the 7-GeV storage ring include a positron beam natural emittance of 8 x 10 -9 m-rad at an average current of 100 mA. Proposed methods for measuring the transverse and longitudinal profiles will be described. Additionally, a research and development effort using an rf gun as a low-emittance source of electrons for injection into the 200- to 650-MeV linac subsystem is underway. This latter system is projected to produce electron beams with a normalized, rms emittance of ∼2 π mm-mrad at peak currents of near one hundred amps. This interesting characterization problem will also be briefly discussed. The combination of both source types within one laboratory facility will stimulate the development of diagnostic techniques in these parameter spaces

Initial performance assessment of the Westinghouse AP600 containment design and related safety issues

International Nuclear Information System (INIS)

Nicolette, V.F.; Washington, K.E.; Tills, J.L.

1991-01-01

This work summarizes the Westinghouse AP600 advanced reactor design assessment calculations performed to date with the CONTAIN code. Correlations for modeling the important heat transfer phenomena are discussed as well. A CONTAIN model of the AP600 was constructed for design basis accident (DBA) calculations. Insights gained from modeling of the smaller-scale Westinghouse Integral Test Facility were incorporated in the development of the AP600 model. The results of the DBA calculations are compared to the results of other researchers to serve as a point of reference for future severe accident calculations. The CONTAIN calculations are reviewed to examine several parameters/phenomena of interest. The results of the calculations are also used to identify limitations of the CONTAIN code regarding application to advanced reactor containment designs. The most recent heat transfer correlations available in the literature are assessed for use in the flow regimes and geometries applicable to the AP600. Use of one of these correlations in CONTAIN may allow for a more accurate assessment of the AP600

AP600 level of automation: United States utility perspective

International Nuclear Information System (INIS)

Bekkerman, A.Y.

1997-01-01

Design of the AP600 advanced nuclear plant man-machine interface system (M-MIS) is guided by the applicable requirements from the Utility Requirements Document (URD). However, the URD has left certain aspects of the M-MIS to be determined by the designer working together with utilities sponsoring the work. This is particularly true in the case of the level of automation to be designed into the M-MIS. Based on experience from currently operating plants, utilities have specified the identity and roles of personnel in the control room, which has led to establishing a number of level of automation issues for the AP600. The key role of automated computerized procedures in the AP600 automation has been determined and resolved. 5 refs

History of the APS Topical Group on Shock Compression of Condensed Matter

International Nuclear Information System (INIS)

Forbes, J W

2001-01-01

In order to provide broader scientific recognition and to advance the science of shock compressed condensed matter, a group of American Physical Society (APS) members worked within the Society to make this field an active part of the APS. Individual papers were presented at APS meetings starting in the 1940's and shock wave sessions were organized starting with the 1967 Pasadena meeting. Shock wave topical conferences began in 1979 in Pullman, WA. Signatures were obtained on a petition in 1984 from a balanced cross-section of the shock wave community to form an APS Topical Group (TG). The APS Council officially accepted the formation of the Shock Compression of Condensed Matter (SCCM) TG at its October 1984 meeting. This action firmly aligned the shock wave field with a major physical science organization. Most early topical conferences were sanctioned by the APS while those held after 1992 were official APS meetings. The topical group organizes a shock wave topical conference in odd numbered years while participating in shock wavehigh pressure sessions at APS general meetings in even numbered years

Noise analysis of a novel hybrid active-passive pixel sensor for medical X-ray imaging

International Nuclear Information System (INIS)

Safavian, N.; Izadi, M.H.; Sultana, A.; Wu, D.; Karim, K.S.; Nathan, A.; Rowlands, J.A.

2009-01-01

Passive pixel sensor (PPS) is one of the most widely used architectures in large area amorphous silicon (a-Si) flat panel imagers. It consists of a detector and a thin film transistor (TFT) acting as a readout switch. While the PPS is advantageous in terms of providing a simple and small architecture suitable for high-resolution imaging, it directly exposes the signal to the noise of data line and external readout electronics, causing significant increase in the minimum readable sensor input signal. In this work we present the operation and noise performance of a hybrid 3-TFT current programmed, current output active pixel sensor (APS) suitable for real-time X-ray imaging. The pixel circuit extends the application of a-Si TFT from conventional switching element to on-pixel amplifier for enhanced signal-to-noise ratio and higher imager dynamic range. The capability of operation in both passive and active modes as well as being able to compensate for inherent instabilities of the TFTs makes the architecture a good candidate for X-ray imaging modalities with a wide range of incoming X-ray intensities. Measurement and theoretical calculations reveal a value for input refferd noise below the 1000 electron noise limit for real-time fluoroscopy. (copyright 2009 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

Development of essential system technologies for advanced reactor

International Nuclear Information System (INIS)

Bae, Y. Y.; Hwang, Y. D.; Cho, B. H. and others

1999-03-01

Basic design of SMART adopts the new advanced technologies which were not applied in the existing 1000MWe PWR. However, the R and D experience on these advanced essential technologies is lacking in domestic nuclear industry. Recently, a research on these advanced technologies has been performed as a part of the mid-and-long term nuclear R and D program, but the research was limited only for the small scale fundamental study. The research on these essential technologies such as helically coiled tube steam generator, self pressurizer, core cooling by natural circulation required for the development of integral reactor SMART have not been conducted in full scale. This project, therefore, was performed for the development of analysis models and methodologies, system analysis and thermal hydraulic experiments on the essential technologies to be applied to the 300MWe capacity of integral reactor SMART and the advanced passive reactor expected to be developed in near future with the emphasis on experimental investigation. (author)

ACR-1000TM Project - Licensing Opportunities and Challenges

International Nuclear Information System (INIS)

Popov, N.; Doerffer, S.; Ion, R.; Hopwood, J.

2011-01-01

Atomic Energy of Canada Limited (AECL) has developed the Advanced CANDU Reactor TM 1 1000 (ACR-1000 TM ) as an evolutionary advancement of the current CANDU 6 reactor. The ACR-1000 design has evolved from AECL's in-depth knowledge of CANDU TM systems, components, and materials, as well as the experience and feedback received from owners and operators of CANDU plants. The ACR design retains the proven strengths and features of CANDU reactors, while incorporating innovations and state-of-the-art technology. It also features major improvements in economics, inherent safety characteristics, and performance, while retaining the proven benefits of the CANDU family of nuclear power plants. To ensure that the ACR design is compliant with Canadian and international requirements, regulatory pre-project reviews of the ACR-1000 (and ACR-700 TM 1 with lower output) were conducted early in the design work. The regulatory feedback from these pre-project regulatory reviews helped AECL to better understand regulatory expectations in Canada, US and the UK, and to make further advancements and improvements in the ACR design to meet the Canadian and international regulatory requirements. This paper provides an overview of the key design features of the ACR-1000 reactor design, and summary of the pre-project reviews by those above-mentioned regulatory bodies, demonstrating opportunities and challenges in licensing process of and pointing to the importance of efficient vendor-regulator interaction. (author)

Survey of the passive safety systems of the BWR 1000 concept from SIEMENS

Energy Technology Data Exchange (ETDEWEB)

Mattern, J; Brettschuh, W; Palavecino, C [SIEMENS, Energieerzeugung, Offenbach (Germany)

1996-12-01

Through the use of passive safety systems and components for accident control in addition to the active systems required for plant operation, a higher degree of safety against core-endangering conditions is achieved which is no longer ruled by complex system engineering dependent on power supply and activation by I and C systems. A low core power density and large water inventories stored inside the reactor pressure vessel as well as inside and outside the containment ensure good plant behaviour in the event of transients or accidents. These passive safety systems - which required neither electric power to function nor I and C systems for actuation, being activated solely on the basis of changes in process variables such as water level, pressure and temperature - provide a grace period of more than 5 days after the onset of accident conditions before manual intervention becomes necessary. 8 figs.

A phase 2 consortium (P2C) trial of 3-aminopyridine-2-carboxaldehyde thiosemicarbazone (3-AP) for advanced adenocarcinoma of the pancreas

Science.gov (United States)

Attia, Steven; Kolesar, Jill; Mahoney, Michelle R.; Pitot, Henry C.; Laheru, Daniel; Heun, James; Huang, Wei; Eickhoff, Jens; Erlichman, Charles

2015-01-01

Summary 3-Aminopyridine-2-carboxaldehyde thiosemicarbazone (3-AP, Triapine®) is a novel small molecule inhibitor of ribonucleotide reductase (RR) with clinical signs of activity in pancreatic cancer. Therefore, the Phase 2 Consortium (P2C) initiated a trial (two single stage studies with planned interim analysis) of 3-AP at 96 mg/m2 intravenously days 1–4 and 15–18 of a 28-day cycle in both chemotherapy-naive and gemcitabine-refractory (GR) patients with advanced pancreatic cancer. The primary endpoint was survival at six months (chemotherapy-naive) and four months (GR). Secondary endpoints were toxicity, response, overall survival, time to progression and mechanistic studies. Fifteen patients were enrolled including one chemotherapy-naïve and 14 GR. The chemotherapy-naïve patient progressed during cycle 1 with grade 3 and 4 toxicities. Of 14 GR patients, seven received two cycles, six received one cycle and one received eight cycles. Progression precluded further treatment in 11 GR patients. Additionally, one died of an ileus in cycle 1 considered related to treatment and two stopped treatment due to toxicity. Five GR patients had grade 4 toxicities possibly related to 3-AP and six GR patients had grade 3 fatigue. Toxicities and lack of meaningful clinical benefit prompted early study closure. Four-month survival in GR patients was 21% (95% CI: 8–58%). Correlative studies confirmed that 3-AP increased the percentage of S-phase buccal mucosal cells, the presence of multidrug resistance gene polymorphisms appeared to predict leukopenia, and baseline pancreatic tumor RR M2 expression was low relative to other tumors treated with 3-AP. In conclusion, this regimen appears inactive against predominantly GR pancreatic cancer. RR M2 protein may not have a critical role in the malignant potential of pancreatic cancer. PMID:18278438

The Demographic Wave: Rethinking Hispanic AP Trends

Science.gov (United States)

Edwards, Kelcey; Sawtell, Ellen

2013-01-01

Presented at the Advanced Placement Annual Conference (APAC) in Las Vegas, NV in July 2013. This presentation reviews new research examining the AP® experience of Hispanic graduates over the past decade. Topics include an in-depth look at the AP Spanish Language and Culture gateway hypothesis and trends in family characteristics such as parent…

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

International Nuclear Information System (INIS)

Wetzel, N.; Scharfe, A.

1998-01-01

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

Feasibility study on emergency passive habitability systems of SPWR

International Nuclear Information System (INIS)

Obata, H.; Tabata, H.; Urakami, M.; Naito, T.

2000-01-01

The major characteristic of the Simplified Pressurized Water Reactor (SPWR) is that safety systems for the emergency core cooling and the core decay heat removal functions are achieved by passive equipment. The AP600 developed in the U.S adopts passive emergency habitability system for the main control room (MCR) and the electrical equipment rooms (EER) by using the concrete of the structures as a heat sink. For the SPWR, alternative natural circulation cooling systems have been investigated: for MCR cooling, a cold water reservoir is used as heat sink; for EER cooling, outside air is instead employed. The distribution of the air-velocity and temperature in those rooms were calculated by using a three-dimensional thermal fluid analysis code. The authors verified the conceptual feasibility of these systems as the emergency passive habitability systems in the SPWR. (author)

Passive radon daughter dosimeters

International Nuclear Information System (INIS)

McElroy, R.G.C.; Johnson, J.R.

1986-03-01

On the basis of an extensive review of the recent literature concerning passive radon daughter dosimeters, we have reached the following conclusions: 1) Passive dosimeters for measuring radon are available and reliable. 2) There does not presently exist an acceptable passive dosimeter for radon daughters. There is little if any hope for the development of such a device in the foreseeable future. 3) We are pessimistic about the potential of 'semi-passive dosimeters' but are less firm about stating categorically that these devices cannot be developed into a useful radon daughter dosimeter. This report documents and justifies these conclusions. It does not address the question of the worker's acceptance of these devices because at the present time, no device is sufficiently advanced for this question to be meaningful. 118 refs

Balancing passive and active systems for evolutionary water cooled reactors

International Nuclear Information System (INIS)

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

1999-01-01

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

SEPTUM MAGNET DESIGN FOR THE APS-U

Energy Technology Data Exchange (ETDEWEB)

Abliz, M.; Jaski, M.; Xiao, A.; Wienands, U.; Cease, H.; Borland, M.; Decker, G.; Kerby, J.

2017-06-25

The Advanced Photon Source is in the process of upgrading its storage ring from a double-bend to a multi-bend lattice as part of the APS Upgrade Project (APS-U). A swap-out injection scheme is planned for the APS-U to keep a constant beam current and to enable a small dynamic aperture. A septum magnet with a minimum thickness of 2 mm and an injection field of 1.06 T has been designed, delivering the required total deflecting angle is 89 mrad with a ring energy of 6 GeV. The stored beam chamber has an 8 mm x 6 mm super-ellipsoidal aperture. The magnet is straight; however, it is tilted in yaw, roll, and pitch from the stored beam chamber to meet the on axis swap out injection requirements for the APS-U lattice. In order to minimize the leakage field inside the stored beam chamber, four different techniques were utilized in the design. As a result, the horizontal deflecting angle of the stored beam was held to only 5 µrad, and the integrated skew quadrupole inside the stored beam chamber was held to 0.09 T. The detailed techniques that were applied to the design, field multipoles, and resulting trajectories of the injected and stored beams are reported.

Updating AP Potential™ Expectancy Tables Involving PSAT/NMSQT® Writing. Research Notes. RN-35

Science.gov (United States)

Ewing, Maureen; Camara, Wayne J.; Millsap, Roger E.; Milewski, Glenn B.

2007-01-01

AP Potential™ is a data-driven tool offered by the College Board that uses scores from the PSAT/NMSQT® to identify students who have the potential to succeed in Advanced Placement Program® (AP®) courses (College Board, 2007). Research showing a moderate-to-strong correlation between PSAT/NMSQT scores and AP Exam scores serves as the basis for this…

Center for Geometrisk Metrologi, CGM ApS

DEFF Research Database (Denmark)

De Chiffre, Leonardo

Denne årsberetning omfatter CGM ApS' akkrediterede virksomhed i kalenderåret 2002. Årsberetningen er udarbejdet til DANAK (Dansk Akkreditering, Erhvervsfremme Styrelsen), som led i opfyldelsen af laboratoriets informationspligt i henhold til gældende regler (Teknisk Forskrift Nr. TF4 af 2000...

Simplified analysis of passive residual heat removal systems for small size PWR's

International Nuclear Information System (INIS)

Botelho, D.A.

1992-02-01

The function and general objectives of a passive residual heat removal system for small size PWR's are defined. The characteristic configuration, the components and the operation modes of this system are concisely described. A preliminary conceptual specification of this system, for a small size PWR of 400 MW thermal, is made analogous to the decay heat removal system of the AP-600 reactor. It is shown by analytic models that such passive systems can dissipate 2% of nominal power within the thermal limits allowed to the reactor fuel elements. (author)

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

International Nuclear Information System (INIS)

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

1993-01-01

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

Advanced configuration of hybrid passive filter for reactive power and harmonic compensation

OpenAIRE

Kececioglu, O. Fatih; Acikgoz, Hakan; Sekkeli, Mustafa

2016-01-01

Harmonics is one of the major power quality problems for power systems. The harmonics can be eliminated by power filters such as passive, active, and hybrid. In this study, a new passive filter configuration has been improved in addition to the existing passive filter configurations. Conventional hybrid passive filters are not successful to compensate rapidly changing reactive power demand. The proposed configure are capable of compensating both harmonics and reactive power at the same time. ...

Reliability prediction for the vehicles equipped with advanced driver assistance systems (ADAS and passive safety systems (PSS

Directory of Open Access Journals (Sweden)

Balbir S. Dhillon

2012-10-01

Full Text Available The human error has been reported as a major root cause in road accidents in today’s world. The human as a driver in road vehicles composed of human, mechanical and electrical components is constantly exposed to changing surroundings (e.g., road conditions, environmentwhich deteriorate the driver’s capacities leading to a potential accident. The auto industries and transportation authorities have realized that similar to other complex and safety sensitive transportation systems, the road vehicles need to rely on both advanced technologies (i.e., Advanced Driver Assistance Systems (ADAS and Passive Safety Systems (PSS (e.g.,, seatbelts, airbags in order to mitigate the risk of accidents and casualties. In this study, the advantages and disadvantages of ADAS as active safety systems as well as passive safety systems in road vehicles have been discussed. Also, this study proposes models that analyze the interactions between human as a driver and ADAS Warning and Crash Avoidance Systems and PSS in the design of vehicles. Thereafter, the mathematical models have been developed to make reliability prediction at any given time on the road transportation for vehicles equipped with ADAS and PSS. Finally, the implications of this study in the improvement of vehicle designs and prevention of casualties are discussed.

A second-generation superconducting undulator cryostat for the APS

Science.gov (United States)

Fuerst, J.; Hasse, Q.; Ivanyushenkov, Y.; Kasa, M.; Shiroyanagi, Y.

2017-12-01

A second-generation cryocooler-based cryostat has been designed and built to support a new helically wound superconducting undulator (SCU) magnet for the Advanced Photon Source (APS) at Argonne National Laboratory (ANL). The design represents an evolution of existing SCU cryostats currently in operation in the APS storage ring. Value engineering and lessons learned have resulted in a smaller, cheaper, and simpler cryostat design compatible with existing planar magnets as well as the new helically wound device. We describe heat load and quench response results, design and operational details, and the “build-to-spec” procurement strategy.

Passive heat removal characteristics of SMART

Energy Technology Data Exchange (ETDEWEB)

Seo, Jae Kwang; Kang, Hyung Seok; Yoon, Joo Hyun; Kim, Hwan Yeol; Cho, Bong Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1998-12-31

A new advanced integral reactor of 330 MWt thermal capacity named SMART (System-Integrated Modular Advanced Reactor) is currently under development in Korea Atomic Energy Research Institute (KAERI) for multi-purpose applications. Modular once-through steam generator (SG) and self-pressurizing pressurizer equipped with wet thermal insulator and cooler are essential components of the SMART. The SMART provides safety systems such as Passive Residual Heat Removal System (PRHRS). In this study, a computer code for performance analysis of the PRHRS is developed by modeling relevant components and systems of the SMART. Using this computer code, a performance analysis of the PRHRS is performed in order to check whether the passive cooling concept using the PRHRS is feasible. The results of the analysis show that PRHRS of the SMART has excellent passive heat removal characteristics. 2 refs., 4 figs., 1 tab. (Author)

Component-Level Prognostics Health Management Framework for Passive Components - Advanced Reactor Technology Milestone: M2AT-15PN2301043

Energy Technology Data Exchange (ETDEWEB)

Ramuhalli, Pradeep; Roy, Surajit; Hirt, Evelyn H.; Prowant, Matthew S.; Pitman, Stan G.; Tucker, Joseph C.; Dib, Gerges; Pardini, Allan F.

2015-06-19

This report describes research results to date in support of the integration and demonstration of diagnostics technologies for prototypical advanced reactor passive components (to establish condition indices for monitoring) with model-based prognostics methods. Achieving this objective will necessitate addressing several of the research gaps and technical needs described in previous technical reports in this series.

APS-U LATTICE DESIGN FOR OFF-AXIS ACCUMULATION

Energy Technology Data Exchange (ETDEWEB)

Sun, Yipeng; Borland, M.; Lindberg, R.; Sajaev, V.

2017-06-25

A 67-pm hybrid-seven-bend achromat (H7BA) lattice is being proposed for a future Advanced Photon Source (APS) multi-bend-achromat (MBA) upgrade project. This lattice design pushes for smaller emittance and requires use of a swap-out (on-axis) injection scheme due to limited dynamic acceptance. Alternate lattice design work has also been performed for the APS upgrade to achieve better beam dynamics performance than the nominal APS MBA lattice, in order to allow off-axis accumulation. Two such alternate H7BA lattice designs, which target a still-low emittance of 90 pm, are discussed in detail in this paper. Although the single-particle-dynamics performance is good, simulations of collective effects indicate that surprising difficulty would be expected accumulating high single-bunch charge in this lattice. The brightness of the 90-pm lattice is also a factor of two lower than the 67-pm H7BA lattice.

The link between exercise and titin passive stiffness.

Science.gov (United States)

Lalande, Sophie; Mueller, Patrick J; Chung, Charles S

2017-09-01

What is the topic of this review? This review focuses on how in vivo and molecular measurements of cardiac passive stiffness can predict exercise tolerance and how exercise training can reduce cardiac passive stiffness. What advances does it highlight? This review highlights advances in understanding the relationship between molecular (titin-based) and in vivo (left ventricular) passive stiffness, how passive stiffness modifies exercise tolerance, and how exercise training may be therapeutic for cardiac diseases with increased passive stiffness. Exercise can help alleviate the negative effects of cardiovascular disease and cardiovascular co-morbidities associated with sedentary behaviour; this may be especially true in diseases that are associated with increased left ventricular passive stiffness. In this review, we discuss the inverse relationship between exercise tolerance and cardiac passive stiffness. Passive stiffness is the physical property of cardiac muscle to produce a resistive force when stretched, which, in vivo, is measured using the left ventricular end diastolic pressure-volume relationship or is estimated using echocardiography. The giant elastic protein titin is the major contributor to passive stiffness at physiological muscle (sarcomere) lengths. Passive stiffness can be modified by altering titin isoform size or by post-translational modifications. In both human and animal models, increased left ventricular passive stiffness is associated with reduced exercise tolerance due to impaired diastolic filling, suggesting that increased passive stiffness predicts reduced exercise tolerance. At the same time, exercise training itself may induce both short- and long-term changes in titin-based passive stiffness, suggesting that exercise may be a treatment for diseases associated with increased passive stiffness. Direct modification of passive stiffness to improve exercise tolerance is a potential therapeutic approach. Titin passive stiffness itself may

Advanced fuel cycles of WWER-1000 reactors

International Nuclear Information System (INIS)

Lunin, G.; Novikov, A.; Pavlov, V.; Pavlovichev, A.

2003-01-01

The present paper considers characteristics of fuel cycles for the WWER-1000 reactor satisfying the following conditions: duration of the campaign at the nominal power is extended from 250 EFPD up to 470 and more ones; fuel enrichment does not exceed 5 wt.%; fuel assemblies maximum burnup does not exceed 55 MWd/kgHM. Along with uranium fuel, the use of mixed Uranium-Plutonium fuel is considered. Calculations were conducted by codes TVS-M, BIPR-7A and PERMAK-A developed in the RRC Kurchatov Institute, verified for the calculations of uranium fuel and certified by GAN RF

Tank 241-AP-106, Grab samples, 6AP-98-1, 6AP-98-2 and 6AP-98-3 Analytical results for the final report

Energy Technology Data Exchange (ETDEWEB)

FULLER, R.K.

1999-02-23

This document is the final report for tank 241-AP-106 grab samples. Three grab samples 6AP-98-1, 6AP-98-2 and 6AP-98-3 were taken from riser 1 of tank 241-AP-106 on May 28, 1998 and received by the 222-S Laboratory on May 28, 1998. Analyses were performed in accordance with the ''Compatability Grab Sampling and Analysis Plan'' (TSAP) (Sasaki, 1998) and the ''Data Quality Objectives for Tank Farms Waste Compatability Program (DQO). The analytical results are presented in the data summary report. No notification limits were exceeded. The request for sample analysis received for AP-106 indicated that the samples were polychlorinated biphenyl (PCB) suspects. The results of this analysis indicated that no PCBs were present at the Toxic Substance Control Act (TSCA) regulated limit of 50 ppm. The results and raw data for the PCB analysis are included in this document.

Tank 241-AP-106, Grab samples, 6AP-98-1, 6AP-98-2 and 6AP-98-3 Analytical results for the final report

International Nuclear Information System (INIS)

FULLER, R.K.

1999-01-01

This document is the final report for tank 241-AP-106 grab samples. Three grab samples 6AP-98-1, 6AP-98-2 and 6AP-98-3 were taken from riser 1 of tank 241-AP-106 on May 28, 1998 and received by the 222-S Laboratory on May 28, 1998. Analyses were performed in accordance with the ''Compatability Grab Sampling and Analysis Plan'' (TSAP) (Sasaki, 1998) and the ''Data Quality Objectives for Tank Farms Waste Compatability Program (DQO). The analytical results are presented in the data summary report. No notification limits were exceeded. The request for sample analysis received for AP-106 indicated that the samples were polychlorinated biphenyl (PCB) suspects. The results of this analysis indicated that no PCBs were present at the Toxic Substance Control Act (TSCA) regulated limit of 50 ppm. The results and raw data for the PCB analysis are included in this document

Investigation of the falling water flow with evaporation for the passive containment cooling system and its scaling-down criteria

Science.gov (United States)

Li, Cheng; Li, Junming; Li, Le

2018-02-01

Falling water evaporation cooling could efficiently suppress the containment operation pressure during the nuclear accident, by continually removing the core decay heat to the atmospheric environment. In order to identify the process of large-scale falling water evaporation cooling, the water flow characteristics of falling film, film rupture and falling rivulet were deduced, on the basis of previous correlation studies. The influences of the contact angle, water temperature and water flow rates on water converge along the flow direction were then numerically obtained and results were compared with the data for AP1000 and CAP1400 nuclear power plants. By comparisons, it is concluded that the water coverage fraction of falling water could be enhanced by either reducing the surface contact angle or increasing the water temperature. The falling water flow with evaporation for AP1000 containment was then calculated and the feature of its water coverage fraction was analyzed. Finally, based on the phenomena identification of falling water flow for AP1000 containment evaporation cooling, the scaling-down is performed and the dimensionless criteria were obtained.

Small break loss of coolant accident analysis of advanced PWR plant designs utilizing DVI line venturis

International Nuclear Information System (INIS)

Kemper, Robert M.; Gagnon, Andre F.; McNamee, Kevin; Cheung, Augustine C.

1995-01-01

The Westinghouse Advanced Passive and evolutionary Pressurizer Water Reactors (i.e. AP600 and APWR) incorporate direct vessel injection (DVI) of emergency core coolant as a means of minimizing the potential spilling of emergency core cooling water during a loss of coolant accident (LOCA). As a result, the most limiting small break LOCA (SBLOCA) event for these designs, with respect core inventory makeup capability, is a postulated double ended rupture of one of the DVI lines. This paper presents the results of a design optimization study that examines the installation of a venturi in the DVI line as a means of limiting the reactor coolant lost from the reactor vessel. The comparison results demonstrate that by incorporating a properly sized venturi in the DVI line, core uncovery concerns as a result of a DVI line break can be eliminated for both the AP600 and APWR plants. (author)

Thrombotic risk assessment in APS: the Global APS Score (GAPSS).

Science.gov (United States)

Sciascia, S; Bertolaccini, M L

2014-10-01

Recently, we developed a risk score for antiphospholipid syndrome (APS) (Global APS Score or GAPSS). This score derived from the combination of independent risk factors for thrombosis and pregnancy loss, taking into account the antiphospholipid antibodies (aPL) profile (criteria and non-criteria aPL), the conventional cardiovascular risk factors, and the autoimmune antibodies profile. We demonstrate that risk profile in APS can be successfully assessed, suggesting that GAPSS can be a potential quantitative marker of APS-related clinical manifestations. © The Author(s) 2014 Reprints and permissions: sagepub.co.uk/journalsPermissions.nav.

Advances in passive neutron instruments for safeguards use

International Nuclear Information System (INIS)

Menlove, H.O.; Krick, M.S.; Langner, D.G.; Miller, M.C.; Stewart, J.E.

1994-01-01

Passive neutron and other nondestructive assay techniques have been used extensively by the International Atomic Energy Agency to verify plutonium metal, powder, mixed oxide, pellets, rods, assemblies, scrap, and liquids. Normally, the coincidence counting rate is used to measure the 240 Pu-effective mass and gamma-ray spectrometry or mass spectrometry is used to verify the plutonium isotopic ratios. During the past few years, the passive neutron detectors have been installed in plants and operated in the unattended/continuous mode. These radiation data with time continuity have made it possible to use the totals counting rate to monitor the movement of nuclear material. Monte Carlo computer codes have been used to optimize the detector designs for specific applications. The inventory sample counter (INVS-III) has been designed to have a higher efficiency (43%) and a larger uniform counting volume than the original INVS. Data analyses techniques have been developed, including the ''known alpha'' and ''known multiplication'' methods that depend on the sample. For scrap and other impure or poorly characterized samples, we have developed multiplicity counting, initially implemented in the plutonium scrap multiplicity counter. For large waste containers such as 200-L drums, we have developed the add-a-source technique to give accurate corrections for the waste-matrix materials. This paper summarizes recent developments in the design and application of passive neutron assay systems

X-ray optics simulation and beamline design for the APS upgrade

Science.gov (United States)

Shi, Xianbo; Reininger, Ruben; Harder, Ross; Haeffner, Dean

2017-08-01

The upgrade of the Advanced Photon Source (APS) to a Multi-Bend Achromat (MBA) will increase the brightness of the APS by between two and three orders of magnitude. The APS upgrade (APS-U) project includes a list of feature beamlines that will take full advantage of the new machine. Many of the existing beamlines will be also upgraded to profit from this significant machine enhancement. Optics simulations are essential in the design and optimization of these new and existing beamlines. In this contribution, the simulation tools used and developed at APS, ranging from analytical to numerical methods, are summarized. Three general optical layouts are compared in terms of their coherence control and focusing capabilities. The concept of zoom optics, where two sets of focusing elements (e.g., CRLs and KB mirrors) are used to provide variable beam sizes at a fixed focal plane, is optimized analytically. The effects of figure errors on the vertical spot size and on the local coherence along the vertical direction of the optimized design are investigated.

Correction magnet power supplies for APS machine

International Nuclear Information System (INIS)

Kang, Y.G.

1991-01-01

The Advanced Photon Source machine requires a number of correction magnets; five kinds for the storage ring, two for the injector synchrotron, and two for the positron accumulator ring. Three types of bipolar power supply will be used for all the correction magnets. This paper describes the design aspects and considerations for correction magnet power supplies for the APS machine. 3 refs., 3 figs., 1 tab

7 CFR 1000.91-1000.92 - [Reserved

Science.gov (United States)

2010-01-01

... 7 Agriculture 9 2010-01-01 2009-01-01 true [Reserved] 1000.91-1000.92 Section 1000.91-1000.92 Agriculture Regulations of the Department of Agriculture (Continued) AGRICULTURAL MARKETING SERVICE (Marketing... Miscellaneous Provisions §§ 1000.91-1000.92 [Reserved] ...

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

International Nuclear Information System (INIS)

2013-09-01

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