ADVANCED SEISMIC BASE ISOLATION METHODS FOR MODULAR REACTORS

International Nuclear Information System (INIS)

Blanford, E.; Keldrauk, E.; Laufer, M.; Mieler, M.; Wei, J.; Stojadinovic, B.; Peterson, P.F.

2010-01-01

Advanced technologies for structural design and construction have the potential for major impact not only on nuclear power plant construction time and cost, but also on the design process and on the safety, security and reliability of next generation of nuclear power plants. In future Generation IV (Gen IV) reactors, structural and seismic design should be much more closely integrated with the design of nuclear and industrial safety systems, physical security systems, and international safeguards systems. Overall reliability will be increased, through the use of replaceable and modular equipment, and through design to facilitate on-line monitoring, in-service inspection, maintenance, replacement, and decommissioning. Economics will also receive high design priority, through integrated engineering efforts to optimize building arrangements to minimize building heights and footprints. Finally, the licensing approach will be transformed by becoming increasingly performance based and technology neutral, using best-estimate simulation methods with uncertainty and margin quantification. In this context, two structural engineering technologies, seismic base isolation and modular steel-plate/concrete composite structural walls, are investigated. These technologies have major potential to (1) enable standardized reactor designs to be deployed across a wider range of sites, (2) reduce the impact of uncertainties related to site-specific seismic conditions, and (3) alleviate reactor equipment qualification requirements. For Gen IV reactors the potential for deliberate crashes of large aircraft must also be considered in design. This report concludes that base-isolated structures should be decoupled from the reactor external event exclusion system. As an example, a scoping analysis is performed for a rectangular, decoupled external event shell designed as a grillage. This report also reviews modular construction technology, particularly steel-plate/concrete construction using

ADVANCED SEISMIC BASE ISOLATION METHODS FOR MODULAR REACTORS

Energy Technology Data Exchange (ETDEWEB)

E. Blanford; E. Keldrauk; M. Laufer; M. Mieler; J. Wei; B. Stojadinovic; P.F. Peterson

2010-09-20

Advanced technologies for structural design and construction have the potential for major impact not only on nuclear power plant construction time and cost, but also on the design process and on the safety, security and reliability of next generation of nuclear power plants. In future Generation IV (Gen IV) reactors, structural and seismic design should be much more closely integrated with the design of nuclear and industrial safety systems, physical security systems, and international safeguards systems. Overall reliability will be increased, through the use of replaceable and modular equipment, and through design to facilitate on-line monitoring, in-service inspection, maintenance, replacement, and decommissioning. Economics will also receive high design priority, through integrated engineering efforts to optimize building arrangements to minimize building heights and footprints. Finally, the licensing approach will be transformed by becoming increasingly performance based and technology neutral, using best-estimate simulation methods with uncertainty and margin quantification. In this context, two structural engineering technologies, seismic base isolation and modular steel-plate/concrete composite structural walls, are investigated. These technologies have major potential to (1) enable standardized reactor designs to be deployed across a wider range of sites, (2) reduce the impact of uncertainties related to site-specific seismic conditions, and (3) alleviate reactor equipment qualification requirements. For Gen IV reactors the potential for deliberate crashes of large aircraft must also be considered in design. This report concludes that base-isolated structures should be decoupled from the reactor external event exclusion system. As an example, a scoping analysis is performed for a rectangular, decoupled external event shell designed as a grillage. This report also reviews modular construction technology, particularly steel-plate/concrete construction using

Preliminary feasibility study of modular reactors

International Nuclear Information System (INIS)

Yamaji, Kenji

1987-01-01

In the future, electric utilities will be required to make a switch-over to a more flexible and dynamic form of power supply due to the slowing growth of power demand, increasing uncertainty, the stagnating economy of increasing scale, the bottleneck of transmission and so on. Nuclear technology would also be required to adapt to this changing environment surrounding its development. The long term prospect of energy demand and nuclear power growth, and the evolution of commercial reactors in Japan are shown. The design of 1,300 MWe advanced LWRs has been completed, and as the reactors of next generation, the ultralarge LWRs of 1,500 - 1,800 MWe are suggested. However, there can be an alternative future for nuclear power development, and in this paper, the possibility for altering the image of conventional nuclear power technology by developing modular reactors which are economical even at small capacity, and can be sited in urban areas just like conventional thermal power plants is examined. The factors for the economical evaluation of modular reactors, learning effect and scale effect on the economy, the case study on a modular high temperature reactor designed by Interatom-GHT, and the possibility of siting in urban areas due to the system of inherent safety are reported. (Kako, I.)

Design study on sodium-cooled middle-scale modular reactor

International Nuclear Information System (INIS)

Shimakawa, Yoshio; Nibe, Nobuaki; Hori, Toru

2002-05-01

In Phase 1 of the 'Feasibility Study on Commercialized Fast Reactor Cycle Systems (F/S)', an advanced loop type reactor has been selected as a promising concept of sodium-cooled middle-scale modular reactor, which has a possibility to fulfill the design requirements of the F/S. In Phase 2 of the F/S, it is planed to precede a preliminary conceptual design of a sodium-cooled middle-scale modular reactor based on the design of the advanced loop type reactor. Through the design study, it is intended to construct such a plant concept that can show its attraction and competitiveness as a commercialized reactor. This report summarizes the results of the design study on the sodium-cooled middle-scale modular reactor performed in JFY2001, which is the first year of Phase 2. As the construction cost of the sodium-cooled middle-scale modular reactor, which has been constructed in Phase 1, was about 10% higher than that of the sodium-cooled large-scale reactor, a new concept of the middle-scale modular reactor, which is expected to be equal to the large-scale reactor from a viewpoint of economic competitiveness, has been re-constructed based on the design of the advanced loop type reactor. After that, fundamental specifications of main systems and components for the new concept have been set, and critical subjects related to safety, structural integrity, thermal hydraulics, operability, maintainability and economy have been examined and evaluated. As a result of this study, the plant concept of the sodium-cooled middle-scale modular reactor has been constructed, which has a prospect to satisfy the economic goal (construction cost: less than 200,000 yens/kWe, etc.) and has a prospect to solve the critical subjects. From now on, reflecting the results of elemental experiments, the preliminary conceptual design of this plant will be preceded toward the selection for narrowing down candidate concepts at the end of Phase 2. (author)

Multi-Application Small Light Water Reactor Final Report

International Nuclear Information System (INIS)

Modro, S.M.; Fisher, J.E.; Weaver, K.D.; Reyes, J.N.; Groome, J.T.; Babka, P.; Carlson, T.M.

2003-01-01

The Multi-Application Small Light Water Reactor (MASLWR) project was conducted under the auspices of the Nuclear Energy Research Initiative (NERI) of the U.S. Department of Energy (DOE). The primary project objectives were to develop the conceptual design for a safe and economic small, natural circulation light water reactor, to address the economic and safety attributes of the concept, and to demonstrate the technical feasibility by testing in an integral test facility. This report presents the results of the project. After an initial exploratory and evolutionary process, as documented in the October 2000 report, the project focused on developing a modular reactor design that consists of a self-contained assembly with a reactor vessel, steam generators, and containment. These modular units would be manufactured at a single centralized facility, transported by rail, road, and/or ship, and installed as a series of self-contained units. This approach also allows for staged construction of an NPP and ''pull and replace'' refueling and maintenance during each five-year refueling cycle. Development of the baseline design concept has been sufficiently completed to determine that it complies with the safety requirements and criteria, and satisfies the major goals already noted. The more significant features of the baseline single-unit design concept include: (1) Thermal Power--150 MWt; (2) Net Electrical Output--35 MWe; (3) Steam Generator Type--Vertical, helical tubes; (4) Fuel UO 2 , 8% enriched; (5) Refueling Intervals--5 years; (6) Life-Cycle--60 years. The economic performance was assessed by designing a power plant with an electric generation capacity in the range of current and advanced evolutionary systems. This approach allows for direct comparison of economic performance and forms a basis for further evaluation, economic and technical, of the proposed design and for the design evolution towards a more cost competitive concept. Applications such as cogeneration

Modular helium reactor for non-electric applications

International Nuclear Information System (INIS)

Shenoy, A.

1997-01-01

The high temperature gas-cooled Modular Helium Reactor (MHR) is an advanced, high efficiency reactor system which can play a vital role in meeting the future energy needs of the world by contributing not only to the generation of electric power, but also the non-electric energy traditionally served by fossil fuels. This paper summarizes work done over 20 years, by several people at General Atomics, how the Modular Helium Reactor can be integrated to provide different non-electric applications during Process Steam/Cogeneration for industrial application, Process Heat for transportation fuel development and Hydrogen Production for various energy applications. The MHR integrates favorably into present petrochemical and primary metal process industries, heavy oil recovery, and future shale oil recovery and synfuel processes. The technical fit of the Process Steam/Cogeneration Modular Helium Reactor (PS/C-MHR) into these processes is excellent, since it can supply the required quantity and high quality of steam without fossil superheating. 12 refs, 25 figs, 2 tabs

Systems analysis for modular versus multi-beam HIF drivers

International Nuclear Information System (INIS)

Meier, W.R.; Logan, B.G.

2004-01-01

Previous modeling for HIF drivers concentrated on designs in which 100 or more beams are grouped in an array and accelerated through a common set of induction cores. The total beam energy required by the target is achieved by the combination of final ion energy, current per beam and number of beams. Economic scaling favors a large number of small (∼1 cm dia.) beams. An alternative architecture has now been investigated, which we refer to as a modular driver. In this case, the driver is subdivided into many (>10) independent accelerators with one or many beams each. A key objective of the modular driver approach is to be able to demonstrate all aspects of the driver (source-to-target) by building a single, lower cost module compared to a full-scale, multi-beam driver. We consider and compare several design options for the modular driver including single-beam designs with solenoid instead of quadrupole magnets in order to transport the required current per module in a single beam, solenoid/quad combinations, and multi-beam, all-quad designs. The drivers are designed to meet the requirements of the hybrid target, which can accommodate a larger spot size than the distributed radiator target that was used for the Robust Point Design. We compare the multi-beam and modular driver configuration for a variety and assumptions and identify key technology advances needed for the modular design

System Design of a Supercritical CO_2 cooled Micro Modular Reactor

International Nuclear Information System (INIS)

Kim, Seong Gu; Cho, Seongkuk; Yu, Hwanyeal; Kim, Yonghee; Jeong, Yong Hoon; Lee, Jeong Ik

2014-01-01

Small modular reactor (SMR) systems that have advantages of little initial capital cost and small restriction on construction site are being developed by many research organizations around the world. Existing SMR concepts have the same objective: to achieve compact size and a long life core. Most of small modular reactors have much smaller size than the large nuclear power plant. However, existing SMR concepts are not fully modularized. This paper suggests a complete modular reactor with an innovative concept for reactor cooling by using a supercritical carbon dioxide. The authors propose the supercritical CO_2 Brayton cycle (S-CO_2 cycle) as a power conversion system to achieve small volume of power conversion unit (PCU) and to contain the reactor core and PCU in one vessel. A conceptual design of the proposed small modular reactor was developed, which is named as KAIST Micro Modular Reactor (MMR). The supercritical CO_2 Brayton cycle for the S-CO_2 cooled reactor core was optimized and the size of turbomachinery and heat exchanger were estimated preliminary. The nuclear fuel composed with UN was proposed and the core lifetime was obtained from a burnup versus reactivity calculation. Furthermore, a system layout with fully passive safety systems for both normal operation and emergency operation was proposed. (author)

Low power modular power generating reactors or Small Modular Reactors (SMR)

International Nuclear Information System (INIS)

Chenais, Jacques

2016-01-01

Electronuclear reactors were small reactors at the beginning, and then tend to be always bigger and more powerful, but since some recent times, several countries specialized in reactor design and fabrication (USA, Russia, China, and South Korea) have been developing Small Modular Reactors (SMR) of less than 300 MW. As France has already produced feasibility studies and is about to launch a SMR development programme, the author comments some specific aspects of this new architecture of reactors, characterises the targeted markets, gives an overview of the various more or less advanced existing concepts: a floating barge in Russia, the SMART 100 MW project in South Korea, several concepts in the USA (the mPower 125 MW, the NuScale 45 MW, the Westinghouse 225 MW, and the HI-SMUR 160 MW projects), the ACP 100 MW in China, the CAREM 27 MW in Argentina. French projects developed by the CEA, EDF, Areva and DCNS are then presented

Adaptive multi-resolution Modularity for detecting communities in networks

Science.gov (United States)

Chen, Shi; Wang, Zhi-Zhong; Bao, Mei-Hua; Tang, Liang; Zhou, Ji; Xiang, Ju; Li, Jian-Ming; Yi, Chen-He

2018-02-01

Community structure is a common topological property of complex networks, which attracted much attention from various fields. Optimizing quality functions for community structures is a kind of popular strategy for community detection, such as Modularity optimization. Here, we introduce a general definition of Modularity, by which several classical (multi-resolution) Modularity can be derived, and then propose a kind of adaptive (multi-resolution) Modularity that can combine the advantages of different Modularity. By applying the Modularity to various synthetic and real-world networks, we study the behaviors of the methods, showing the validity and advantages of the multi-resolution Modularity in community detection. The adaptive Modularity, as a kind of multi-resolution method, can naturally solve the first-type limit of Modularity and detect communities at different scales; it can quicken the disconnecting of communities and delay the breakup of communities in heterogeneous networks; and thus it is expected to generate the stable community structures in networks more effectively and have stronger tolerance against the second-type limit of Modularity.

Small Modular Reactors (468th Brookhaven Lecture)

International Nuclear Information System (INIS)

Bari, Robert

2011-01-01

With good reason, much more media attention has focused on nuclear power plants than solar farms, wind farms, or hydroelectric plants during the past month and a half. But as nations around the world demand more energy to power everything from cell phone batteries to drinking water pumps to foundries, nuclear plants are the only non-greenhouse-gas producing option that can be built to operate almost anywhere, and can continue to generate power during droughts, after the sun sets, and when winds die down. To supply this demand for power, designers around the world are competing to develop more affordable nuclear reactors of the future: small modular reactors. Brookhaven Lab is working with DOE to ensure that these reactors are designed to be safe for workers, members of surrounding communities, and the environment and to ensure that the radioactive materials and technology will only be used for peaceful purposes, not weapons. In his talk, Bari will discuss the advantages and challenges of small modular reactors and what drives both international and domestic interest in them. He will also explain how Brookhaven Lab and DOE are working to address the challenges and provide a framework for small modular reactors to be commercialized.

Small modular reactor (SMR) development plan in Korea

Energy Technology Data Exchange (ETDEWEB)

Shin, Yong-Hoon, E-mail: chaotics@snu.ac.kr; Park, Sangrok; Kim, Byong Sup; Choi, Swongho; Hwang, Il Soon [Nuclear Transmutation Energy Research Center, Seoul National University, Bldg.31-1, 1 Gwanak-ro, Gwanak-gu, Seoul, Republic of Korea, 151-742 (Korea, Republic of)

2015-04-29

Since the first nuclear power was engaged in Korean electricity grid in 1978, intensive research and development has been focused on localization and standardization of large pressurized water reactors (PWRs) aiming at providing Korean peninsula and beyond with economical and safe power source. With increased priority placed on the safety since Chernobyl accident, Korean nuclear power R and D activity has been diversified into advanced PWR, small modular PWR and generation IV reactors. After the outbreak of Fukushima accident, inherently safe small modular reactor (SMR) receives growing interest in Korea and Europe. In this paper, we will describe recent status of evolving designs of SMR, their advantages and challenges. In particular, the conceptual design of lead-bismuth cooled SMR in Korea, URANUS with 40∼70 MWe is examined in detail. This paper will cover a framework of the program and a strategy for the successful deployment of small modular reactor how the goals would entail and the approach to collaboration with other entities.

Modular Stellarator Fusion Reactor concept

International Nuclear Information System (INIS)

Miller, R.L.; Krakowski, R.A.

1981-08-01

A preliminary conceptual study is made of the Modular Stellarator Reactor (MSR). A steady-state ignited, DT-fueled, magnetic fusion reactor is proposed for use as a central electric-power station. The MSR concept combines the physics of the classic stellarator confinement topology with an innovative, modular-coil design. Parametric tradeoff calculations are described, leading to the selection of an interim design point for a 4-GWt plant based on Alcator transport scaling and an average beta value of 0.04 in an l = 2 system with a plasma aspect ratio of 11. The physics basis of the design point is described together with supporting magnetics, coil-force, and stress computations. The approach and results presented herein will be modified in the course of ongoing work to form a firmer basis for a detailed conceptual design of the MSR

The modular high temperature gas cooled reactor

International Nuclear Information System (INIS)

Lutz, D.E.; Lipps, A.J.

1984-01-01

Due to relatively high operating temperatures, the gas-cooled reactor has the potential to serve a wide variety of energy applications. This paper discusses the energy applications which can be served by the modular HTGR, the magnitude of the potential markets, and the HTGR product cost incentives relative to fossil fuel competition. Advantages of the HTGR modular systems are presented along with a description of the design features and performance characteristics of the current reference HTGR modular systems

Advanced In-Core Fuel Cycles for the Gas Turbine-Modular Helium Reactor

Energy Technology Data Exchange (ETDEWEB)

Talamo, Alberto

2006-04-15

Amid generation IV of nuclear power plants, the Gas Turbine - Modular Helium Reactor, designed by General Atomics, is the only core with an energy conversion efficiency of 50%; the safety aspects, coupled to construction and operation costs lower than ordinary Light Water Reactors, renders the Gas Turbine - Modular Helium reactor rather unequaled. In the present studies we investigated the possibility to operate the GT-MHR with two types of fuels: LWRs waste and thorium; since thorium is made of only fertile {sup 232}Th, we tried to mix it with pure {sup 233}U, {sup 235}U or {sup 239}Pu; ex post facto, only uranium isotopes allow the reactor operation, that induced us to examine the possibility to use a mixture of uranium, enriched 20% in {sup 235}U, and thorium. We performed all calculations by the MCNP and MCB codes, which allowed to model the reactor in a very detailed three-dimensional geometry and to describe the nuclides transmutation in a continuous energy approach; finally, we completed our studies by verifying the influence of the major nuclear data libraries, JEFF, JENDL and ENDF/B, on the obtained results.

The Modular Helium Reactor for Hydrogen Production

International Nuclear Information System (INIS)

E. Harvego; M. Richards; A. Shenoy; K. Schultz; L. Brown; M. Fukuie

2006-01-01

For electricity and hydrogen production, an advanced reactor technology receiving considerable international interest is a modular, passively-safe version of the high-temperature, gas-cooled reactor (HTGR), known in the U.S. as the Modular Helium Reactor (MHR), which operates at a power level of 600 MW(t). For hydrogen production, the concept is referred to as the H2-MHR. Two concepts that make direct use of the MHR high-temperature process heat are being investigated in order to improve the efficiency and economics of hydrogen production. The first concept involves coupling the MHR to the Sulfur-Iodine (SI) thermochemical water splitting process and is referred to as the SI-Based H2-MHR. The second concept involves coupling the MHR to high-temperature electrolysis (HTE) and is referred to as the HTE-Based H2-MHR

Proliferation resistance considerations for remote small modular reactors

Energy Technology Data Exchange (ETDEWEB)

Whitlock, J., E-mail: whitlockj@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Sprinkle, J., E-mail: j.sprinkle@iaea.org [International Atomic Energy Agency, Vienna (Austria)

2013-07-01

Remotely located Small Modular Reactors at the low end of energy production (on the order of 10 MWe, referenced here as Very Small Modular Reactors or VSMRs) present unique proliferation resistance advantages and challenges. Addressing these challenges in the most efficient manner may not only be desirable, but necessary, for development of this technology. Incorporation of safeguards considerations early in the design process (Safeguards by Design) along with safety, security, economics and other key drivers, is of importance. Operational Transparency may become an essential aspect of the safeguards approach for such systems. (author)

Initiating Events for Multi-Reactor Plant Sites

Energy Technology Data Exchange (ETDEWEB)

Muhlheim, Michael David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Poore, III, Willis P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2014-09-01

Inherent in the design of modular reactors is the increased likelihood of events that initiate at a single reactor affecting another reactor. Because of the increased level of interactions between reactors, it is apparent that the Probabilistic Risk Assessments (PRAs) for modular reactor designs need to specifically address the increased interactions and dependencies.

Small modular reactors are 'crucial technology'

Science.gov (United States)

Johnston, Hamish

2018-03-01

Small modular nuclear reactors (SMRs) offer a way for the UK to reduce carbon dioxide emissions from electricity generation, while allowing the country to meet the expected increase in demand for electricity from electric vehicles and other uses.

Design study on sodium-cooled middle-scale modular reactor

International Nuclear Information System (INIS)

Kisohara, Naoyuki; Hishida, Masahiko; Nibe, Nobuaki

2003-09-01

In Phase 1 of the 'Feasibility Study on Commercialized Fast Reactor Cycle Systems (F/S)', an advanced loop type reactor has been selected as a promising concept of sodium-cooled middle-scale modular reactor, which has a possibility to fulfill the design requirements of the F/S. This report summarizes the results of the design study on the sodium-cooled middle-scale modular reactor performed in JFY2002, which is the second year of Phase 2. The construction cost of the sodium-cooled middle-scale modular reactor, which has been constructed in JFY2002, was almost achieved the economical goal. But its achievability was not sufficient to accept the concept. In order to reduce the construction cost, the plant concept has been re-constructed based on the 50 MWe plant studied in JFY2002. After that, fundamental specifications of main systems and components for the new concept have been set, and critical subjects have been examined and evaluated. In addition, in order to achieve the further cost reduction, the plant with simplified secondary system, the plant with electric magnetic pump in secondary system, and the fuel handling system are examined and evaluated. As a result of this study, the plant concept of the sodium-cooled middle-scale modular reactor has been constructed, which has a prospect to satisfy the economic goal (construction cost: less than 200,000 yens/kWe, etc.) and has a prospect to solve the critical subjects. From now on, reflecting the results of elemental experiments, the preliminary conceptual design of this plant will be preceded toward the selection for narrowing down candidate concepts at the end of Phase 2. (author)

Preliminary Study for Conceptual Design of Advanced Long Life Small Modular Fast Reactor

Energy Technology Data Exchange (ETDEWEB)

Tak, Taewoo; Choe, Jiwon; Jeong, Yongjin; Lee, Deokjung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Kim, T. K. [Argonne National Laboratory, Argonne (United States)

2015-05-15

As one of the non-water coolant Small-Modular Reactor (SMR) core concepts for use in the mid- to long-term, ANL has proposed a 100 MWe Advanced sodium-cooled Fast Reactor core concept (AFR-100) targeting a small grid, transportable from pre-licensed factories to the remote plant site for affordable supply. Various breed-and-burn core concepts have been proposed to extend the reactor cycle length, which includes CANDLE with a cigar-type depletion strategy, TerraPower reactors with fuel shuffling for effective breeding, et al. UNIST has also proposed an ultra-long cycle fast reactor (UCFR) core concept having the power rating of 1000 MWe. By adopting the breed-and-burn strategies, the UCFR core can maintain criticality for a targeting reactor lifetime of 60 years without refueling. The objective of this project is to develop an advanced long-life SMR core concept by adopting both the small modular design features of the AFR-100 and the long-life breed-and-burn concept of the UCFR. A conceptual design of long life small modular fast reactor is under development by adopting both the small modular design features of the AFR-100 and the long-life breed-and-burn concept of the UCFR. The feasibility of the long-life fast reactor concepts was reviewed to obtain the core design guidelines and the reactor design requirements of long life small modular fast reactor were proposed in this study.

Modular reactor head shielding system

International Nuclear Information System (INIS)

Jacobson, E. B.

1985-01-01

An improved modular reactor head shielding system is provided that includes a frame which is removably assembled on a reactor head such that no structural or mechanical alteration of the head is required. The shielding system also includes hanging assemblies to mount flexible shielding pads on trolleys which can be moved along the frame. The assemblies allow individual pivoting movement of the pads. The pivoting movement along with the movement allowed by the trolleys provides ease of access to any point on the reactor head. The assemblies also facilitate safe and efficient mounting of the pads directly to and from storage containers such that workers have additional shielding throughout virtually the entire installation and removal process. The flexible shielding pads are designed to interleave with one another when assembled around the reactor head for substantially improved containment of radiation leakage

Westinghouse Small Modular Reactor (SMR) Programe

International Nuclear Information System (INIS)

Shulyak, Nick

2014-01-01

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

Economical and engineering aspects of modular-type fast reactors

International Nuclear Information System (INIS)

Kirillov, E.V.; Demidova, L.S.

1989-01-01

Economical and engineering characteristics for SAFR and PRISM modular-type reactors are analyzed on the basis of foreign papers. Dependence of economical characteristics for SAFR modules on their output is shown. Cost of power generation for the NPPs with PRISM reactor, LWR reactor and for coal thermal power plant is presented

DCNS invents the immersed civil modular reactor

International Nuclear Information System (INIS)

Guilhem, Jean

2013-01-01

SMRs (Small and Modular Reactors) are a response to networks smaller than 10 GW. They are under study in USA, Russia and China, and a Korean one is already certified. In France, DCNS proposes Flexblue, a compact modular 160 MWe reactor, completely immersed 60 to 100 m under the sea level, few kilometres off the coasts. Almost invisible and protected by the marine environment, it will benefit from an infinite cold source which will ensure a high safety level. It is designed to produce 1 MWh at less than 100 euros. It will be retrieved onshore for its dismantling at the end of its service life. Its operation is said to be neutral for surrounding ecosystems with a tritium release more than 90 per cent less than that of onshore power plants

Development of small simplified modular reactors

International Nuclear Information System (INIS)

Hiki, Hideaki; Nakamaru, Mikihide

2003-01-01

The small simplified modular reactor, which is being development with matured BWR technologies together with innovative systems/components, will provide attractiveness for the energy market in the world due to its flexibility in energy demands as well as in site conditions, its high potential in reducing investment risk and its safety feature facilitating public acceptance. The flexibility is achieved by small output of 300 MWe and capability of long operating cycle (refueling intervals). The investment potential is expected from simplification/innovation in design such as natural circulation core cooling with the bottom located short core, internal upper entry control rod drives (CRDs) with ring-type dryers and hull structure building concept. The natural circulation core eliminates recirculation pumps as well as needs for maintenance of such pumps. The internal CRDs shorten the height of the reactor vessel (RPV) and consequently shorten the primary containment vessel (PCV). The hull structure facilitates modular arrangement, design standardization and factory fabrication. The safety feature mainly consists of large water inventory above the core without large penetration below the top of the core, passive containment cooling system (PCCS), passive auto-catalytic recombiner and in-vessel retention (IVR) capability. The large inventory increases the system response time in case of design base accidents including loss of coolant accidents. The PCCS suppresses PCV pressure by steam condensation without and AC power. The recombiner decreases hydrogen concentration in the PCV in case of a severe accident. The IVR could cool the molten core inside the RPV if the core should be damaged by loss of core coolability. These innovative systems/components featured in the small simplified modular reactor will stimulate global energy market. (author)

NEPTUNE: a modular system for light-water reactor calculation

International Nuclear Information System (INIS)

Bouchard, J.; Kanevoky, A.; Reuss, P.

1975-01-01

A complete modular system of light water reactor calculations has been designed. It includes basic nuclear data processing, the APOLLO phase: transport calculations for cells, multicells, fuel assemblies or reactors, the NEPTUNE phase: reactor calculations. A fuel management module, devoted to the automatic determination of the best shuffling strategy is included in NEPTUNE [fr

Modular coils: a promising toroidal-reactor-coil system

International Nuclear Information System (INIS)

Chu, T.K.; Furth, H.P.; Johnson, J.L.; Ludescher, C.; Weimer, K.E.

1981-04-01

The concept of modular coils originated from a need to find reactor-relevant stellarator windings, but its usefulness can be extended to provide an externally applied, additional rotational transform in tokamaks. Considerations of (1) basic principles of modular coils, (2) types of coils, (3) types of configurations (general, helically symmetric, helically asymmetric, with magnetic well, with magnetic hill), (4) types of rotational transform profile, and (5) structure and origin of ripples are given. These results show that modular coils can offer a wide range of vacuum magnetic field configurations, some of which cannot be obtained with the classical stellarator or torsatron coil configuration

Advanced Mesh-Enabled Monte carlo capability for Multi-Physics Reactor Analysis

Energy Technology Data Exchange (ETDEWEB)

Wilson, Paul; Evans, Thomas; Tautges, Tim

2012-12-24

This project will accumulate high-precision fluxes throughout reactor geometry on a non- orthogonal grid of cells to support multi-physics coupling, in order to more accurately calculate parameters such as reactivity coefficients and to generate multi-group cross sections. This work will be based upon recent developments to incorporate advanced geometry and mesh capability in a modular Monte Carlo toolkit with computational science technology that is in use in related reactor simulation software development. Coupling this capability with production-scale Monte Carlo radiation transport codes can provide advanced and extensible test-beds for these developments. Continuous energy Monte Carlo methods are generally considered to be the most accurate computational tool for simulating radiation transport in complex geometries, particularly neutron transport in reactors. Nevertheless, there are several limitations for their use in reactor analysis. Most significantly, there is a trade-off between the fidelity of results in phase space, statistical accuracy, and the amount of computer time required for simulation. Consequently, to achieve an acceptable level of statistical convergence in high-fidelity results required for modern coupled multi-physics analysis, the required computer time makes Monte Carlo methods prohibitive for design iterations and detailed whole-core analysis. More subtly, the statistical uncertainty is typically not uniform throughout the domain, and the simulation quality is limited by the regions with the largest statistical uncertainty. In addition, the formulation of neutron scattering laws in continuous energy Monte Carlo methods makes it difficult to calculate adjoint neutron fluxes required to properly determine important reactivity parameters. Finally, most Monte Carlo codes available for reactor analysis have relied on orthogonal hexahedral grids for tallies that do not conform to the geometric boundaries and are thus generally not well

Advanced modularity design for the MIT pebble bed reactor

Energy Technology Data Exchange (ETDEWEB)

Kadak, Andrew C. [Department of Nuclear Engineering, Massachusetts Institute of Technology, 77 Massachusetts Avenue, 24-202 Cambridge, MA 02139-4307 (United States)]. E-mail: kadak@mit.edu; Berte, Marc V. [Department of Nuclear Engineering, Massachusetts Institute of Technology, 77 Massachusetts Avenue, 24-202 Cambridge, MA 02139-4307 (United States)]. E-mail: mvberte@yahoo.com

2006-03-15

The future of all reactors will depend on whether they can be economically built and operated. One of the major impediments to new nuclear construction is the capital cost due in large part to the length of construction time and complexity of the plant. Pebble bed reactors offer the opportunity to reduce the complexity of the plant because the number of safety systems required is significantly reduced due to the inherent safety of the technology. However, because of its small size, the capital cost per kilowatt is likely to be large if traditional construction approaches are followed. This strongly suggests the need for innovative construction concepts to reduce the construction time and cost. MIT has proposed a modularity approach in which the plant is pre-built in space-frame type modules which are built in factories. These space frames would contain all the equipment contained in a given volume. Once equipment in the space frame is installed, the space frame would then be shipped to the site and assembled 'lego-style.' Studies presently underway have demonstrated the feasibility of the concept. Thermal stress analysis has been performed and an integrated design with the space frames has been developed. It is expected that this modularity approach will significantly shorten construction time and expense. This paper proposes a concept for further development, not a final design for the entire plant.

Advanced modularity design for the MIT pebble bed reactor

International Nuclear Information System (INIS)

Kadak, Andrew C.; Berte, Marc V.

2006-01-01

The future of all reactors will depend on whether they can be economically built and operated. One of the major impediments to new nuclear construction is the capital cost due in large part to the length of construction time and complexity of the plant. Pebble bed reactors offer the opportunity to reduce the complexity of the plant because the number of safety systems required is significantly reduced due to the inherent safety of the technology. However, because of its small size, the capital cost per kilowatt is likely to be large if traditional construction approaches are followed. This strongly suggests the need for innovative construction concepts to reduce the construction time and cost. MIT has proposed a modularity approach in which the plant is pre-built in space-frame type modules which are built in factories. These space frames would contain all the equipment contained in a given volume. Once equipment in the space frame is installed, the space frame would then be shipped to the site and assembled 'lego-style.' Studies presently underway have demonstrated the feasibility of the concept. Thermal stress analysis has been performed and an integrated design with the space frames has been developed. It is expected that this modularity approach will significantly shorten construction time and expense. This paper proposes a concept for further development, not a final design for the entire plant

Overview of the Westinghouse Small Modular Reactor building layout

Energy Technology Data Exchange (ETDEWEB)

Cronje, J. M. [Westinghouse Electric Company LLC, Centurion (South Africa); Van Wyk, J. J.; Memmott, M. J. [Westinghouse Electric Company LLC, Cranberry Township, PA (United States)

2012-07-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the third in a series of four papers, which describe the design and functionality of the Westinghouse SMR. It focuses in particular upon the plant building layout and modular design of the Westinghouse SMR. In the development of small modular reactors, the building layout is an area where the safety of the plant can be improved by applying new design approaches. This paper will present an overview of the Westinghouse SMR building layout and indicate how the design features improve the safety and robustness of the plant. The Westinghouse SMR is designed with no shared systems between individual reactor units. The main buildings inside the security fence are the nuclear island, the rad-waste building, the annex building, and the turbine building. All safety related equipment is located in the nuclear island, which is a seismic class 1 building. To further enhance the safety and robustness of the design, the reactor, containment, and most of the safety related equipment are located below grade on the nuclear island. This reduces the possibility of severe damage from external threats or natural disasters. Two safety related ultimate heat sink (UHS) water tanks that are used for decay heat removal are located above grade, but are redundant and physically separated as far as possible for improved safety. The reactor and containment vessel are located below grade in the center of the nuclear island. The rad-waste and other radioactive systems are located on the bottom floors to limit the radiation exposure to personnel. The Westinghouse SMR safety trains are completely separated into four unconnected quadrants of the building, with access between quadrants only allowed

Emergency systems and protection equipment of modular steam generators for fast reactors

International Nuclear Information System (INIS)

Matal, O.

The requirements are discussed for accident protection of modular steam generators for fast reactors. Accident protection is assessed for a modular through-flow steam generator and for a natural circulation modular steam generator. Benefits and constraints are shown and possible improvements are outlined for accident protection of liquid sodium fired modular steam generators. (Kr)

Evaluation of the Gas Turbine Modular Helium Reactor

Energy Technology Data Exchange (ETDEWEB)

1994-02-01

Recent advances in gas-turbine and heat exchanger technology have enhanced the potential for a Modular Helium Reactor (MHR) incorporating a direct gas turbine (Brayton) cycle for power conversion. The resulting Gas Turbine Modular Helium Reactor (GT-MHR) power plant combines the high temperature capabilities of the MHR with the efficiency and reliability of modern gas turbines. While the passive safety features of the steam cycle MHR (SC-MHR) are retained, generation efficiencies are projected to be in the range of 48% and steam power conversion systems, with their attendant complexities, are eliminated. Power costs are projected to be reduced by about 20%, relative to the SC-MHR or coal. This report documents the second, and final, phase of a two-part evaluation that concluded with a unanimous recommendation that the direct cycle (DC) variant of the GT-MHR be established as the commercial objective of the US Gas-Cooled Reactor Program. This recommendation has been endorsed by industrial and utility participants and accepted by the US Department of Energy (DOE). The Phase II effort, documented herein, concluded that the DC GT-MHR offers substantial technical and economic advantages over both the IDC and SC systems. Both the DC and IDC were found to offer safety advantages, relative to the SC, due to elimination of the potential for water ingress during power operations. This is the dominant consequence event for the SC. The IDC was judged to require somewhat less development than the direct cycle, while the SC, which has the greatest technology base, incurs the least development cost and risk. While the technical and licensing requirements for the DC were more demanding, they were judged to be incremental and feasible. Moreover, the DC offers significant performance and cost improvements over the other two concepts. Overall, the latter were found to justify the additional development needs.

Evaluation of the Gas Turbine Modular Helium Reactor

International Nuclear Information System (INIS)

1994-02-01

Recent advances in gas-turbine and heat exchanger technology have enhanced the potential for a Modular Helium Reactor (MHR) incorporating a direct gas turbine (Brayton) cycle for power conversion. The resulting Gas Turbine Modular Helium Reactor (GT-MHR) power plant combines the high temperature capabilities of the MHR with the efficiency and reliability of modern gas turbines. While the passive safety features of the steam cycle MHR (SC-MHR) are retained, generation efficiencies are projected to be in the range of 48% and steam power conversion systems, with their attendant complexities, are eliminated. Power costs are projected to be reduced by about 20%, relative to the SC-MHR or coal. This report documents the second, and final, phase of a two-part evaluation that concluded with a unanimous recommendation that the direct cycle (DC) variant of the GT-MHR be established as the commercial objective of the US Gas-Cooled Reactor Program. This recommendation has been endorsed by industrial and utility participants and accepted by the US Department of Energy (DOE). The Phase II effort, documented herein, concluded that the DC GT-MHR offers substantial technical and economic advantages over both the IDC and SC systems. Both the DC and IDC were found to offer safety advantages, relative to the SC, due to elimination of the potential for water ingress during power operations. This is the dominant consequence event for the SC. The IDC was judged to require somewhat less development than the direct cycle, while the SC, which has the greatest technology base, incurs the least development cost and risk. While the technical and licensing requirements for the DC were more demanding, they were judged to be incremental and feasible. Moreover, the DC offers significant performance and cost improvements over the other two concepts. Overall, the latter were found to justify the additional development needs

An evaluation of reactor cooling and coupled hydrogen production processes using the modular helium reactor

International Nuclear Information System (INIS)

Harvego, E.A.; Reza, S.M.M.; Richards, M.; Shenoy, A.

2006-01-01

The high-temperature characteristics of the modular helium reactor (MHR) make it a strong candidate for producing hydrogen using either thermochemical or high-temperature electrolysis (HTE) processes. Using heat from the MHR to drive a sulfur-iodine (SI) thermochemical hydrogen production process has been the subject of a U.S. Department of Energy sponsored Nuclear Engineering Research Initiative (NERI) project led by General Atomics, with participation from the Idaho National Laboratory (INL) and Texas A and M University. While the focus of much of the initial work was on the SI thermochemical production of hydrogen, recent activities included development of a preconceptual design for an integral HTE hydrogen production plant driven by the process heat and electricity produced by a 600 MW MHR. This paper describes ATHENA analyses performed to evaluate alternative primary system cooling configurations for the MHR to minimize peak reactor vessel and core temperatures while achieving core helium outlet temperatures in the range of 900-1000 deg. C that are needed for the efficient production of hydrogen using either the SI or HTE process. The cooling schemes investigated are intended to ensure peak fuel temperatures do not exceed specified limits under normal or transient upset conditions, and that reactor vessel temperatures do not exceed American Society of Mechanical Engineers (ASME) code limits for steady-state or transient conditions using standard light water reactor vessel materials. Preconceptual designs for SI and HTE hydrogen production plants driven by one or more 600 MW MHRs at helium outlet temperatures in the range of 900-1000 deg. C are described and compared. An initial SAPHIRE model to evaluate the reliability, maintainability, and availability of the SI hydrogen production plant is also described. Finally, a preliminary flowsheet for a conceptual design of an HTE hydrogen production plant coupled to a 600 MW modular helium reactor is presented and

Introducing small modular reactors into Canada

International Nuclear Information System (INIS)

Humphries, J.R.

2012-01-01

In recent years there has been a growing interest in smaller, simpler reactors for generating electricity and process heat. This is evidenced in the growing body of literature and the increasingly frequent meetings and conferences on the subject. The interest in Small Modular Reactors (SMRs) is driven to a large extent by the desire to reduce capital costs, to reduce greenhouse gas emissions, to replace retiring fossil plants that do not meet today's environmental standards, and to provide power in locations away from large electrical grids. These drivers are as important in Canada as they are in the U.S., where the design and licensing of SMRs is being most vigorously pursued. They have led to a growing interest in Canada as a potentially significant market for SMRs, particularly in the Western Provinces of Alberta and Saskatchewan and in the remote First Nations communities of Northern Canada. There is a growing body of literature addressing the regulation and licensing of Small Modular Reactors in the U.S. Issues being identified in there can generally be categorized as licensing framework issues, licensing application issues, and design and manufacturing issues. Many of these issues are embedded in the US regulatory framework and can only be resolved through changes in the regulations. For the most part these issues are equally applicable in Canada and will need to be addressed in introducing SMRs here. A significant difference, however, is that these issues can be addressed within the Canadian regulatory framework without requiring changes in the regulations. The CNSC has taken a very proactive stance regarding the licensing of small reactors in Canada. They have published two new Regulatory Documents stipulating the requirements for licensing small reactors. A key feature is that they allow the application of a 'graded approach' in which the stringency of the design measures and analyses applied are commensurate with the level of risk posed by

Modular Stellarator Reactor conceptual design study

International Nuclear Information System (INIS)

Miller, R.L.; Bathke, C.G.

1983-01-01

A conceptual design study of the Modular Stellarator Reactor is summarized. The physics basis of the approach is elucidated with emphasis on magnetics performance optimization. Key engineering features of the fusion power core are described. Comparisons with an analogous continuous-helical-coil (torsatron) system are made as the basis of a technical and economic assessment

Modular stellarator reactor conceptual design study

International Nuclear Information System (INIS)

Miller, R.L.; Krakowski, R.A.; Bathke, C.G.

1983-01-01

A conceptual design study of the Modular Stellarator Reactor is summarized. The physics basis of the approach is elucidated with emphasis on magnetics performance optimization. Key engineering features of the fusion power core are described. Comparisons with an analogous continuous-helical-coil (torsatron) system are made as the basis of a technical and economic assessment

Small modular reactors: Simpler, safer, cheaper?

International Nuclear Information System (INIS)

Vujić, Jasmina; Bergmann, Ryan M.; Škoda, Radek; Miletić, Marija

2012-01-01

Nuclear energy can play a very significant long-term role for meeting the world’s increasing energy demands, while simultaneously addressing challenges associated with global climate and environmental impact. Many nations of the world, particularly the Asia/Pacific Rim countries, are actively engaged in a major expansion of their nuclear energy complex. The degree to which nuclear energy can address long-term energy needs, either globally or regionally, will be dictated by the pace and adequacy of technical and policy solutions for waste, safety, security, and non-proliferation issues, as well as the capital cost of construction. Small Modular Reactors (SMRs) could successfully address several of these issues. SMRs offer simpler, standardized, and safer modular design by being factory built, requiring smaller initial capital investment, and having shorter construction times. The SMRs could be small enough to be transportable, could be used in isolated locations without advanced infrastructure and without power grid, or could be clustered in a single site to provide a multi-module, large capacity power plant. This paper summarizes some of the basic features of SMRs for early deployment, several advanced SMR concepts, and points out the benefits and challenges in regulatory, economical, safety and security issues. -- Highlights: ► We held a summer forum on SMR technologies at UC Berkeley in July 2010. ► Advantages and disadvantages, technical and economic, of each design were discussed. ► Further literature searches were also done and this paper summarizes prominent designs. ► We conclude SMRs have no large problems preventing their introduction into the nuclear market.

Progress in modular-stellarator fusion-power-reactor conceptual designs

International Nuclear Information System (INIS)

Sviatoslavsky, I.N.; Van Sciver, S.W.; Kulcinski, G.L.

1982-01-01

Recent encouraging experimental results on stellarators/torsatrons/heliotrons (S/T/H) have revived interest in these concepts as possible fusion power reactors. The use of modular coils to generate the stellarator topology has added impetus to this renewed interest. Studies of the modular coil approach to stellarators by UW-Madison and Los Alamos National Laboratory are summarized in this paper

MHTGR [Modular High-Temperature Gas-Cooled Reactor] technology development plan

International Nuclear Information System (INIS)

Homan, F.J.; Neylan, A.J.

1988-01-01

This paper presents the approach used to define the technology program needed to support design and licensing of a Modular High-Temperature Gas-Cooled Reactor (MHTGR). The MHTGR design depends heavily on data and information developed during the past 25 years to support large HTGR (LHTGR) designs. The technology program focuses on MHTGR-specific operating and accident conditions, and on validation of models and assumptions developed using LHTGR data. The technology program is briefly outlined, and a schedule is presented for completion of technology work which is consistent with completion of a Final Safety Summary Analysis Report (FSSAR) by 1992

Pre-Conceptual Design of a Fluoride-Salt-Cooled Small Modular Advanced High Temperature Reactor (SmAHTR)

Energy Technology Data Exchange (ETDEWEB)

Greene, Sherrell R [ORNL; Gehin, Jess C [ORNL; Holcomb, David Eugene [ORNL; Carbajo, Juan J [ORNL; Ilas, Dan [ORNL; Cisneros, Anselmo T [ORNL; Varma, Venugopal Koikal [ORNL; Corwin, William R [ORNL; Wilson, Dane F [ORNL; Yoder Jr, Graydon L [ORNL; Qualls, A L [ORNL; Peretz, Fred J [ORNL; Flanagan, George F [ORNL; Clayton, Dwight A [ORNL; Bradley, Eric Craig [ORNL; Bell, Gary L [ORNL; Hunn, John D [ORNL; Pappano, Peter J [ORNL; Cetiner, Sacit M [ORNL

2011-02-01

This document presents the results of a study conducted at Oak Ridge National Laboratory during 2010 to explore the feasibility of small modular fluoride salt-cooled high temperature reactors (FHRs). A preliminary reactor system concept, SmATHR (for Small modular Advanced High Temperature Reactor) is described, along with an integrated high-temperature thermal energy storage or salt vault system. The SmAHTR is a 125 MWt, integral primary, liquid salt cooled, coated particle-graphite fueled, low-pressure system operating at 700 C. The system employs passive decay heat removal and two-out-of-three , 50% capacity, subsystem redundancy for critical functions. The reactor vessel is sufficiently small to be transportable on standard commercial tractor-trailer transport vehicles. Initial transient analyses indicated the transition from normal reactor operations to passive decay heat removal is accomplished in a manner that preserves robust safety margins at all times during the transient. Numerous trade studies and trade-space considerations are discussed, along with the resultant initial system concept. The current concept is not optimized. Work remains to more completely define the overall system with particular emphasis on refining the final fuel/core configuration, salt vault configuration, and integrated system dynamics and safety behavior.

Modular Stellarator Fusion Reactor (MSR) concept

International Nuclear Information System (INIS)

Miller, R.L.; Krakowski, R.A.

1981-01-01

A preliminary conceptual study has been made of the Modulator Stellarator Reactor (MSR) as a stedy-state, ignited, DT-fueled, magnetic fusion reactor. The MSR concept combines the physics of classic stellarator confinement with an innovative, modular-coil design. Parametric tradeoff calculations are described, leading to the selection of an interim design point for a 4.8-GWt plant based on Alcator transport scaling and an average beta value of 0.04 in an l = 2 system with a plasma aspect ratio of 11. Neither an economic analysis nor a detailed conceptual engineering design is presented here, as the primary intent of this scoping study is the elucidation of key physics tradeoffs, constraints, and uncertainties for the ultimate power-reactor embodiment

The ESKOM pebble bed modular reactor

International Nuclear Information System (INIS)

Gittus, J.H.

1999-01-01

An audit has been made of the design, construction, safety, economics and marketability of the ESKOM pebble bed modular reactor (PBMR). In this paper that audit is briefly summarized. The principal conclusions of the audit are as follows. The design is sound. It is a logical development of the designs proposed for other, modern, high-temperature gas-cooled reactors. More than 80% of the cost of constructing and commissioning a series of PBMRs would be spent in South Africa. The PBMR is much safer than existing nuclear power reactors and for many practical purposes it may be treated as a conventional chemical plant. The PBMR is economically competitive with thermal power stations. There is a substantial global market for the PBMR. (author)

Study on the safety and on international developments of small modular reactors (SMR). Final report; Studie zur Sicherheit und zu internationalen Entwicklungen von Small Modular Reactors (SMR). Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Buchholz, Sebastian; Kruessenberg, Anne; Schaffrath, Andreas; Zipper, Reinhard

2015-05-15

This report documents the work and results of the project RS1521 Study of Safety and International Development of Small Modular Reactors (SMR). The aims of this study can be summarized as - setting-up of a sound overview on SMR, - identification of essential issues of reactor safety research and future R and D projects, - identification of needs for adaption of system codes of GRS used in reactor safety research. The sound overview consists of the descriptions of in total 69 SMR (Small and Medium Sized Rector) concepts (32 light water reactors (LWR), 22 liquid metal cooled reactors (LMR), 2 heavy water reactors, 9 gas cooled reactors (GCR) and 4 molten salt reactors (MSR)). It provides information about the core, the cooling circuits and the safety systems. The quality of the given specifications depends on their availability and public accessibility. Using the safety requirements for nuclear power plants and the fundamental safety functions, the safety relevant issues of the described SMR concepts were identified. The systems and measures used in the safety requirements were summarized in table form. Finally it was evaluated whether these systems and measures can be already simulated with the nuclear simulation chain of GRS and where further code development and validation is necessary. The results of this study can be summarized as follows: Many of the current SMR concepts are based on integral design. Here the main components like steam generators, intermediate heat exchangers or - in case of forced convection core cooling - main cooling pumps are located within the reactor pressure vessel. Most of the SMR fulfil highest safety standards and their safety concepts are mainly based on passive safety systems. The safety of theses reactors is achieved indefinitely without energy supply or additional measures of the operators. Since SMR's aim is not only to produce electricity but also couple them with chemical or physical process plants, the safety aspects of

The Gas Turbine - Modular Helium Reactor: A Promising Option for Near Term Deployment

International Nuclear Information System (INIS)

LaBar, Malcolm P.

2002-01-01

The Gas Turbine - Modular Helium Reactor (GT-MHR) is an advanced nuclear power system that offers unparalleled safety, high thermal efficiency, environmental advantages, and competitive electricity generation costs. The GT-MHR module couples a gas-cooled modular helium reactor (MHR) with a high efficiency modular Brayton cycle gas turbine (GT) energy conversion system. The reactor and power conversion systems are located in a below grade concrete silo that provides protection against sabotage. The GT-MHR safety is achieved through a combination of inherent safety characteristics and design selections that take maximum advantage of the gas-cooled reactor coated particle fuel, helium coolant and graphite moderator. The GT-MHR is projected to be economically competitive with alternative electricity generation technologies due to the high operating temperature of the gas-cooled reactor, high thermal efficiency of the Brayton cycle power conversion system, high fuel burnup (>100,000 MWd/MT), and low operation and maintenance requirements. (author)

System Dynamics Modeling of interactive cost factors for small modular reactors

International Nuclear Information System (INIS)

Ahn, Nam Sung; Lee, Keun Dae; Yoon, Suk Ho

2011-01-01

As a part of the Study on Economic Efficiency and Marketability of small modular reactors project, we at Nemo partners NEC consulting corporation were studying the various cost factors on small modular reactors (SMRs). To have a better knowledge of the interaction between the cost factors, System Dynamics Modeling has been developed. This model will contribute to our understanding of the interaction on the major factors effecting on the unit cost of SMRs to the SMRs' market share in the market economics as competition

Wishful thinking and real problems: Small modular reactors, planning constraints, and nuclear power in Jordan

International Nuclear Information System (INIS)

Ramana, M.V.; Ahmad, Ali

2016-01-01

Jordan plans to import two conventional gigawatt scale nuclear reactors from Russia that are expensive and too large for Jordan's current electricity grid. Jordan efforts to establish nuclear power might become easier in some ways if the country were to construct Small Modular Reactors, which might be better suited to Jordan's financial capabilities and its smaller electrical grid capacity. But, the SMR option raises new problems, including locating sites for multiple reactors, finding water to cool these reactors, and the higher cost of electricity generation. Jordan's decision has important implications for its energy planning as well as for the market for SMRs. - Highlights: •Jordan is planning to purchase two large reactors from Russia. •Large reactors would be inappropriate to Jordan's small electricity grid. •Small modular reactors would be more appropriate to Jordan's grid, but have problems. •The market for small modular reactors will be smaller than often projected. •Jordan should consider the financial impact of building a large nuclear reactor.

Model and control of the isolated multi-modular converter

DEFF Research Database (Denmark)

Verdugo, Cristian; Candela, Jose Ignacio; Blaabjerg, Frede

2017-01-01

Multi-modular converters have proven to be one of the most suitable topology to be used in high and power applications. The modularity of these configurations offer several advantages, such as: high quality voltage, redundancy and high efficiency. In these converters, the series connection...... voltages. However, increasing the number of conversion stages, increase also the losses and the implementation costs. In order to overcome these drawbacks, a new configuration of multi-modular converter is proposed in this paper. The proposed converter provides isolation with one conversion stage by means...... of using low frequency transformers in each module. The main principle behind this concept, the control structure and simulation results are presented to validate the proposed configuration....

Calculation and modular properties of multi-loop superstring amplitudes

International Nuclear Information System (INIS)

Danilov, G S

2012-01-01

Multi-loop superstring amplitude is calculated in the conventional gauge where Grassmann moduli are carried by the 2D gravitino field. Generally, instead of the modular symmetry, the amplitudes hold the symmetry under modular transformations added by relevant transformations of the 2D local supersymmetry. If a number of loops are larger than 3, the integration measures are not modular forms. In this case the expression for the amplitude contains an integral over the bound of the fundamental region of the modular group. (paper)

High-temperature reactor in modular construction

International Nuclear Information System (INIS)

Mueller, F.U.; Reutler, H.; Ullrich, M.

1981-01-01

Together with other reactors of the same type a gas-cooled, small-sized high-temperature reactor is to be assembled into a plant with modular design. The reactor vessel can be withdrawn as a whole after shutdown, removal of the fuel element charge, disassembly of the control rods, and opening of the closure of the safety containment. All apertures for the inlet and outlet of the cooling gas are located in the ground plate of the reactor. The lower part of the reactor cavern serves as inlet space for the cool gas, while the heated gas is let in through a line of a heat sink, e.g. a heat exchanger. The ground plate is connected with the hot gas line or with an inserted hot gas collecting room by means of a simple plug connection which is released automatically when the reactor vessel is withdrawn. The cooling gas, which is put into circulation by a blower and led through special conducting systems, is also used for cooling the outer metal jacket of the hot gas line. A second design is described according to which the reactor and heat exchanger are superposed in a safety containment, such as applied for pressurized water-cooled nuclear reactors. (orig.) [de

Preliminary Design of S-CO2 Brayton Cycle for KAIST Micro Modular Reactor

International Nuclear Information System (INIS)

Kim, Seong Gu; Kim, Min Gil; Bae, Seong Jun; Lee, Jeong Ik

2013-01-01

This paper suggests a complete modular reactor with an innovative concept of reactor cooling by using a supercritical carbon dioxide directly. Authors propose the supercritical CO 2 Brayton cycle (S-CO 2 cycle) as a power conversion system to achieve small volume of power conversion unit (PCU) and to contain the core and PCU in one vessel for the full modularization. This study suggests a conceptual design of small modular reactor including PCU which is named as KAIST Micro Modular Reactor (MMR). As a part of ongoing research of conceptual design of KAIST MMR, preliminary design of power generation cycle was performed in this study. Since the targets of MMR are full modularization of a reactor system with S-CO 2 coolant, authors selected a simple recuperated S-CO 2 Brayton cycle as a power conversion system for KAIST MMR. The size of components of the S-CO 2 cycle is much smaller than existing helium Brayton cycle and steam Rankine cycle, and whole power conversion system can be contained with core and safety system in one containment vessel. From the investigation of the power conversion cycle, recompressing recuperated cycle showed higher efficiency than the simple recuperated cycle. However the volume of heat exchanger for recompressing cycle is too large so more space will be occupied by heat exchanger in the recompressing cycle than the simple recuperated cycle. Thus, authors consider that the simple recuperated cycle is more suitable for MMR. More research for the KAIST MMR will be followed in the future and detailed information of reactor core and safety system will be developed down the road. More refined cycle layout and design of turbomachinery and heat exchanger will be performed in the future study

Reactivity balance for a soluble boron-free small modular reactor

Directory of Open Access Journals (Sweden)

Lezani van der Merwe

2018-06-01

Full Text Available Elimination of soluble boron from reactor design eliminates boron-induced reactivity accidents and leads to a more negative moderator temperature coefficient. However, a large negative moderator temperature coefficient can lead to large reactivity feedback that could allow the reactor to return to power when it cools down from hot full power to cold zero power. In soluble boron-free small modular reactor (SMR design, only control rods are available to control such rapid core transient.The purpose of this study is to investigate whether an SMR would have enough control rod worth to compensate for large reactivity feedback. The investigation begins with classification of reactivity and completes an analysis of the reactivity balance in each reactor state for the SMR model.The control rod worth requirement obtained from the reactivity balance is a minimum control rod worth to maintain the reactor critical during the whole cycle. The minimum available rod worth must be larger than the control rod worth requirement to manipulate the reactor safely in each reactor state. It is found that the SMR does have enough control rod worth available during rapid transient to maintain the SMR at subcritical below k-effectives of 0.99 for both hot zero power and cold zero power. Keywords: Control Rod Worth, Reactivity Balance, Reactivity Feedback, Small Modular Reactor, Soluble Boron Free

Small size modular fast reactors in large scale nuclear power

International Nuclear Information System (INIS)

Zrodnikov, A.V.; Toshinsky, G.I.; Komlev, O.G.; Dragunov, U.G.; Stepanov, V.S.; Klimov, N.N.; Kopytov, I.I.; Krushelnitsky, V.N.

2005-01-01

The report presents an innovative nuclear power technology (NPT) based on usage of modular type fast reactors (FR) (SVBR-75/100) with heavy liquid metal coolant (HLMC) i. e. eutectic lead-bismuth alloy mastered for Russian nuclear submarines' (NS) reactors. Use of this NPT makes it possible to eliminate a conflict between safety and economic requirements peculiar to the traditional reactors. Physical features of FRs, an integral design of the reactor and its small power (100 MWe), as well as natural properties of lead-bismuth coolant assured realization of the inherent safety properties. This made it possible to eliminate a lot of safety systems necessary for the reactor installations (RI) of operating NPPs and to design the modular NPP which technical and economical parameters are competitive not only with those of the NPP based on light water reactors (LWR) but with those of the steam-gas electric power plant. Multipurpose usage of transportable reactor modules SVBR-75/100 of entirely factory manufacture assures their production in large quantities that reduces their fabrication costs. The proposed NPT provides economically expedient change over to the closed nuclear fuel cycle (NFC). When the uranium-plutonium fuel is used, the breeding ratio is over one. Use of proposed NPT makes it possible to considerably increase the investment attractiveness of nuclear power (NP) with fast neutron reactors even today at low costs of natural uranium. (authors)

Experience in Reviewing Small Modular Reactor Technology

International Nuclear Information System (INIS)

Ahmad Nabil Abdul Rahim; Alfred, S.L.; Phongsakorn, P.

2015-01-01

Malaysia is in the stage of conducting Preliminary Technical Feasibility Study for the Deployment of Small Modular Reactor (SMR). There are different types of SMR, some already under construction in Argentina (CAREM) and China (HTR-PM) - (light water reactor and high temperature reactor technologies), others with near-term deployment such as SMART in South Korea, ACP100 in China, mPower and NuScale in the US, and others with longer term deployment prospects (liquid-metal cooled reactor technologies). The study was mainly to get an overview of the technology available in the market. The SMR ranking in the study was done through listing out the most deployable technology in the market according to their types. As a new comer country, the proven technology with an excellent operation history will usually be the main consideration points. (author)

Working Towards Unified Safety Design Criteria for Modular High Temperature Gas-cooled Reactor Designs

International Nuclear Information System (INIS)

Reitsma, Frederik; Silady, Fred; Kunitomi, Kazuhiko

2014-01-01

The Nuclear Power Development Section of the IAEA recently received approval for a Coordinated Research Project (CRP) to investigate and make proposals on modular High Temperature Gas-cooled Reactor (HTGR) Safety design criteria. It is expected that these criteria would consider past experience and existing safety standards in the light of modular HTGR material and design characteristics to propose safety design criteria. It will consider the deterministic and risk-informed safety design standards that apply to the wide spectrum of Off- normal events under development worldwide for existing and planned HTGRs. The CRP would also take into account lessons from the Fukushima Daiichi accident, clarifying the safety approach and safety evaluation criteria for design and beyond design basis events, including those events that can affect multiple reactor modules and/or are dependent on the application proximate to the plant site. (e. g., industrial process steam/heat). The logical flow of criteria is from the fundamental inherent safety characteristics of modular HTGRs and associated expected performance characteristics, to the safety functions required to ensure those characteristics during the wide spectrum of Off-normal events, and finally to specific criteria related to those functions. This is detailed in the paper with specific examples included of how it may be applied. The results of the CRP will be made available to the member states and HTGR community. (author)

A small modular fast reactor as starting point for industrial deployment of fast reactors

International Nuclear Information System (INIS)

Chang, Yoon I.; Lo Pinto, Pierre; Konomura, Mamoru

2006-01-01

The current commercial reactors based on light water technology provide 17% of the electricity worldwide owing to their reliability, safety and competitive economics. In the near term, next generation reactors are expected to be evolutionary type, taking benefits of extensive LWR experience feedbacks and further improved economics and safety provisions. For the long term, however, sustainable energy production will be required due to continuous increase of the human activities, environmental concerns such as greenhouse effect and the need of alternatives to fossil fuels as long term energy resources. Therefore, future generation commercial reactors should meet some criteria of sustainability that the current generation cannot fully satisfy. In addition to the current objectives of economics and safety, waste management, resource extension and public acceptance become other major objectives among the sustainability criteria. From this perspective, two questions can be raised: what reactor type can meet the sustainability criteria, and how to proceed to an effective deployment in harmony with the high reliability and availability of the current nuclear reactor fleet. There seems to be an international consensus that the fast spectrum reactor, notably the sodium-cooled system is most promising to meet all of the long term sustainability criteria. As for the latter, we propose a small modular fast reactor project could become a base to prepare the industrial infrastructure. The paper has the following contents: - Introduction; - SMFR project; - Core design; - Supercritical CO 2 Brayton cycle; - Near-term reference plant; - Advanced designs; - Conclusions. To summarize, the sodium-cooled fast reactor is currently recognized as the technology of choice for the long term nuclear energy expansion, but some research and development are required to optimize and validate advanced design solutions. A small modular fast reactor can satisfy some existing near-term market niche

Multi-kilowatt modularized spacecraft power processing system development

International Nuclear Information System (INIS)

Andrews, R.E.; Hayden, J.H.; Hedges, R.T.; Rehmann, D.W.

1975-07-01

A review of existing information pertaining to spacecraft power processing systems and equipment was accomplished with a view towards applicability to the modularization of multi-kilowatt power processors. Power requirements for future spacecraft were determined from the NASA mission model-shuttle systems payload data study which provided the limits for modular power equipment capabilities. Three power processing systems were compared to evaluation criteria to select the system best suited for modularity. The shunt regulated direct energy transfer system was selected by this analysis for a conceptual design effort which produced equipment specifications, schematics, envelope drawings, and power module configurations

Modular High Temperature Gas-Cooled Reactor Safety Basis and Approach

Energy Technology Data Exchange (ETDEWEB)

David Petti; Jim Kinsey; Dave Alberstein

2014-01-01

Various international efforts are underway to assess the safety of advanced nuclear reactor designs. For example, the International Atomic Energy Agency has recently held its first Consultancy Meeting on a new cooperative research program on high temperature gas-cooled reactor (HTGR) safety. Furthermore, the Generation IV International Forum Reactor Safety Working Group has recently developed a methodology, called the Integrated Safety Assessment Methodology, for use in Generation IV advanced reactor technology development, design, and design review. A risk and safety assessment white paper is under development with respect to the Very High Temperature Reactor to pilot the Integrated Safety Assessment Methodology and to demonstrate its validity and feasibility. To support such efforts, this information paper on the modular HTGR safety basis and approach has been prepared. The paper provides a summary level introduction to HTGR history, public safety objectives, inherent and passive safety features, radionuclide release barriers, functional safety approach, and risk-informed safety approach. The information in this paper is intended to further the understanding of the modular HTGR safety approach. The paper gives those involved in the assessment of advanced reactor designs an opportunity to assess an advanced design that has already received extensive review by regulatory authorities and to judge the utility of recently proposed new methods for advanced reactor safety assessment such as the Integrated Safety Assessment Methodology.

Modular high-temperature gas-cooled reactor simulation using parallel processors

International Nuclear Information System (INIS)

Ball, S.J.; Conklin, J.C.

1989-01-01

The MHPP (Modular HTGR Parallel Processor) code has been developed to simulate modular high-temperature gas-cooled reactor (MHTGR) transients and accidents. MHPP incorporates a very detailed model for predicting the dynamics of the reactor core, vessel, and cooling systems over a wide variety of scenarios ranging from expected transients to very-low-probability severe accidents. The simulations routines, which had originally been developed entirely as serial code, were readily adapted to parallel processing Fortran. The resulting parallelized simulation speed was enhanced significantly. Workstation interfaces are being developed to provide for user (operator) interaction. In this paper the benefits realized by adapting previous MHTGR codes to run on a parallel processor are discussed, along with results of typical accident analyses

Fabrication and Testing of a Modular Micro-Pocket Fission Detector Instrumentation System for Test Nuclear Reactors

Science.gov (United States)

Reichenberger, Michael A.; Nichols, Daniel M.; Stevenson, Sarah R.; Swope, Tanner M.; Hilger, Caden W.; Roberts, Jeremy A.; Unruh, Troy C.; McGregor, Douglas S.

2018-01-01

Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Measurement of the neutron-flux distribution within the reactor core provides a more complete understanding of the operating conditions in the reactor than typical ex-core sensors. Micro-Pocket Fission Detectors have been developed and tested previously but have been limited to single-node operation and have utilized highly specialized designs. The development of a widely deployable, multi-node Micro-Pocket Fission Detector assembly will enhance nuclear research capabilities. A modular, four-node Micro-Pocket Fission Detector array was designed, fabricated, and tested at Kansas State University. The array was constructed from materials that do not significantly perturb the neutron flux in the reactor core. All four sensor nodes were equally spaced axially in the array to span the fuel-region of the reactor core. The array was filled with neon gas, serving as an ionization medium in the small cavities of the Micro-Pocket Fission Detectors. The modular design of the instrument facilitates the testing and deployment of numerous sensor arrays. The unified design drastically improved device ruggedness and simplified construction from previous designs. Five 8-mm penetrations in the upper grid plate of the Kansas State University TRIGA Mk. II research nuclear reactor were utilized to deploy the array between fuel elements in the core. The Micro-Pocket Fission Detector array was coupled to an electronic support system which has been specially developed to support pulse-mode operation. The Micro-Pocket Fission Detector array composed of four sensors was used to monitor local neutron flux at a constant reactor power of 100 kWth at different axial locations simultaneously. The array was positioned at five different radial locations within the core to emulate the deployment of multiple arrays and develop a 2-dimensional measurement of

On modular stellarator reactor coils

International Nuclear Information System (INIS)

Rau, F.; Harmeyer, E.; Kisslinger, J.; Wobig, H.

1985-01-01

Modular twisted coils are discussed which produce magnetic fields of the Advanced Stellarator WENDELSTEIN VII-AS type. Reducing the number coils/FP offers advantage for maintenance of coils, but increases the magnetic ripple and B m /B o . Computation of force densities within the coils of ASR and ASB yield local maximum values of about 80 and 180 MN/m 3 , respectively. A system of mutual coil support is being developed. Twisted coils in helical arrangement provide a reactor-sized HELIAC system. In order to reduce the magnetic ripple, a large number of 14 coils/FP in special arrangement is used

Studies of a modular advanced stellarator reactor ASRA6C

International Nuclear Information System (INIS)

Boehme, G.; Jentzsch, K.; Komarek, P.; Maurer, W.; El-Guebaly, L.A.; Emmert, G.A.; Kulcinski, G.L.; Larsen, E.M.; Sanatarius, J.F.; Schawan, M.E.; Scharer, J.E.; Sviatoslavski, I.N.; Vogelsang, W.F.; Walstrom, P.L.; Wittenberg, L.J.; Grieger, G.; Harmeyer, E.; Herrnegger, F.; Kisslinger, J.; Rau, F.; Wobig, H.

1987-05-01

This study is directed towards the clarification of critical issues of advanced modular stellerator reactors exploiting the inherent potential of steady state operation, and is not a point design study of a reactor. Critical technology issues arise from the three-dimensional magnetic field structure. The first wall, blanket and shield are more complex than those of axi-symmetric systems, but this is eased at moderate to large aspect ratio typical of stellerators. Several blanket options have been studied and a thin blanket (21 cm) was the first choice for the design. Superconducting modular coils were investigated with respect to the conductor and mechanical supports. From the analysis of forces and stresses caused by the electromagnetic loads the coils are considered to be feasible, although shear stresses might pose a critical issue. Demountable intermagnetic support elements were designed for use at separation areas between the cryostat modules. A scheme for remote reactor maintenance was also developed. The plasma physics issues of different configurations were studied using extrapolations of transort behaviour and equilibrium from theory and present experiments. These studies indicate that the confinement and equilibrium behaviour is adequate for ignited operation at an average value of 5% beta. Impurities may pose a critical issue. Several impurity control operations were investigated; a pumped limiter configuration utilizing the 'ergodic layer' at the plasma edge was chosen for edge plasma and impurity control. A general conclusion of the study is that the modular stellerator configuration offers interesting prospects regarding the development towards steady-state reactors. (orig.)

Studies of a modular advanced stellarator reactor ASRA6C

International Nuclear Information System (INIS)

Boehme, G.; El-Guebaly, L.A.; Emmert, G.A.; Grieger, G.; Harmeyer, E.; Herrnegger, F.; Huebener, J.; Jentzsch, K.; Kisslinger, J.; Komarek, P.; Kulcinski, G.L.; Larsen, E.M.; Maurer, W.; Rau, F.; Santarius, J.F.; Sawan, M.E.; Scharer, J.E.; Sviatoslavsky, I.N.; Vogelsang, W.F.; Walstrom, P.L.; Wittenberg, L.J.; Wobig, H.

1987-06-01

This study is directed towards the clarification of critical issues of advanced modular stellerator reactors exploiting the inherent potential of steady state operation, and is not a point design study of a reactor. Critical technology issues arise from the three-dimensional magnetic field structure. The first wall, blanket and shield are more complex than those of axi-symmetric systems, but this is eased at moderate to large aspect ratio typical of stellarators. Several blanket options have been studied and a thin blanket (21 cm) was the first choice for the design. Superconducting modular coils were investigated with respect to the conductor and mechanical supports. From the analysis of forces and stresses caused by the electromagnetic loads the coils are considered to be feasible, although shear stresses might pose a critical issue. Demountable intermagnetic support elements were designed for use at separation areas between the cryostat modules. A scheme for remote reactor maintenance was also developed. The plasma physics issues of different configurations were studied using extrapolations of transport behaviour and equilibrium from theory and present experiments. These studies indicate that the confinement and equilibrium behaviour is adequate for ignited operation at an average value of 5% beta. Impurities may pose a critical issue. Several impurity control operations were investigated; a pumped limiter configuration utilizing the 'ergodic layer' at the plasma edge was chosen for edge plasma and impurity control. A general conclusion of the study is that the modular stellerator configuration offers interesting prospects regarding the development towards steady-state reactors. (orig.) [de

Study on layout and construction concept of DMS (modular simplified medium small reactor)

International Nuclear Information System (INIS)

Shizuka Hirako; Yuusuke Shimizu; Shigeru Yokouchi; Yoshinori Iimura; Yuuji Yasuda; Kumiaki Moriya; Takahiko Hida

2005-01-01

Nuclear power is expected to become the main source of electric power generation in Japan for reasons of energy security and prevention of CO 2 emissions. In addition, the recent slowdown of electric power demand and the liberalization of the electric power market are accelerating medium and small sized reactor development. Under these circumstances, DMS's (modular simplified and medium small reactors) have been developed as 400 MWe class LWR's supported by the Japan Atomic Power Company. In the development of medium and small sized reactors, the most important point is how to overcome the scale demerits. To this end, we have pursued not only the simplification of systems and equipment but also the standardization of layout and construction. The main technical feature of DMS's is the adoption of a natural circulation reactor with short length fuel. Short length fuel enables the reduction of RPV height as well as construction volume of the PCV and building volume. A natural circulation reactor has considerable rationalizing effects such as the elimination of re-circulation pumps and their drive power source. By applying simplified systems and equipment, a rationalized layout and construction method are adopted. To improve the constructability by means of modular construction methods, steel containment is applied. The PCV size is reduced to 17 m in diameter and 24 m in height by applying a dish-shaped drywell and eccentric RPV arrangement. By applying a compact PCV and concentrated equipment arrangement in building, it can be confirmed that the ratio of building volume per unit power is equivalent to that of existing large sized ABWRs. Furthermore, a steel plate reinforced concrete structure (SC structure) is applied to the building layout. The application of the compact PCV (steel containment) and the SC structure makes it easier to apply a large-scale module, such as an integrated steel containment and SC structure module, and an integrated multi-layer BM (building

Advances in global development and deployment of small modular reactors and incorporating lessons learned from the Fukushima Daiichi accident into the designs of engineered safety features of advanced reactors

International Nuclear Information System (INIS)

Hadid Subki, M.; )

2014-01-01

The IAEA has been facilitating the Member States in incorporating the lessons-learned from the Fukushima Dai-ichi Accident into the designs of engineered safety features of advanced reactors, including small modular reactors. An extended assessment is required to address challenges for advancing reactor safety in the new evolving generation of SMR plants to preserve the historic lessons in safety, through: assuring the diversity in emergency core cooling systems following loss of onsite AC power; ensuring diversity in reactor depressurization following a transient or accident; confirming independence in reactor trip and safety systems for sensors, power supplies and actuation systems, and finally diversity in maintaining containment integrity following a severe accident

Gas turbine modular helium reactor in cogeneration; Turbina de gas reactor modular con helio en cogeneracion

Energy Technology Data Exchange (ETDEWEB)

Leon de los Santos, G. [UNAM, Facultad de Ingenieria, Division de Ingenieria Electrica, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Mexico, D. F. (Mexico)], e-mail: tesgleon@gmail.com

2009-10-15

This work carries out the thermal evaluation from the conversion of nuclear energy to electric power and process heat, through to implement an outline gas turbine modular helium reactor in cogeneration. Modeling and simulating with software Thermo flex of Thermo flow the performance parameters, based on a nuclear power plant constituted by an helium cooled reactor and helium gas turbine with three compression stages, two of inter cooling and one regeneration stage; more four heat recovery process, generating two pressure levels of overheat vapor, a pressure level of saturated vapor and one of hot water, with energetic characteristics to be able to give supply to a very wide gamma of industrial processes. Obtaining a relationship heat electricity of 0.52 and efficiency of net cogeneration of 54.28%, 70.2 MW net electric, 36.6 MW net thermal with 35% of condensed return to 30 C; for a supplied power by reactor of 196.7 MW; and with conditions in advanced gas turbine of 850 C and 7.06 Mpa, assembly in a shaft, inter cooling and heat recovery in cogeneration. (Author)

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

International Nuclear Information System (INIS)

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

2014-01-01

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

Molten salt small modular reactors (MSSMRs): from DMSR to SmAHTR

International Nuclear Information System (INIS)

LeBlanc, D.

2013-01-01

Molten salt reactors were developed extensively from the 1950s to 1970s as a thermal breeder alternative on the Thorium-U233 cycle. Simplified designs running as fluid fuel convertors without salt processing as well as TRISO fueled, salt cooled reactors both hold much promise as potential small modular reactors. A background will be presented along with the most likely routes forward for a Canadian development program. (author)

Human Reliability Analysis for Small Modular Reactors

Energy Technology Data Exchange (ETDEWEB)

Ronald L. Boring; David I. Gertman

2012-06-01

Because no human reliability analysis (HRA) method was specifically developed for small modular reactors (SMRs), the application of any current HRA method to SMRs represents tradeoffs. A first- generation HRA method like THERP provides clearly defined activity types, but these activity types do not map to the human-system interface or concept of operations confronting SMR operators. A second- generation HRA method like ATHEANA is flexible enough to be used for SMR applications, but there is currently insufficient guidance for the analyst, requiring considerably more first-of-a-kind analyses and extensive SMR expertise in order to complete a quality HRA. Although no current HRA method is optimized to SMRs, it is possible to use existing HRA methods to identify errors, incorporate them as human failure events in the probabilistic risk assessment (PRA), and quantify them. In this paper, we provided preliminary guidance to assist the human reliability analyst and reviewer in understanding how to apply current HRA methods to the domain of SMRs. While it is possible to perform a satisfactory HRA using existing HRA methods, ultimately it is desirable to formally incorporate SMR considerations into the methods. This may require the development of new HRA methods. More practicably, existing methods need to be adapted to incorporate SMRs. Such adaptations may take the form of guidance on the complex mapping between conventional light water reactors and small modular reactors. While many behaviors and activities are shared between current plants and SMRs, the methods must adapt if they are to perform a valid and accurate analysis of plant personnel performance in SMRs.

Control, Co-generation, and Sensor Placement Strategy for Integral Small Modular Reactors

International Nuclear Information System (INIS)

Upadhyaya, Belle-R.; Fan, Li; Hines, J.-Wesley; Perillo, Sergio-R. P.

2011-01-01

The development of Small Modular Reactors (SMR) has multiple applications for electricity generation, process heat, hydrogen production, and others. The results of research, development, and demonstration (RD and D) of load-following control design for multiple modules, nuclear desalination, and sensor placement strategy for enhanced fault detection and isolation, are presented in this paper. The technologies are demonstrated with application to an integral pressurized water reactor (IPWR) such as the IRIS reactor. The outcomes of this RD and D include the development of a complete dynamic model of the IRIS system, load following control under dual-module steam mixing, nuclear desalination with a multi-stage flash (MSF) desalination plant, and automated technique for sensor allocation in a combined reactor and balance-of-plant system. The dynamic performance of a nuclear power station comprised of two IRIS reactor modules, operating simultaneously with a common steam header with steam mixing, was evaluated. The control problem addressed 'load-following' scenarios, such as varying load during the day or reduced consumption during the weekend. To solve this problem, a single-module IRIS MATLAB-Simulink model was developed and used to quantify the responses from both modules. The resulting model was subjected to eight different perturbation cases to analyze its capability of detecting small perturbations, therefore testing its robustness and sensitivity. The prospects of using nuclear energy for seawater desalination on a large scale can be very attractive since desalination is an energy intensive process that can utilize the heat from a nuclear reactor and/or the electricity produced by such plants. Small modular reactors, ranging from 50 MWe to 300 MWe, offer the largest potential as coupling options to nuclear desalination systems. However, coupling a nuclear plant and a desalination plant involves a number of issues that have to be addressed. Among these issues

Control, Co-generation, and Sensor Placement Strategy for Integral Small Modular Reactors

Energy Technology Data Exchange (ETDEWEB)

Upadhyaya, Belle-R.; Fan, Li; Hines, J.-Wesley [University of Tennessee, Knoxville (United States); Perillo, Sergio-R. P. [Instituto de Pesquisas Energeticas e Nucleares, Sao Paulo (Brazil)

2011-08-15

The development of Small Modular Reactors (SMR) has multiple applications for electricity generation, process heat, hydrogen production, and others. The results of research, development, and demonstration (RD and D) of load-following control design for multiple modules, nuclear desalination, and sensor placement strategy for enhanced fault detection and isolation, are presented in this paper. The technologies are demonstrated with application to an integral pressurized water reactor (IPWR) such as the IRIS reactor. The outcomes of this RD and D include the development of a complete dynamic model of the IRIS system, load following control under dual-module steam mixing, nuclear desalination with a multi-stage flash (MSF) desalination plant, and automated technique for sensor allocation in a combined reactor and balance-of-plant system. The dynamic performance of a nuclear power station comprised of two IRIS reactor modules, operating simultaneously with a common steam header with steam mixing, was evaluated. The control problem addressed 'load-following' scenarios, such as varying load during the day or reduced consumption during the weekend. To solve this problem, a single-module IRIS MATLAB-Simulink model was developed and used to quantify the responses from both modules. The resulting model was subjected to eight different perturbation cases to analyze its capability of detecting small perturbations, therefore testing its robustness and sensitivity. The prospects of using nuclear energy for seawater desalination on a large scale can be very attractive since desalination is an energy intensive process that can utilize the heat from a nuclear reactor and/or the electricity produced by such plants. Small modular reactors, ranging from 50 MWe to 300 MWe, offer the largest potential as coupling options to nuclear desalination systems. However, coupling a nuclear plant and a desalination plant involves a number of issues that have to be addressed. Among these

Severe water ingress accident analysis for a Modular High Temperature Gas Cooled Reactor

International Nuclear Information System (INIS)

Zhang Zuoyi; Scherer, Winfried

1997-01-01

This paper analyzes the severe water ingress accidents in the SIEMENS 200MW Modular High Temperature Gas Cooled Reactor (HTR-Module) under the assumption of no active safety protection systems in order to find the safety margin of the current HTR-Module design. A water, steam and helium multi-phase cavity model is originally developed and implemented in the DSNP simulation system. The developed DSNP system is used to simulate the primary circuit of HTR-Module power plant. The comparisons of the models with the TINTE calculations validate the current simulation. After analyzing the effects of blower separation on water droplets, the wall heat storage, etc., it is found that the maximum H 2 O density increase rate in the reactor core is smaller than 0.3 kg/(m 3 s). The liquid water vaporization in the steam generator and H 2 O transport from the steam generator to the reactor core reduces the impulse of the H 2 O in the reactor core. The nuclear reactivity increase caused by the water ingress leads to a fast power excursion, which, however, is inherently counterbalanced by negative feedback effects. Concerning the integrity of the fuel elements, the safety relevant temperature limit of 1600degC was not reached in any case. (author)

Modular Lead-Bismuth Fast Reactors in Nuclear Power

Directory of Open Access Journals (Sweden)

Vladimir Petrochenko

2012-09-01

Full Text Available On the basis of the unique experience of operating reactors with heavy liquid metal coolant–eutectic lead-bismuth alloy in nuclear submarines, the concept of modular small fast reactors SVBR-100 for civilian nuclear power has been developed and validated. The features of this innovative technology are as follows: a monoblock (integral design of the reactor with fast neutron spectrum, which can operate using different types of fuel in various fuel cycles including MOX fuel in a self-providing mode. The reactor is distinct in that it has a high level of self-protection and passive safety, it is factory manufactured and the assembled reactor can be transported by railway. Multipurpose application of the reactor is presumed, primarily, it can be used for regional power to produce electricity, heat and for water desalination. The Project is being realized within the framework of state-private partnership with joint venture OJSC “AKME-Engineering” established on a parity basis by the State Atomic Energy Corporation “Rosatom” and the Limited Liability Company “EuroSibEnergo”.

Advanced I&C for Fault-Tolerant Supervisory Control of Small Modular Reactors

Energy Technology Data Exchange (ETDEWEB)

Cole, Daniel G. [Univ. of Pittsburgh, PA (United States)

2018-01-30

In this research, we have developed a supervisory control approach to enable automated control of SMRs. By design the supervisory control system has an hierarchical, interconnected, adaptive control architecture. A considerable advantage to this architecture is that it allows subsystems to communicate at different/finer granularity, facilitates monitoring of process at the modular and plant levels, and enables supervisory control. We have investigated the deployment of automation, monitoring, and data collection technologies to enable operation of multiple SMRs. Each unit's controller collects and transfers information from local loops and optimize that unit’s parameters. Information is passed from the each SMR unit controller to the supervisory controller, which supervises the actions of SMR units and manage plant processes. The information processed at the supervisory level will provide operators the necessary information needed for reactor, unit, and plant operation. In conjunction with the supervisory effort, we have investigated techniques for fault-tolerant networks, over which information is transmitted between local loops and the supervisory controller to maintain a safe level of operational normalcy in the presence of anomalies. The fault-tolerance of the supervisory control architecture, the network that supports it, and the impact of fault-tolerance on multi-unit SMR plant control has been a second focus of this research. To this end, we have investigated the deployment of advanced automation, monitoring, and data collection and communications technologies to enable operation of multiple SMRs. We have created a fault-tolerant multi-unit SMR supervisory controller that collects and transfers information from local loops, supervise their actions, and adaptively optimize the controller parameters. The goal of this research has been to develop the methodologies and procedures for fault-tolerant supervisory control of small modular reactors. To achieve

Loading Analysis of Modular Multi-level Converter for Offshore High-voltage DC Application under Various Grid Faults

DEFF Research Database (Denmark)

Liu, Hui; Ma, Ke; Loh, Poh Chiang

2016-01-01

challenges but may also result in overstressed components for the modular multi-level converter. However, the thermal loading of the modular multi-level converter under various grid faults has not yet been clarified. In this article, the power loss and thermal performance of the modular multi-level converter...... low-voltage ride-through strongly depend on the types and severity values of grid voltage dips. The thermal distribution among the three phases of the modular multi-level converter may be quite uneven, and some devices are much more stressed than the normal operating condition, which may...

Power optimization in the STAR-LM modular natural convection reactor system. Topic 2.1 advanced reactor power plants

International Nuclear Information System (INIS)

Spencer, B.W.; Sienicki, J.J.; Farmer, M.T.

2001-01-01

The secure, transportable, autonomous reactor (STAR) project addresses the needs of developing countries and independent power producers for a small (300 MWt), multi-purpose energy system. The STAR-LM variant described here is a liquid metal cooled, fast spectrum reactor system. Previous development of a reference STAR-LM design resulted in a 300 MWt modular, pool- type reactor based on criteria for factory fabrication of modules, full transportability of modules (barge, rail, overland), fast construction and startup, and semi-autonomous operation. Earlier work on the reference 300 MWt concept focused first on addressing whether 100% natural circulation heat transport was achievable under the module size constraints for full transportability and under the coolant and cladding peak temperature limitations imposed by the existing Russian database for ferritic-martensitic core material with oxide-layer corrosion protection. Secondly, owing to uncertainties and limitations in the available Russian materials compatibility database, the objective of the reference design was to address how low the coolant and cladding peak temperatures could be commensurate with achieving 300 MWt power level with 100% natural circulation in a fully transportable module size. In the present work we have refocused the approach to attempt to maximize the power achievable in the reactor module based on preserving the criteria for full module transportability and remaining within the materials compatibility database limits. (author)

Power optimization in the star-LM modular, natural convection reactor system

International Nuclear Information System (INIS)

Spencer, B.W.; Sienicki, J.J.; Farmer, M.T.

2001-01-01

The secure, transportable, autonomous reactor (STAR) project addresses the needs of developing countries and independent power producers for small, multi-purpose energy systems, which operate near autonomously for very long term. The STAR-LM variant described here is a liquid metal cooled, fast reactor system. Previous development of STAR-LM resulted in a 300 MWt modular, pool-type reactor based on criteria for factory fabrication, full transportability (barge, overland, rail), and fast construction and startup. Steam generator modules are placed directly into the primary heat transport circuit, eliminating the intermediate heat transport loop. Natural convection heat transport at all power levels eliminates the need for main coolant pumps. Seismic isolation eliminates concern about seismic and sloshing-related loads in the pool configuration. Even end-of-spectrum postulated events such as loss-of-heat sink with failure to scram are terminated passively by inherent core power shutdown, and decay heat is passively rejected to the atmospheric air inexhaustible heat sink by guard vessel exterior cooling. Recent concept development has focused on maximizing the power achievable in a small module size based on preserving key criteria for: full spectrum of modes of module transport from factory to site (including rail transport); 100% natural circulation heat transport; ultra-long core cartridge lifetime; coolant and cladding peak temperatures well within the existing (Russian) database for Pb/Bi coolant and ferritic steel core materials. (author)

Technology selection for offshore underwater small modular reactors

International Nuclear Information System (INIS)

Shivan, Koroush; Ballinger, Ronald; Buongiorno, Jacopo; Forsberg, Charles; Kazimi, Mujid; Todreas, Neil

2016-01-01

This work examines the most viable nuclear technology options for future underwater designs that would meet high safety standards as well as good economic potential, for construction in the 2030-2040 time frame. The top five concepts selected from a survey of 13 nuclear technologies were compared to a small modular pressurized water reactor (PWR) designed with a conventional layout. In order of smallest to largest primary system size where the reactor and all safety systems are contained, the top five designs were: (1) a lead-bismuth fast reactor based on the Russian SVBR-100; (2) a novel organic cooled reactor; (3) an innovative superheated water reactor; (4) a boiling water reactor based on Toshiba's LSBWR; and (5) an integral PWR featuring compact steam generators. A similar study on potential attractive power cycles was also performed. A condensing and recompression supercritical CO 2 cycle and a compact steam Rankine cycle were designed. It was found that the hull size required by the reactor, safety systems and power cycle can be significantly reduced (50-80%) with the top five designs compared to the conventional PWR. Based on the qualitative economic consideration, the organic cooled reactor and boiling water reactor designs are expected to be the most cost effective options

Technology selection for offshore underwater small modular reactors

Energy Technology Data Exchange (ETDEWEB)

Shivan, Koroush; Ballinger, Ronald; Buongiorno, Jacopo; Forsberg, Charles; Kazimi, Mujid; Todreas, Neil [Dept. of Nuclear Science and Engineering, Massachusetts Institute of Technology, Cambridge (United States)

2016-12-15

This work examines the most viable nuclear technology options for future underwater designs that would meet high safety standards as well as good economic potential, for construction in the 2030-2040 time frame. The top five concepts selected from a survey of 13 nuclear technologies were compared to a small modular pressurized water reactor (PWR) designed with a conventional layout. In order of smallest to largest primary system size where the reactor and all safety systems are contained, the top five designs were: (1) a lead-bismuth fast reactor based on the Russian SVBR-100; (2) a novel organic cooled reactor; (3) an innovative superheated water reactor; (4) a boiling water reactor based on Toshiba's LSBWR; and (5) an integral PWR featuring compact steam generators. A similar study on potential attractive power cycles was also performed. A condensing and recompression supercritical CO{sub 2} cycle and a compact steam Rankine cycle were designed. It was found that the hull size required by the reactor, safety systems and power cycle can be significantly reduced (50-80%) with the top five designs compared to the conventional PWR. Based on the qualitative economic consideration, the organic cooled reactor and boiling water reactor designs are expected to be the most cost effective options.

Technology Selection for Offshore Underwater Small Modular Reactors

Directory of Open Access Journals (Sweden)

Koroush Shirvan

2016-12-01

Full Text Available This work examines the most viable nuclear technology options for future underwater designs that would meet high safety standards as well as good economic potential, for construction in the 2030–2040 timeframe. The top five concepts selected from a survey of 13 nuclear technologies were compared to a small modular pressurized water reactor (PWR designed with a conventional layout. In order of smallest to largest primary system size where the reactor and all safety systems are contained, the top five designs were: (1 a lead–bismuth fast reactor based on the Russian SVBR-100; (2 a novel organic cooled reactor; (3 an innovative superheated water reactor; (4 a boiling water reactor based on Toshiba's LSBWR; and (5 an integral PWR featuring compact steam generators. A similar study on potential attractive power cycles was also performed. A condensing and recompression supercritical CO2 cycle and a compact steam Rankine cycle were designed. It was found that the hull size required by the reactor, safety systems and power cycle can be significantly reduced (50–80% with the top five designs compared to the conventional PWR. Based on the qualitative economic consideration, the organic cooled reactor and boiling water reactor designs are expected to be the most cost effective options.

Xenon-induced power oscillations in a generic small modular reactor

Science.gov (United States)

Kitcher, Evans Damenortey

As world demand for energy continues to grow at unprecedented rates, the world energy portfolio of the future will inevitably include a nuclear energy contribution. It has been suggested that the Small Modular Reactor (SMR) could play a significant role in the spread of civilian nuclear technology to nations previously without nuclear energy. As part of the design process, the SMR design must be assessed for the threat to operations posed by xenon-induced power oscillations. In this research, a generic SMR design was analyzed with respect to just such a threat. In order to do so, a multi-physics coupling routine was developed with MCNP/MCNPX as the neutronics solver. Thermal hydraulic assessments were performed using a single channel analysis tool developed in Python. Fuel and coolant temperature profiles were implemented in the form of temperature dependent fuel cross sections generated using the SIGACE code and reactor core coolant densities. The Power Axial Offset (PAO) and Xenon Axial Offset (XAO) parameters were chosen to quantify any oscillatory behavior observed. The methodology was benchmarked against results from literature of startup tests performed at a four-loop PWR in Korea. The developed benchmark model replicated the pertinent features of the reactor within ten percent of the literature values. The results of the benchmark demonstrated that the developed methodology captured the desired phenomena accurately. Subsequently, a high fidelity SMR core model was developed and assessed. Results of the analysis revealed an inherently stable SMR design at beginning of core life and end of core life under full-power and half-power conditions. The effect of axial discretization, stochastic noise and convergence of the Monte Carlo tallies in the calculations of the PAO and XAO parameters was investigated. All were found to be quite small and the inherently stable nature of the core design with respect to xenon-induced power oscillations was confirmed. Finally, a

Industrial applications of multi-functional, multi-phase reactors

NARCIS (Netherlands)

Harmsen, G.J.; Chewter, L.A.

1999-01-01

To reveal trends in the design and operation of multi-functional, multi-phase reactors, this paper describes, in historical sequence, three industrial applications of multi-functional, multi-phase reactors developed and operated by Shell Chemicals during the last five decades. For each case, we

Development of a system model for advanced small modular reactors.

Energy Technology Data Exchange (ETDEWEB)

Lewis, Tom Goslee,; Holschuh, Thomas Vernon,

2014-01-01

This report describes a system model that can be used to analyze three advance small modular reactor (SMR) designs through their lifetime. Neutronics of these reactor designs were evaluated using Monte Carlo N-Particle eXtended (MCNPX/6). The system models were developed in Matlab and Simulink. A major thrust of this research was the initial scoping analysis of Sandias concept of a long-life fast reactor (LLFR). The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional light water reactors (LWRs) or other SMR designs (e.g. high temperature gas reactor (HTGR)). The system model has subroutines for lifetime reactor feedback and operation calculations, thermal hydraulic effects, load demand changes and a simplified SCO2 Brayton cycle for power conversion.

The Research on Operation Strategy of Nuclear Power Plant with Multi-reactors

Energy Technology Data Exchange (ETDEWEB)

Fang, Maoyao; Peng, Minjun; Cheng Shouyu [Harbin Engineering University, Harbin (China)

2014-08-15

In this paper, the operation characteristics and control strategy of nuclear power plant (NPP) with multi-modular pressurized water reactors (PWR) were researched through simulation. The main objective of this research was to ensure the coordinated operation and satisfy the convenience of turbine-generator and reactor's load adjustment in NPP with multi-reactors (MR). According to the operation characteristics of MR-NPP, the operation and control strategy was proposed, which was 'he average allocation of load for each reactor and maintaining average temperature of coolant at a constant? The control system was designed based the operation and control strategy. In order to research the operation characteristics and control strategy of MR-NPP, the paper established the transient analysis model which included the reactors and thermal hydraulic models, turbine model, could simulate and analyze on different operating conditions such as load reducing, load rising. Based on the proposed operation and control strategy and simulation models, the paper verified and validated the operation strategy and control system through load reducing, load rising. The results of research simulation showed that the operation strategy was feasible and can make the MR-NPP running safely as well as steadily on different operating conditions.

The Research on Operation Strategy of Nuclear Power Plant with Multi-reactors

International Nuclear Information System (INIS)

Fang, Maoyao; Peng, Minjun; Cheng Shouyu

2014-01-01

In this paper, the operation characteristics and control strategy of nuclear power plant (NPP) with multi-modular pressurized water reactors (PWR) were researched through simulation. The main objective of this research was to ensure the coordinated operation and satisfy the convenience of turbine-generator and reactor's load adjustment in NPP with multi-reactors (MR). According to the operation characteristics of MR-NPP, the operation and control strategy was proposed, which was 'he average allocation of load for each reactor and maintaining average temperature of coolant at a constant? The control system was designed based the operation and control strategy. In order to research the operation characteristics and control strategy of MR-NPP, the paper established the transient analysis model which included the reactors and thermal hydraulic models, turbine model, could simulate and analyze on different operating conditions such as load reducing, load rising. Based on the proposed operation and control strategy and simulation models, the paper verified and validated the operation strategy and control system through load reducing, load rising. The results of research simulation showed that the operation strategy was feasible and can make the MR-NPP running safely as well as steadily on different operating conditions

Feasibility study on small modular reactors for modern microgrids

Energy Technology Data Exchange (ETDEWEB)

Islam, R.; Gabbar, H.A., E-mail: hossam.gabbar@uoit.ca [Univ. of Ontario Inst. of Technology, Faculty of Energy Systems and Nuclear Science, Oshawa, Ontario (Canada)

2013-07-01

Microgrid is a solution of conventional power grid problem and offer sustainable decentralized power system. Microgrid with modern distributed energy resources (DER) could play an important role to alleviate dependency on the main electricity grid. Distributed energy resource comprises wind turbine, solar photovoltaic, diesel generator, gas engine, micro turbine, fuel cells, etc.Due to the gap between typical loads and supply within microgrid, larger scale energy generation could provide a possible solution to balance power demand and supply. Feasibility study of Small Nuclear Power Plant, such as Small Modular reactor (SMR), within microgrids could be achieved via different cases. To achieve the target, a comprehensive feasibility study is conducted on microgrid with SMR through electricity generation profiles, geographical and environmental assessment, as well as cost analysis using simulation practices and data analysis.Also potency of SMRs is analyzed. Parameters and Key Performance Indicators (KPIs) could be analyzed to achieve feasible solution of microgrids with small modular reactor (SMR) to improve the overall microgrid performance.The study shows that SMR could be a feasible solution if microgrid parameters are selected properly. (author)

Feasibility study on small modular reactors for modern microgrids

International Nuclear Information System (INIS)

Islam, R.; Gabbar, H.A.

2013-01-01

Microgrid is a solution of conventional power grid problem and offer sustainable decentralized power system. Microgrid with modern distributed energy resources (DER) could play an important role to alleviate dependency on the main electricity grid. Distributed energy resource comprises wind turbine, solar photovoltaic, diesel generator, gas engine, micro turbine, fuel cells, etc.Due to the gap between typical loads and supply within microgrid, larger scale energy generation could provide a possible solution to balance power demand and supply. Feasibility study of Small Nuclear Power Plant, such as Small Modular reactor (SMR), within microgrids could be achieved via different cases. To achieve the target, a comprehensive feasibility study is conducted on microgrid with SMR through electricity generation profiles, geographical and environmental assessment, as well as cost analysis using simulation practices and data analysis.Also potency of SMRs is analyzed. Parameters and Key Performance Indicators (KPIs) could be analyzed to achieve feasible solution of microgrids with small modular reactor (SMR) to improve the overall microgrid performance.The study shows that SMR could be a feasible solution if microgrid parameters are selected properly. (author)

The role of small modular reactors in enhancing energy security in developing countries

International Nuclear Information System (INIS)

Kessides, I. N.; Kuznestov, V.

2018-01-01

In recent years, small modular reactors (SMRs) have been attracting considerable attention around the world. SMR designs incorporate innovative approaches to achieve simplicity, modularity and speed of build, passive safety features, proliferation resistance, and reduced financial risk. The incremental capacity expansion associated with SMR deployment could provide a better match (than the large-scale reactors) to the limited grid capacity of many developing countries. Because of their lower capital requirements, SMRs could also effectively address the energy needs of small developing countries with limited financial resources. Although SMRs can have substantially higher specific capital costs as compared to large-scale reactors, they may nevertheless enjoy significant economic benefits due to shorter build times, accelerated learning effects and co-siting economies, temporal and sizing flexibility of deployment, and design simplification. (author)

The role of the Small Modular Reactors for isolated electric systems in Brazil

Energy Technology Data Exchange (ETDEWEB)

Mattos, João Roberto Loureiro de; Dias, Marcio Soares, E-mail: jrmattos@cdtn.br, E-mail: marciod@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2017-07-01

Small modular reactors (SMRs) are being considered as a useful option to reduce the total capital costs of nuclear power plants nad to provide power in small or off-grid systems. The SMRs offer simpler, standardized and safer modular design by being factory built, requiring smaller initial capital investment, and having shorter construction times. SMRs could be small enough to be transportable, could be used in isolated locations without advanced infrastructure and without power grid, or could be clustered in a single site to provide a multi-module, large capacity power plant. In this sense, Brazil has about 200 isolated systems in the Amazon region with no connection to the national grid. SMRs could be suitable to serve some of these smaller energy markets with less developed infrastructure, replacing large diesel generators, with expensive fuel cost, often noisy, highly polluting and have low overall efficiency. Despite the economics of SMRs have yet to be proven, the Brazilian growth of electricity generation will demand all available sources of energy and some propose not to discard their evaluation in terms of SMR applications. This paper addresses the potential application of SMRs with focus in specific Brazilian regions. (author)

The role of the Small Modular Reactors for isolated electric systems in Brazil

International Nuclear Information System (INIS)

Mattos, João Roberto Loureiro de; Dias, Marcio Soares

2017-01-01

Small modular reactors (SMRs) are being considered as a useful option to reduce the total capital costs of nuclear power plants nad to provide power in small or off-grid systems. The SMRs offer simpler, standardized and safer modular design by being factory built, requiring smaller initial capital investment, and having shorter construction times. SMRs could be small enough to be transportable, could be used in isolated locations without advanced infrastructure and without power grid, or could be clustered in a single site to provide a multi-module, large capacity power plant. In this sense, Brazil has about 200 isolated systems in the Amazon region with no connection to the national grid. SMRs could be suitable to serve some of these smaller energy markets with less developed infrastructure, replacing large diesel generators, with expensive fuel cost, often noisy, highly polluting and have low overall efficiency. Despite the economics of SMRs have yet to be proven, the Brazilian growth of electricity generation will demand all available sources of energy and some propose not to discard their evaluation in terms of SMR applications. This paper addresses the potential application of SMRs with focus in specific Brazilian regions. (author)

Sensitivity study of the factors affecting first-of-a-kind and Nth-of-a-kind Small Modular Reactor investment costs - 15108

International Nuclear Information System (INIS)

Boldon, L.; Liu, L.; Sabharwall, P.; Schneider, E.

2015-01-01

Conventional nuclear reactors require significant investment and pose financial, regulatory, and political risks. Small modular reactors (SMRs), on the other hand, provide an attractive alternative due to substantially reduced up-front investment costs and financial risks; more flexible deployment schedules and operations; modular fabrication and faster onsite assembly; and incorporation into hybrid energy systems or even cogeneration applications such as desalination, district heating, and industrial process heat. This paper aims to develop a general methodology of assessing the economics of first-of-a-kind (FOAK) and nth-of-a-kind (NOAK) SMRs as part of an effort to determine the conditions that would make them competitive with other generation technologies. Ultimately, this paper will provide background information on the various factors impacting nuclear reactor capital and operational costs; detail technology independent cost estimates for water-cooled SMRs; and provide a parametric study on learning and construction and deployment effects. This study demonstrates precisely why the financial risks are drastically reduced for an SMR site, making SMR nuclear projects feasible for many companies, rather than just a handful which are able to finance multi-billion dollar projects and handle the potential construction or licensing delays or cost overruns

Thermal-hydraulic code selection for modular high temperature gas-cooled reactors

Energy Technology Data Exchange (ETDEWEB)

Komen, E M.J.; Bogaard, J.P.A. van den

1995-06-01

In order to study the transient thermal-hydraulic system behaviour of modular high temperature gas-cooled reactors, the thermal-hydraulic computer codes RELAP5, MELCOR, THATCH, MORECA, and VSOP are considered at the Netherlands Energy Research Foundation ECN. This report presents the selection of the most appropriate codes. To cover the range of relevant accidents, a suite of three codes is recommended for analyses of HTR-M and MHTGR reactors. (orig.).

Pebble bed modular reactor safeguards: developing new approaches and implementing safeguards by design

Energy Technology Data Exchange (ETDEWEB)

Beyer, Brian David [Los Alamos National Laboratory; Beddingfield, David H [Los Alamos National Laboratory; Durst, Philip [INL; Bean, Robert [INL

2010-01-01

The design of the Pebble Bed Modular Reactor (PBMR) does not fit or seem appropriate to the IAEA safeguards approach under the categories of light water reactor (LWR), on-load refueled reactor (OLR, i.e. CANDU), or Other (prismatic HTGR) because the fuel is in a bulk form, rather than discrete items. Because the nuclear fuel is a collection of nuclear material inserted in tennis-ball sized spheres containing structural and moderating material and a PBMR core will contain a bulk load on the order of 500,000 spheres, it could be classified as a 'Bulk-Fuel Reactor.' Hence, the IAEA should develop unique safeguards criteria. In a multi-lab DOE study, it was found that an optimized blend of: (i) developing techniques to verify the plutonium content in spent fuel pebbles, (ii) improving burn-up computer codes for PBMR spent fuel to provide better understanding of the core and spent fuel makeup, and (iii) utilizing bulk verification techniques for PBMR spent fuel storage bins should be combined with the historic IAEA and South African approaches of containment and surveillance to verify and maintain continuity of knowledge of PBMR fuel. For all of these techniques to work the design of the reactor will need to accommodate safeguards and material accountancy measures to a far greater extent than has thus far been the case. The implementation of Safeguards-by-Design as the PBMR design progresses provides an approach to meets these safeguards and accountancy needs.

Technology development for the modular high-temperature gas-cooled reactor

International Nuclear Information System (INIS)

Homan, F.J.; Turner, R.F.

1989-01-01

In the USA the Modular High-Temperature Gas-Cooled Reactor is in an advanced stage of design. The related HTGR program areas, the approaches to these programs along with sample results and a description of how these data are used are highlighted in the paper. (author). Figs and tabs

Performance Estimation of Supercritical Co2 Micro Modular Reactor (MMR) for Varying Cooling Air Temperature

International Nuclear Information System (INIS)

Ahn, Yoonhan; Kim, Seong Gu; Cho, Seong Kuk; Lee, Jeong Ik

2015-01-01

A Small Modular Reactor (SMR) receives interests for the various application such as electricity co-generation, small-scale power generation, seawater desalination, district heating and propulsion. As a part of SMR development, supercritical CO2 Micro Modular Reactor (MMR) of 36.2MWth in power is under development by the KAIST research team. To enhance the mobility, the entire system including the power conversion system is designed for the full modularization. Based on the preliminary design, the thermal efficiency is 31.5% when CO2 is sufficiently cooled to the design temperature. A supercritical CO2 MMR is designed to supply electricity to the remote regions. The ambient temperature of the area can influence the compressor inlet temperature as the reactor is cooled with the atmospheric air. To estimate the S-CO2 cycle performance for various environmental conditions, A quasi-static analysis code is developed. For the off design performance of S-CO2 turbomachineries, the experimental result of Sandia National Lab (SNL) is utilized

A new small modular high-temperature gas-cooled reactor plant concept based on proven technology

International Nuclear Information System (INIS)

McDonald, C.F.; Goodjohn, A.J.

1982-01-01

Based on the established and proven high-temperature gas-cooled reactor (HTGR) technologies from the Peach Bottom 1 and Fort St. Vrain utility-operated units, a new small modular HTGR reactor is currently being evaluated. The basic nuclear reactor heat source, with a prismatic core, is being designed so that the decay heat can be removed by passive means (i.e., natural circulation). Although this concept is still in the preconceptual design stage, emphasis is being placed on establishing an inherently safe or benign concept which, when engineered, will have acceptable capital cost and power generation economics. The proposed new HTGR concept has a variety of applications, including electrical power generation, cogeneration, and high-temperature process heat. This paper discusses the simplest application, i.e., a steam Rankine cycle electrical power generating version. The gas-cooled modular reactor concepts presented are based on a graphite moderated prismatic core of low-power density (i.e., 4.1 W/cm 3 ) with a thermal rating of 250 MW(t). With the potential for inherently safe characteristics, a new small reactor could be sited close to industrial and urban areas to provide electrical power and thermal heating needs (i.e., district and space heating). Incorporating a multiplicity of small modular units to provide a larger power output is also discussed. The potential for a small, inherently safe HTGR reactor concept is highlighted

French Modular Impoundment: Final Cost and Performance Evaluation

Energy Technology Data Exchange (ETDEWEB)

Drown, Peter [French Development Enterprises, LLC, North Billerica, MA (United States); French, Bill [French Development Enterprises, LLC, North Billerica, MA (United States)

2017-05-17

This report comprises the Final Cost and Performance Report for the Department of Energy Award # EE0007244, the French Modular Impoundment (aka the “French Dam”.) The French Dam is a system of applying precast modular construction to water control structures. The “French Dam” is a term used to cover the construction means/methods used to construct or rehabilitate dams, diversion structures, powerhouses, and other hydraulic structures which impound water and are covered under FDE’s existing IP (Patents # US8414223B2; US9103084B2.)

A modular multi-microcomputer system for on-line vibration diagnostics

International Nuclear Information System (INIS)

Saedtler, E.

1988-01-01

A new modular multi-microprocessor system for on-line vibration monitoring and diagnostics of PWRs is described. The aim of the system is to make feasible an early detection of increasing failures in relevant regions of a reactor plant, to verify the mechanical integrity of the investigated components, and to improve therefore the operational safety of the plant. After a discussion of the implemented surveillance methods and algorithms, which are based on hierarchical structured identification (estimation) and statistical pattern recognition tools, the system architecture (software and hardware) is portrayed. The classification scheme itself works sequential so that samples (or features) can arrive on-line. This on-line classification is important in order to take necessary actions in time. Furthermore, the system has learning capabilities, which means it is adaptable to different, varying states and plant conditions. The main features of the system are presented and its contribution to an automation of complex surveillance and monitoring tasks is shown. (author)

Gas turbine modular helium reactor in cogeneration

International Nuclear Information System (INIS)

Leon de los Santos, G.

2009-10-01

This work carries out the thermal evaluation from the conversion of nuclear energy to electric power and process heat, through to implement an outline gas turbine modular helium reactor in cogeneration. Modeling and simulating with software Thermo flex of Thermo flow the performance parameters, based on a nuclear power plant constituted by an helium cooled reactor and helium gas turbine with three compression stages, two of inter cooling and one regeneration stage; more four heat recovery process, generating two pressure levels of overheat vapor, a pressure level of saturated vapor and one of hot water, with energetic characteristics to be able to give supply to a very wide gamma of industrial processes. Obtaining a relationship heat electricity of 0.52 and efficiency of net cogeneration of 54.28%, 70.2 MW net electric, 36.6 MW net thermal with 35% of condensed return to 30 C; for a supplied power by reactor of 196.7 MW; and with conditions in advanced gas turbine of 850 C and 7.06 Mpa, assembly in a shaft, inter cooling and heat recovery in cogeneration. (Author)

PCU arrangement of a supercritical CO{sub 2} cooled micro modular reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Seong Gu; Baik, Seungjoon; Cho, Seong Kuk; Oh, Bong Seong; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of)

2016-05-15

As part of the SMR(Small Modular Reactor)s development effort, the authors propose a concept of supercritical CO{sub 2} (S-CO{sub 2}) cooled fast reactor combined with the S-CO{sub 2} Brayton cycle. The reactor concept is named as KAIST Micro Modular Reactor (MMR). The S-CO{sub 2} Brayton cycle has many strong points when it is used for SMR's power conversion unit. It occupies small footprints due to the compact cycle components and simple layout. Thus, a concept of one module containing the S-CO{sub 2} cooled fast reactor and power conversion system is possible. This module can be shipped via ground transportation (by trailer) or marine transportation. In this study, the authors propose a new conceptual layout for the S-CO{sub 2} cooled direct cycle while considering various issues for arranging cycle components. The new design has an improved cycle efficiency (from 31% to 34%) than the earlier version of MMR by reducing pressure drops in the heat exchangers. As a more efficient option, a recompression recuperated cycle was also designed. It improves 5% of thermal efficiency while 18tons of mass can be added in comparison to the simple recuperated cycle. Even if we adopt recompression cycle as a PCU, the weight of module (152tons) is less than the ground transportable limit (260tons)

LOCA Analysis of KAIST-Micro Modular Reactor with Modified GAMMA+ code

Energy Technology Data Exchange (ETDEWEB)

Oh, Bong Seong; Ahn, Yoon Han; Kim, Seong Gu; Bae, Seong Jun; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of)

2016-05-15

The supercritical carbon dioxide (S-CO{sub 2}) power cycle is being seriously investigated around the world due to its simple layout, quite high efficiency around 500 .deg. C turbine inlet temperature, etc. By combining these two ideas, the KAIST research team developed a S-CO{sub 2} cooled SMR, called KAIST-Micro Modular reactor (MMR), which is targeting transportability and electricity supply for remote region. Therefore, requirements of MMR design are factory fabrication of the total system including power conversion system to be transported and air cooling to be independent from the site selection. Until now, steady performances and sizes of components were evaluated. Thus, in this paper a transient performance of the MMR are simulated with special focus on the loss of coolant accident (LOCA) at cold leg pipe. The MMR is a newly suggested innovative small modular reactor concept by the KAIST research team. Since the MMR is cooled by supercritical CO{sub 2}, general safety codes for conventional reactors have limitations. Thus, GAMMA+ code for the transient analysis of a gas-cooled reactor was selected and modified for the S-CO{sub 2} power system. After the modification of GAMMA+ code, LOCA is simulated, which is considered as one of the most limiting accidents in terms of safety of nuclear power plant.

Thorium-Based Fuel Cycles in the Modular High Temperature Reactor

Institute of Scientific and Technical Information of China (English)

CHANG Hong; YANG Yongwei; JING Xingqing; XU Yunlin

2006-01-01

Large stockpiles of civil-grade as well as weapons-grade plutonium have been accumulated in the world from nuclear power or other programs of different countries. One alternative for the management of the plutonium is to incinerate it in the high temperature reactor (HTR). The thorium-based fuel cycle was studied in the modular HTR to reduce weapons-grade plutonium stockpiles, while producing no additional plutonium or other transuranic elements. Three thorium-uranium fuel cycles were also investigated. The thorium absorption cross sections of the resolved and unresolved resonances were generated using the ZUT-DGL code based on existing resonance data. The equilibrium core of the modular HTR was calculated and analyzed by means of the code VSOP'94. The results show that the modular HTR can incinerate most of the initially loaded plutonium amounting to about 95.3% net 239Pu for weapons-grade plutonium and can effectively utilize the uranium and thorium in the thorium-uranium fuel cycles.

Water-ingress analysis for the 200 MWe pebble-bed modular high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Zheng Yanhua; Shi Lei; Wang Yan

2010-01-01

Water ingress into the primary circuit is generally recognized as one of the severe accidents with potential hazard to the modular high temperature gas-cooled reactor adopting steam-turbine cycle, which will cause a positive reactivity introduction, as well as the chemical corrosion of graphite fuel elements and reflector structure material. Besides, increase of the primary pressure may result in the opening of the safety valves, consequently leading the release of radioactive isotopes and flammable water gas. The analysis of such a kind of important and particular accident is significant to verify the inherent safety characteristics of the modular HTR plants. Based on the preliminary design of the 200 MWe high temperature gas-cooled reactor pebble-bed modular (HTR-PM), the design basis accident of a double-ended guillotine break of one heating tube and the beyond design basis accident of a large break of the main steam collection plate have been analyzed by using TINTE code, which is a special transient analysis program for high temperature gas-cooled reactors. Some safety relevant concerns, such as the fuel temperature, the primary loop pressure, the graphite corrosion, the water gas releasing amount, as well as the natural convection influence on the condition of failing to close the blower flaps, have been studied in detail. The calculation results indicate that even under some severe hypothetical postulates, the HTR-PM is able to keep the inherent safeties of the modular high temperature gas-cooled reactor and has a relatively good natural plant response, which will not result in environmental radiation hazard.

Small Modular Reactors: Institutional Assessment

Energy Technology Data Exchange (ETDEWEB)

Joseph Perkowski, Ph.D.

2012-06-01

? Objectives include, among others, a description of the basic development status of “small modular reactors” (SMRs) focused primarily on domestic activity; investigation of the domestic market appeal of modular reactors from the viewpoints of both key energy sector customers and also key stakeholders in the financial community; and consideration of how to proceed further with a pro-active "core group" of stakeholders substantially interested in modular nuclear deployment in order to provide the basis to expedite design/construction activity and regulatory approval. ? Information gathering was via available resources, both published and personal communications with key individual stakeholders; published information is limited to that already in public domain (no confidentiality); viewpoints from interviews are incorporated within. Discussions at both government-hosted and private-hosted SMR meetings are reflected herein. INL itself maintains a neutral view on all issues described. Note: as per prior discussion between INL and CAP, individual and highly knowledgeable senior-level stakeholders provided the bulk of insights herein, and the results of those interviews are the main source of the observations of this report. ? Attachment A is the list of individual stakeholders consulted to date, including some who provided significant earlier assessments of SMR institutional feasibility. ? Attachments B, C, and D are included to provide substantial context on the international status of SMR development; they are not intended to be comprehensive and are individualized due to the separate nature of the source materials. Attachment E is a summary of the DOE requirements for winning teams regarding the current SMR solicitation. Attachment F deserves separate consideration due to the relative maturity of the SMART SMR program underway in Korea. Attachment G provides illustrative SMR design features and is intended for background. Attachment H is included for overview

Proliferation resistance of small modular reactors fuels

Energy Technology Data Exchange (ETDEWEB)

Polidoro, F.; Parozzi, F. [RSE - Ricerca sul Sistema Energetico,Via Rubattino 54, 20134, Milano (Italy); Fassnacht, F.; Kuett, M.; Englert, M. [IANUS, Darmstadt University of Technology, Alexanderstr. 35, D-64283 Darmstadt (Germany)

2013-07-01

In this paper the proliferation resistance of different types of Small Modular Reactors (SMRs) has been examined and classified with criteria available in the literature. In the first part of the study, the level of proliferation attractiveness of traditional low-enriched UO{sub 2} and MOX fuels to be used in SMRs based on pressurized water technology has been analyzed. On the basis of numerical simulations both cores show significant proliferation risks. Although the MOX core is less proliferation prone in comparison to the UO{sub 2} core, it still can be highly attractive for diversion or undeclared production of nuclear material. In the second part of the paper, calculations to assess the proliferation attractiveness of fuel in typical small sodium cooled fast reactor show that proliferation risks from spent fuel cannot be neglected. The core contains a highly attractive plutonium composition during the whole life cycle. Despite some aspects of the design like the sealed core that enables easy detection of unauthorized withdrawal of fissile material and enhances proliferation resistance, in case of open Non-Proliferation Treaty break-out, weapon-grade plutonium in sufficient quantities could be extracted from the reactor core.

Integration of multi-interface conversion channel using FPGA for modular photonic network

Science.gov (United States)

Janicki, Tomasz; Pozniak, Krzysztof T.; Romaniuk, Ryszard S.

2010-09-01

The article discusses the integration of different types of interfaces with FPGA circuits using a reconfigurable communication platform. The solution has been implemented in practice in a single node of a distributed measurement system. Construction of communication platform has been presented with its selected hardware modules, described in VHDL and implemented in FPGA circuits. The graphical user interface (GUI) has been described that allows a user to control the operation of the system. In the final part of the article selected practical solutions have been introduced. The whole measurement system resides on multi-gigabit optical network. The optical network construction is highly modular, reconfigurable and scalable.

The Pebble Bed Modular Reactor: An obituary

Energy Technology Data Exchange (ETDEWEB)

Thomas, Steve, E-mail: stephen.thomas@gre.ac.u [Public Services International Research Unit (PSIRU), Business School, University of Greenwich, 30 Park Row, London SE10 9LS (United Kingdom)

2011-05-15

The High Temperature Gas-cooled Reactor (HTGR) has exerted a peculiar attraction over nuclear engineers. Despite many unsuccessful attempts over half a century to develop it as a commercial power reactor, there is still a strong belief amongst many nuclear advocates that a highly successful HTGR technology will emerge. The most recent attempt to commercialize an HTGR design, the Pebble Bed Modular Reactor (PBMR), was abandoned in 2010 after 12 years of effort and the expenditure of a large amount of South African public money. This article reviews this latest attempt to commercialize an HTGR design and attempts to identify which issues have led to its failure and what lessons can be learnt from this experience. It concludes that any further attempts to develop HTGRs using Pebble Bed technology should only be undertaken if there is a clear understanding of why earlier attempts have failed and a high level of confidence that earlier problems have been overcome. It argues that the PBMR project has exposed serious weaknesses in accountability mechanisms for the expenditure of South African public money. - Research highlights: {yields} In this study we examine the reasons behind the failure of the South African PBMR programme. {yields} The study reviews the technical issues that have arisen and lessons for future reactor developments. {yields} The study also identifies weaknesses in the accountability mechanisms for public spending.

Utility industry evaluation of the Modular High-Temperature Gas-Cooled Reactor

International Nuclear Information System (INIS)

Burstein, S.; Bitel, J.S.; Tramm, T.R.; High, M.D.; Neils, G.H.; Tomonto, J.R.; Weinberg, C.J.

1990-02-01

A team of utility industry representatives evaluated the Modular High Temperature Gas-Cooled Reactor plant design, a current design created by an industrial team led by General Atomics under Department of Energy sponsorship and with support provided by utilities through Gas-Cooled Reactor Associates. The utility industry team concluded that the plant design should be considered a viable application of an advanced nuclear concept and deserves continuing development. Specific comments and recommendations are provided as a contribution toward improving a very promising plant design. 2 refs

Importance of the multi-modules study in PSA; Importancia del estudio de multi-modulos en APS

Energy Technology Data Exchange (ETDEWEB)

Gonzalez R, V. J.; Nelson E, P. F., E-mail: judith_gonzalez_rodriguez@outlook.es [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)

2015-09-15

The current approach that has taken the Probabilistic Safety Analysis (PSA) consists of doing all the APS analysis including the existence of multi-units in the nuclear power plants (NPP), this new approach seeks to analyze the risk of site, evaluating all reactors together. The main reasons for this trend are: the accident occurred on March 2011 in Fukushima Daiichi in Japan, with serious consequences in more than one reactor of the NPP and the current planning and construction of new Small Modular Reactors, which host more than one module on the same NPP and are connected to a single control room. This study analyzes how to model the risk of a multi-module NPP. In 2013, the ASME/ANS standard for advanced reactors that are not light-water reactors was published, in which the requirements to realize a PSA including multi-units or modules are shown; however, does not describe the methodology to do that. This article presents a methodology to calculate the risk of the site in a PBMR plant with two modules. This methodology consists of two models of trees of different events, one that evaluates to a single PBMR module and another that evaluates the two modules together. Both models are responsible to show their differences and compare results to finally demonstrate the need for new methodologies for risk analysis site in multi-modules and units. (Author)

Proposed Advanced Reactor Adaptation of the Standard Review Plan NUREG-0800 Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors

Energy Technology Data Exchange (ETDEWEB)

Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Poore, III, Willis P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holbrook, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States); Moe, Wayne [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sofu, Tanju [Argonne National Lab. (ANL), Argonne, IL (United States)

2017-03-01

This report proposes adaptation of the previous regulatory gap analysis in Chapter 4 (Reactor) of NUREG 0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. The proposed adaptation would result in a Chapter 4 review plan applicable to certain advanced reactors. This report addresses two technologies: the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for adaptation because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-based description in Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail.

A small-scale modular reactor for electric source for remote places

International Nuclear Information System (INIS)

Anon.

2002-01-01

Use of a small-scale modular reactor (SMR) as an electric source for remote places is one of scenarios for actual use of SMR parallel to alternative source of present nuclear power stations and co-generation source at urban suburbs, there is not only an actual experience to construct and operate for power source for military use in U.S.A. on 1950s to 1960s, but also four nuclear reactors (LWGR, 12 MW) in Vilyvino Nuclear Power Station in far northern district in Russia are under operation. Recently, Department of Energy in U.S.A. prepared the 'Report to Congress on Small Modular Nuclear Reactors' evaluating on feasibility of SMR as a power source for remote places according to requirement of the Congress. This report evaluated a feasibility study on nine SMRs in the world with 10 to 50 MW of output as electric source for remote places on economical efficiency and so on, together with analysis of their design concepts, to conclude that 'they could perform beginning of operations on 2000s because of no large technical problems and keeping a level capable of competing with power generation cost at remote place on its present economical efficiency'. Here was introduced on outlines of this report. (G.K.)

Pebble bed modular reactor - The first Generation IV reactor to be constructed

International Nuclear Information System (INIS)

Ion, S.; Nicholls, D.; Matzie, R.; Matzner, D.

2004-01-01

Substantial interest has been generated in advanced reactors over the past few years. This interest is motivated by the view that new nuclear power reactors will be needed to provide low carbon generation of electricity and possibly hydrogen to support the future growth in demand for both of these commodities. Some governments feel that substantially different designs will be needed to satisfy the desires for public perception, improved safety, proliferation resistance, reduced waste and competitive economics. This has motivated the creation of the Generation IV Nuclear Energy Systems programme in which ten countries have agreed on a framework for international cooperation in research for advanced reactors. Six designs have been selected for continued evaluation, with the objective of deployment by 2030. One of these designs is the very high temperature reactor (VHTR), which is a thermal neutron spectrum system with a helium-cooled core utilising carbon-based fuel. The pebble bed modular reactor (PBMR), being developed in South Africa through a worldwide international collaborative effort led by Eskom, the national utility, will represent a key milestone on the way to achievement of the VHTR design objectives, but in the much nearer term. This paper outlines the design objectives, safety approach and design details of the PBMR, which is already at a very advanced stage of development. (author)

Passive Decay Heat Removal System for Micro Modular Reactor

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

Licensing process characteristics of Small Modular Reactors and spent nuclear fuel repository

Energy Technology Data Exchange (ETDEWEB)

Söderholm, Kristiina, E-mail: kristiina.soderholm@fortum.com [Fortum Power (Finland); Tuunanen, Jari, E-mail: jari.tuunanen@fortum.com [Fortum Power (Finland); Amaba, Ben, E-mail: baamaba@us.ibm.com [IBM Complex Systems (United States); Bergqvist, Sofia, E-mail: sofia.bergqvist@se.ibm.com [IBM Rational Software (Sweden); Lusardi, Paul, E-mail: plusardi@nuscalepower.com [NuScale Power (United States)

2014-09-15

Highlights: • We examine the licensing process challenges of modular nuclear facilities. • We compare the features of Small Modular Reactors and spent nuclear fuel repository. • We present the need of nuclear licensing simplification. • Part of the licensing is proposed to be internationally applicable. • Systems engineering and requirements engineering benefits are presented. - Abstract: This paper aims to increase the understanding of the licensing processes characteristics of Small Modular Reactors (SMR) compared with licensing of spent nuclear fuel repository. The basis of the SMR licensing process development lies in licensing processes used in Finland, France, the UK, Canada and the USA. These countries have been selected for this study because of their various licensing processes and recent actions in the new NPP construction. Certain aspects of the aviation industry licensing process have also been studied and selected practices have been investigated as possibly suitable for use in nuclear licensing. Suitable features for SMR licensing are emphasized and suggested. The licensing features of the spent nuclear fuel deep repository along with similar features of SMR licensing are discussed. Since there are similar types of challenges of lengthy licensing time frames, as well as modular features to be taken into account in licensing, these two different nuclear industry fields can be compared. The main SMR features to take into account in licensing are: • Standardization of the design. • Modularity. • Mass production. • Serial construction. Modularity can be divided into two different categories: the first category is simply a single power plant unit constructed of independently engineered modules (e.g. construction process for Westinghouse AP-1000 NPP) and the second one a power plant composed of many reactor modules, which are manufactured in factories and installed as needed (e.g. NuScale Power SMR design). The deep underground repository

Optimization of coupled multiphysics methodology for safety analysis of pebble bed modular reactor

Science.gov (United States)

Mkhabela, Peter Tshepo

The research conducted within the framework of this PhD thesis is devoted to the high-fidelity multi-physics (based on neutronics/thermal-hydraulics coupling) analysis of Pebble Bed Modular Reactor (PBMR), which is a High Temperature Reactor (HTR). The Next Generation Nuclear Plant (NGNP) will be a HTR design. The core design and safety analysis methods are considerably less developed and mature for HTR analysis than those currently used for Light Water Reactors (LWRs). Compared to LWRs, the HTR transient analysis is more demanding since it requires proper treatment of both slower and much longer transients (of time scale in hours and days) and fast and short transients (of time scale in minutes and seconds). There is limited operation and experimental data available for HTRs for validation of coupled multi-physics methodologies. This PhD work developed and verified reliable high fidelity coupled multi-physics models subsequently implemented in robust, efficient, and accurate computational tools to analyse the neutronics and thermal-hydraulic behaviour for design optimization and safety evaluation of PBMR concept The study provided a contribution to a greater accuracy of neutronics calculations by including the feedback from thermal hydraulics driven temperature calculation and various multi-physics effects that can influence it. Consideration of the feedback due to the influence of leakage was taken into account by development and implementation of improved buckling feedback models. Modifications were made in the calculation procedure to ensure that the xenon depletion models were accurate for proper interpolation from cross section tables. To achieve this, the NEM/THERMIX coupled code system was developed to create the system that is efficient and stable over the duration of transient calculations that last over several tens of hours. Another achievement of the PhD thesis was development and demonstration of full-physics, three-dimensional safety analysis

Westinghouse Small Modular Reactor nuclear steam supply system design

Energy Technology Data Exchange (ETDEWEB)

Memmott, M. J.; Harkness, A. W.; Van Wyk, J. [Westinghouse Electric Company LLC, 600 Cranberry Woods Drive, Cranberry Twp. PA 16066 (United States)

2012-07-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the first in a series of four papers which describe the design and functionality of the Westinghouse SMR. Also described in this series are the key drivers influencing the design of the Westinghouse SMR and the unique passive safety features of the Westinghouse SMR. Several critical motivators contributed to the development and integration of the Westinghouse SMR design. These design driving motivators dictated the final configuration of the Westinghouse SMR to varying degrees, depending on the specific features under consideration. These design drivers include safety, economics, AP1000{sup R} reactor expertise and experience, research and development requirements, functionality of systems and components, size of the systems and vessels, simplicity of design, and licensing requirements. The Westinghouse SMR NSSS consists of an integral reactor vessel within a compact containment vessel. The core is located in the bottom of the reactor vessel and is composed of 89 modified Westinghouse 17x17 Robust Fuel Assemblies (RFA). These modified fuel assemblies have an active core length of only 2.4 m (8 ft) long, and the entirety of the core is encompassed by a radial reflector. The Westinghouse SMR core operates on a 24 month fuel cycle. The reactor vessel is approximately 24.4 m (80 ft) long and 3.7 m (12 ft) in diameter in order to facilitate standard rail shipping to the site. The reactor vessel houses hot and cold leg channels to facilitate coolant flow, control rod drive mechanisms (CRDM), instrumentation and cabling, an intermediate flange to separate flow and instrumentation and facilitate simpler refueling, a pressurizer, a straight tube, recirculating steam

The HTR modular power reactor system. Qualification of fuel elements and materials

International Nuclear Information System (INIS)

Heidenreich, U.; Breitling, H.; Nieder, R.; Ohly, W.; Mittenkuehler, A.; Ragoss, H.; Seehafer, H.J.; Wirtz, K.; Serafin, N.

1989-01-01

For further development of the HTR modular power reactor system (HTR-M-KW), the project activities for 'Qualification of fuel elements and materials' reported here cover the work for specifying the qualifications to be met by metallic and ceramic materials, taking into account the design-based requirements and the engineered safety requirements. The fission product retention data determined for the HTR modular reactor fuel elements could be better confirmed by evaluation of the experiments, and have been verified by various calculation methods for different operating conditions. The qualification of components was verified by strength analyses including a benchmark calculation for specified normal operation and emergencies; the results show a convenient behaviour of the components and their materials. In addition, a fuel element burnup measuring system was designed that applies Cs-137 gamma spectroscopy; its feasibility was checked by appropriate analyses, and qualification work is in progress. The installation of a prototype measurement system is the task for project No. 03 IAT 211. (orig.) [de

The modular high-temperature gas-cooled reactor - a new production reactor

International Nuclear Information System (INIS)

Nulton, J.D.

1990-01-01

One of the reactor concepts being considered for application as a new production reactor (NPR) is a 350-MW(thermal) modular high-temperature gas-cooled reactor (MHTGR). The proposed MHTGR-NPR is based on the design of the commercial MHTGR and is being developed by a team that includes General Atomics and Combustion Engineering. The proposed design includes four modules combined into a production block that includes a shared containment, a spent-fuel storage facility, and other support facilities. The MHTGR has a helium-cooled, graphite-moderated, graphite-reflected annular core formed from prismatic graphite fuel blocks. The MHTGR fuel consists of highly enriched uranium oxycarbide (UCO) microsphere fuel particles that are coated with successive layers of pyrolytic carbon (PyC) and silicon carbide (SiC). Tritium-producing targets consist of enriched 6 Li aluminate microsphere target particles that are coated with successive layers of PyC and SiC similar to the fuel microspheres. Normal reactivity control is implemented by articulated control rods that can be inserted into channels in the inner and outer reflector blocks. Shutdown heat removal is accomplished by a single shutdown heat exchanger and electric motor-driven circulator located in the bottom of the reactor vessel. Current plans are to stack spent fuel elements in dry, helium-filled, water-cooled wells and store them for ∼1 yr before reprocessing. All phases of MHTGR fuel reprocessing have been demonstrated

Draft pre-application safety evaluation report for the modular high-temperature gas-cooled reactor

International Nuclear Information System (INIS)

Williams, P.M.; King, T.L.; Wilson, J.N.

1989-03-01

This draft safety evaluation report (SER) presents the preliminary results of a pre-application design review for the standard modular high-temperature gas-cooled reactor (MHTGR) (Project 672). The MHTGR conceptual design was submitted by the U.S. Department of Energy (DOE) in accordance with the U.S. Nuclear Regulatory Commission(NRC) 'Statement of Policy for the Regulation of Advanced Nuclear Power Plants' (51 FR 24643), which provides for early Commission review and interaction. The standard MHTGR consists of four identical reactor modules, each with a thermal output of 350 MWt, coupled with two steam turbine-generator sets to produce a total plant electrical output of 540 MWe. The reactors are helium cooled and graphite moderated and utilize ceramically coated particle-type nuclear fuel. The design includes passive reactor-shutdown and decay-heat-removal features. The staff and its contractors at the Oak Ridge National Laboratory and the Brookhaven National Laboratory have reviewed this design with emphasis on those unique provisions in the design that accomplish the key safety functions of reactor shutdown, decay-heat removal, and containment of radioactive material. This report presents the NRC staff's technical evaluation of those features in the MHTGR design important to safety, including their proposed research and testing needs. In addition this report presents the criteria proposed by the NRC staff to judge the acceptability of the MHTGR design and, where possible, includes statements on the potential of the MHTGR to meet these criteria. However, it should be recognized that final conclusions in all matters discussed in this report require approval by the Commission. Final determination on the acceptability of the MHTGR standard design is contingent on receipt and evaluation of additional information requested from DOE pertaining to the adequacy of the containment design and on the following: (1) satisfactory resolution of open safety issues identified

Small modular reactor: First-of-a-Kind (FOAK) and Nth-of-a-Kind (NOAK) Economic Analysis

Energy Technology Data Exchange (ETDEWEB)

Boldon, Lauren M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sabharwall, Piyush [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2014-08-01

Small modular reactors (SMRs) refer to any reactor design in which the electricity generated is less than 300 MWe. Often medium sized reactors with power less than 700 MWe are also grouped into this category. Internationally, the development of a variety of designs for SMRs is booming with many designs approaching maturity and even in or nearing the licensing stage. It is for this reason that a generalized yet comprehensive economic model for first of a kind (FOAK) through nth of a kind (NOAK) SMRs based upon rated power, plant configuration, and the fiscal environment was developed. In the model, a particular project’s feasibility is assessed with regards to market conditions and by commonly utilized capital budgeting techniques, such as the net present value (NPV), internal rate of return (IRR), Payback, and more importantly, the levelized cost of energy (LCOE) for comparison to other energy production technologies. Finally, a sensitivity analysis was performed to determine the effects of changing debt, equity, interest rate, and conditions on the LCOE.

Hydrogen generation using the modular helium reactor

International Nuclear Information System (INIS)

Richards, M.; Shenoy, A.

2004-01-01

Process heat from a high-temperature nuclear reactor can be used to drive a set of chemical reactions, with the net result of splitting water into hydrogen and oxygen. For example, process heat at temperatures in the range 850 deg.C to 950 deg.C can drive the sulfur-iodine (SI) thermochemical process to produce hydrogen with high efficiency. Electricity can also be used to split water, using conventional, low-temperature electrolysis. An example of a hybrid process is high-temperature electrolysis (HTE), in which process heat is used to generate steam, which is then supplied to an electrolyser to generate hydrogen. In this paper we investigate the coupling of the Modular Helium Reactor (MHR) to the SI process and HTE. These concepts are referred to as the H2-MHR. Optimization of the MHR core design to produce higher coolant outlet temperatures is also discussed. The use of fixed orifices to control the flow distribution is a promising design solution for increasing the coolant outlet temperature without increasing peak fuel temperatures significantly

Small modular reactors: current status, economic aspects and licensing

International Nuclear Information System (INIS)

Zimbron E, E.; Puente E, F.

2014-10-01

Interest for nuclear energy had resurgence since the beginning of the new century. This was a consequence of the new world conditions and needs: increasing energy demands (mainly from developing countries), awareness of the importance of energetic security and the necessity of limiting carbon emissions. In this nuclear boom, Small Modular Reactors (SMRs) develop and start to consolidate as a viable option for the present energy market. Their modular characteristics, lower initial capital cost, passive safety features and their niche applications, situate them as a technology with various advantages. The following study will present and analysis that will help to comprehend the SMRs present status. Information will show planned reactors, reactors in construction and in operation, advantages and challenges of their implementation, relevant economic aspects and important licensing factors that need to be highlighted. The analysis showed that the SMR technology is still in an initial stage that could reach and important development point in the next ten years. In this period, many of the reactors that are in design stage or that are through their licensing process might be constructed and could be getting ready for a commercial status. On the other hand, it has been observed that there are two main economic factors that need to be considered for any SMRs implementation project. First, the costs (initial, operation, maintenance, fuel and decommissioning) and second their possible niche market applications. Additionally, it has been noted that the licensing process is one of the greatest challenges for SMR general development. Licensing is mainly related to topic such as Emergency Planning Zone, first-of-a-kind engineering, passive safety features, proliferation resistance, multiple module designs and staffing. Previous information will serve as a base for carrying out a feasibility assessment analysis for SMR in Mexico. This part will be the last section of the project

Small modular reactors: current status, economic aspects and licensing

Energy Technology Data Exchange (ETDEWEB)

Zimbron E, E. [Instituto Tecnologico de Estudios Superiores de Monterrey, Campus Santa Fe, Av. Carlos Lazo No. 100, Santa Fe, 01389 Mexico D. F. (Mexico); Puente E, F., E-mail: erick.zimbron@gmail.com [ININ, Direccion de Investigacion Cientifica, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2014-10-15

Interest for nuclear energy had resurgence since the beginning of the new century. This was a consequence of the new world conditions and needs: increasing energy demands (mainly from developing countries), awareness of the importance of energetic security and the necessity of limiting carbon emissions. In this nuclear boom, Small Modular Reactors (SMRs) develop and start to consolidate as a viable option for the present energy market. Their modular characteristics, lower initial capital cost, passive safety features and their niche applications, situate them as a technology with various advantages. The following study will present and analysis that will help to comprehend the SMRs present status. Information will show planned reactors, reactors in construction and in operation, advantages and challenges of their implementation, relevant economic aspects and important licensing factors that need to be highlighted. The analysis showed that the SMR technology is still in an initial stage that could reach and important development point in the next ten years. In this period, many of the reactors that are in design stage or that are through their licensing process might be constructed and could be getting ready for a commercial status. On the other hand, it has been observed that there are two main economic factors that need to be considered for any SMRs implementation project. First, the costs (initial, operation, maintenance, fuel and decommissioning) and second their possible niche market applications. Additionally, it has been noted that the licensing process is one of the greatest challenges for SMR general development. Licensing is mainly related to topic such as Emergency Planning Zone, first-of-a-kind engineering, passive safety features, proliferation resistance, multiple module designs and staffing. Previous information will serve as a base for carrying out a feasibility assessment analysis for SMR in Mexico. This part will be the last section of the project

Probabilistic safety assessment framework of pebble-bed modular high-temperature gas-cooled reactor

International Nuclear Information System (INIS)

Liu Tao; Tong Jiejuan; Zhao Jun; Cao Jianzhu; Zhang Liguo

2009-01-01

After an investigation of similar reactor type probabilistic safety assessment (PSA) framework, Pebble-bed Modular High-Temperature Gas-cooled Reactor (HTR-PM) PSA framework was presented in correlate with its own design characteristics. That is an integral framework which spreads through event sequence structure with initiating events at the beginning and source term categories in the end. The analysis shows that it is HTR-PM design feature that determines its PSA framework. (authors)

Systems analysis of a 100-MWe modular liquid metal cooled reactor

International Nuclear Information System (INIS)

Morris, E.E.; Rhow, S.K.; Switick, D.M.

1985-01-01

The response of a 100-MWe modular liquid metal cooled reactor to unprotected loss of flow and/or loss of primary heat removal accidents is analyzed using the systems analysis code SASSYS. The reactor response is tracked for the first 1000 s following a postulated upset in the primary heat removal system. The calculations do not take credit for the functioning of any decay heat removal other than through the secondary system. In addition to the power rating, other features of the reactor are an average sodium temperature rise of 148 K, a sodium void worth (counting the core and upper axial blanket) of 1.89 $, and 3.6 $ of Doppler feedback due to a uniform e-fold fuel temperature increase

An Artificial Neural Network Compensated Output Feedback Power-Level Control for Modular High Temperature Gas-Cooled Reactors

Directory of Open Access Journals (Sweden)

Zhe Dong

2014-02-01

Full Text Available Small modular reactors (SMRs could be beneficial in providing electricity power safely and also be viable for applications such as seawater desalination and heat production. Due to its inherent safety features, the modular high temperature gas-cooled reactor (MHTGR has been seen as one of the best candidates for building SMR-based nuclear power plants. Since the MHTGR dynamics display high nonlinearity and parameter uncertainty, it is necessary to develop a nonlinear adaptive power-level control law which is not only beneficial to the safe, stable, efficient and autonomous operation of the MHTGR, but also easy to implement practically. In this paper, based on the concept of shifted-ectropy and the physically-based control design approach, it is proved theoretically that the simple proportional-differential (PD output-feedback power-level control can provide asymptotic closed-loop stability. Then, based on the strong approximation capability of the multi-layer perceptron (MLP artificial neural network (ANN, a compensator is established to suppress the negative influence caused by system parameter uncertainty. It is also proved that the MLP-compensated PD power-level control law constituted by an experientially-tuned PD regulator and this MLP-based compensator can guarantee bounded closed-loop stability. Numerical simulation results not only verify the theoretical results, but also illustrate the high performance of this MLP-compensated PD power-level controller in suppressing the oscillation of process variables caused by system parameter uncertainty.

Baseline Concept Description of a Small Modular High Temperature Reactor

Energy Technology Data Exchange (ETDEWEB)

Gougar, Hans D. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2014-10-01

The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNP were all both ‘small’ or medium-sized and ‘modular’ by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOE’s ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the

Baseline Concept Description of a Small Modular High Temperature Reactor

Energy Technology Data Exchange (ETDEWEB)

Hans Gougar

2014-05-01

The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNP were all both ‘small’ or medium-sized and ‘modular’ by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOE’s ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the

Classification of functional interactions from multi-electrodes data using conditional modularity analysis

Science.gov (United States)

Makhtar, Siti Noormiza; Senik, Mohd Harizal

2018-02-01

The availability of massive amount of neuronal signals are attracting widespread interest in functional connectivity analysis. Functional interactions estimated by multivariate partial coherence analysis in the frequency domain represent the connectivity strength in this study. Modularity is a network measure for the detection of community structure in network analysis. The discovery of community structure for the functional neuronal network was implemented on multi-electrode array (MEA) signals recorded from hippocampal regions in isoflurane-anaesthetized Lister-hooded rats. The analysis is expected to show modularity changes before and after local unilateral kainic acid (KA)-induced epileptiform activity. The result is presented using color-coded graphic of conditional modularity measure for 19 MEA nodes. This network is separated into four sub-regions to show the community detection within each sub-region. The results show that classification of neuronal signals into the inter- and intra-modular nodes is feasible using conditional modularity analysis. Estimation of segregation properties using conditional modularity analysis may provide further information about functional connectivity from MEA data.

Modeling and performance of the MHTGR [Modular High-Temperature Gas-Cooled Reactor] reactor cavity cooling system

International Nuclear Information System (INIS)

Conklin, J.C.

1990-04-01

The Reactor Cavity Cooling System (RCCS) of the Modular High- Temperature Gas-Cooled Reactor (MHTGR) proposed by the U.S. Department of Energy is designed to remove the nuclear afterheat passively in the event that neither the heat transport system nor the shutdown cooling circulator subsystem is available. A computer dynamic simulation for the physical and mathematical modeling of and RCCS is described here. Two conclusions can be made form computations performed under the assumption of a uniform reactor vessel temperature. First, the heat transferred across the annulus from the reactor vessel and then to ambient conditions is very dependent on the surface emissivities of the reactor vessel and RCCS panels. These emissivities should be periodically checked to ensure the safety function of the RCCS. Second, the heat transfer from the reactor vessel is reduced by a maximum of 10% by the presence of steam at 1 atm in the reactor cavity annulus for an assumed constant in the transmission of radiant energy across the annulus can be expected to result in an increase in the reactor vessel temperature for the MHTGR. Further investigation of participating radiation media, including small particles, in the reactor cavity annulus is warranted. 26 refs., 7 figs., 1 tab

KWU's modular approach to HTR commercialization

International Nuclear Information System (INIS)

Frewer, H.; Weisbrodt, I.

1983-01-01

As a way of avoiding the uncertainties, delays and unacceptable commercial risks which have plagued advanced reactor projects in Germany, KWU is advocating a modular approach to commercialization of the high-temperature reactor (HTR), using small size standard reactor units. KWU has received a contract for the study of a co-generation plant based on this modular system. Features of the KWU modular HTR, process heat, gasification, costs and future development are discussed. (UK)

Safety aspects of the Modular High-Temperature Gas-Cooled Reactor (MHTGR)

International Nuclear Information System (INIS)

Silady, F.A.; Millunzi, A.C.

1989-08-01

The Modular High-Temperature Gas-Cooled Reactor (MHTGR) is an advanced reactor concept under development through a cooperative program involving the US Government, the nuclear industry and the utilities. The design utilizes the basic high-temperature gas-cooled reactor (HTGR) features of ceramic fuel, helium coolant, and a graphite moderator. The qualitative top-level safety requirement is that the plant's operation not disturb the normal day-to-day activities of the public. The MHTGR safety response to events challenging the functions relied on to retain radionuclides within the coated fuel particles has been evaluated. A broad range of challenges to core heat removal have been examined which include a loss of helium pressure and a simultaneous loss of forced cooling of the core. The challenges to control of heat generation have considered not only the failure to insert the reactivity control systems, but the withdrawal of control rods. Finally, challenges to control chemical attack of the ceramic coated fuel have been considered, including catastrophic failure of the steam generator allowing water ingress or of the pressure vessels allowing air ingress. The plant's response to these extreme challenges is not dependent on operator action and the events considered encompass conceivable operator errors. In the same vein, reliance on radionuclide retention within the full particle and on passive features to perform a few key functions to maintain the fuel within acceptable conditions also reduced susceptibility to external events, site-specific events, and to acts of sabotage and terrorism. 4 refs., 14 figs., 1 tab

Update on Small Modular Reactors Dynamics System Modeling Tool -- Molten Salt Cooled Architecture

Energy Technology Data Exchange (ETDEWEB)

Hale, Richard Edward [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Cetiner, Sacit M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Fugate, David L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Qualls, A L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Borum, Robert C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Chaleff, Ethan S. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Rogerson, Doug W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Batteh, John J. [Modelon Corporation (Sweden); Tiller, Michael M. [Xogeny Corporation, Canton, MI (United States)

2014-08-01

The Small Modular Reactor (SMR) Dynamic System Modeling Tool project is in the third year of development. The project is designed to support collaborative modeling and study of various advanced SMR (non-light water cooled) concepts, including the use of multiple coupled reactors at a single site. The objective of the project is to provide a common simulation environment and baseline modeling resources to facilitate rapid development of dynamic advanced reactor SMR models, ensure consistency among research products within the Instrumentation, Controls, and Human-Machine Interface (ICHMI) technical area, and leverage cross-cutting capabilities while minimizing duplication of effort. The combined simulation environment and suite of models are identified as the Modular Dynamic SIMulation (MoDSIM) tool. The critical elements of this effort include (1) defining a standardized, common simulation environment that can be applied throughout the program, (2) developing a library of baseline component modules that can be assembled into full plant models using existing geometry and thermal-hydraulic data, (3) defining modeling conventions for interconnecting component models, and (4) establishing user interfaces and support tools to facilitate simulation development (i.e., configuration and parameterization), execution, and results display and capture.

Importance of the multi-modules study in PSA

International Nuclear Information System (INIS)

Gonzalez R, V. J.; Nelson E, P. F.

2015-09-01

The current approach that has taken the Probabilistic Safety Analysis (PSA) consists of doing all the APS analysis including the existence of multi-units in the nuclear power plants (NPP), this new approach seeks to analyze the risk of site, evaluating all reactors together. The main reasons for this trend are: the accident occurred on March 2011 in Fukushima Daiichi in Japan, with serious consequences in more than one reactor of the NPP and the current planning and construction of new Small Modular Reactors, which host more than one module on the same NPP and are connected to a single control room. This study analyzes how to model the risk of a multi-module NPP. In 2013, the ASME/ANS standard for advanced reactors that are not light-water reactors was published, in which the requirements to realize a PSA including multi-units or modules are shown; however, does not describe the methodology to do that. This article presents a methodology to calculate the risk of the site in a PBMR plant with two modules. This methodology consists of two models of trees of different events, one that evaluates to a single PBMR module and another that evaluates the two modules together. Both models are responsible to show their differences and compare results to finally demonstrate the need for new methodologies for risk analysis site in multi-modules and units. (Author)

The gas turbine-modular helium reactor (GT-MHR), high efficiency, cost competitive, nuclear energy for the next century

International Nuclear Information System (INIS)

Zgliczynski, J.B.; Silady, F.A.; Neylan, A.J.

1994-04-01

The Gas Turbine-Modular Helium Reactor (GT-MHR) is the result of coupling the evolution of a small passively safe reactor with key technology developments in the US during the last decade: large industrial gas turbines, large active magnetic bearings, and compact, highly effective plate-fin heat exchangers. The GT-MHR is the only reactor concept which provides a step increase in economic performance combined with increased safety. This is accomplished through its unique utilization of the Brayton cycle to produce electricity directly with the high temperature helium primary coolant from the reactor directly driving the gas turbine electrical generator. This cannot be accomplished with another reactor concept. It retains the high levels of passive safety and the standardized modular design of the steam cycle MHTGR, while showing promise for a significant reduction in power generating costs by increasing plant net efficiency to a remarkable 47%

Application of S-CO_2 Cycle for Small Modular Reactor coupled with Desalination System

International Nuclear Information System (INIS)

Lee, Won Woong; Bae, Seong Jun; Lee, Jeong Ik

2016-01-01

The Korean small modular reactor, SMART (System-integrated Modular Advanced ReacTor, 100MWe), is designed to achieve enhanced safety and improved economics through reliable passive safety systems, a system simplification and component modularization. SMART can generate electricity and provide water by seawater desalination. However, due to the desalination aspect of SMART, the total amount of net electricity generation is decreased from 100MWe to 90MWe. The authors suggest in this presentation that the reduction of electricity generation can be replenished by applying S-CO_2 power cycle technology. The S-CO_2 Brayton cycle, which is recently receiving significant attention as the next generation power conversion system, has some benefits such as high cycle efficiency, simple configuration, compactness and so on. In this study, the cycle performance analysis of the S-CO_2 cycles for SMART with desalination system is conducted. The simple recuperated S-CO_2 cycle is revised for coupling with desalination system. The three revised layout are proposed for the cycle performance comparison. In this results of the 3rd revised layout, the cycle efficiency reached 37.8%, which is higher than the efficiency of current SMART with the conventional power conversion system 30%

The modular high-temperature gas-cooled reactor (MHTGR) in the US

International Nuclear Information System (INIS)

Neylan, A.J.; Graf, D.F.; Millunzi, A.C.

1987-01-01

GA Technologies Inc. and other U.S. corporations, in a cooperative program with the U.S. Department of Energy, is developing a Modular High-Temperature Gas-Cooled Reactor (MHTGR) that will provide highly reliable, economic, nuclear power. The MHTGR system assures maximum safety to the public, the owner/operator, and the environment. The MHTGR is being designed to meet and exceed rigorous requirements established by the user industry for availability, operation and maintenance, plant investment protection, safety and licensing, siting flexibility and economics. The plant will be equally attractive for deployment and operation in the U.S., other major industrialized nations including Korea, Japan, and the Republic of China, as well as the developing nations. The High-Temperature Gas-Cooled Reactor (HTGR) is an advanced, third generation nuclear power system which incorporates distinctive technical features, including the use of pressurized helium as a coolant, graphite as the moderator and core structural material, and fuel in the form of ceramic coated uranium particles. The modular HTGR builds upon generic gas-cooled reactor experience and specific HTGR programs and projects. The MHTGR offers unique technological features and the opportunity for the cooperative international development of an advanced energy system that will help assure adaquate world energy resources for the future. Such international joint venturing of energy development can offer significant benefits to participating industries and governments and also provides a long term solution to the complex problems of the international balance of payments

The design features of integrated modular water reactor (IMR)

International Nuclear Information System (INIS)

Kanagawa, T.; Goto, M.; Usui, S.; Suzuta, T.; Serizawa, A.; Kunugi, T.; Yamauchi, T.; Itoh, G.; Matsumura, T.

2004-01-01

Small-to-medium-sized (300-600 MWe) reactors are required for the electric power market in the near future (2010-2030). The main theme in the development of small-to-medium-sized reactor is how to realize competitive cost against other energy sources. As measures to this disadvantage, greatly simplified and small-scale design is needed. From such point of view, Integrated Modular Water Reactor (IMR), whose electric output power is 350 MWe, adopts integrated and high temperature two-phase natural circulation system for the primary system. In this design, main coolant pipes, a pressurizer, and reactor coolant pumps are not needed, and the sizes of the reactor vessel and steam generators are minimized. Additionally, to enhance the economy of the whole plant, fluid systems, and Instrumentation and Control systems of IMR have also been reviewed to make them simplest and smallest taking the advantage of the IMR concept and the state of the art technologies. For example, the integrated primary system and the stand-alone direct heat removal system make the safety system very simple, i.e., no injection, no containment spray, no emergency AC power, etc. The chemical and volume control system is also simplified by eliminating the boron control system and the seal water system of reactor coolant pumps. In this paper, the status of the IMR development and the outline of the IMR design efforts to achieve the simplest and smallest plant are presented. (authors)

Safety and international development of small modular reactors (SMR). A study of GRS

Energy Technology Data Exchange (ETDEWEB)

Buchholz, Sebastian; Kruessenberg, Anne; Schaffrath, Andreas [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Garching (Germany). Bereich Reaktorsicherheitsforschung

2015-11-15

The abbreviation SMR stands for Small Modular Reactor and describes reactors with low power output. One reactor module, composed of primary, secondary and, where necessary, intermediate circuit and auxiliary systems, may be transported to the construction site as a whole or in few parts only and can therefore be connected quickly to the grid. Various modules can form a larger nuclear power plant and additional modules may be added one by one, while the others are in operation. Designers develop SMR for the deployment mainly in remote, sparsely populated areas or near cities respectively. SMR may provide in both cases electricity, district heating and potable water.

Multi-modular neural networks for the classification of e+e- hadronic events

International Nuclear Information System (INIS)

Proriol, J.

1994-01-01

Some multi-modular neural network methods of classifying e + e - hadronic events are presented. We compare the performances of the following neural networks: MLP (multilayer perceptron), MLP and LVQ (learning vector quantization) trained sequentially, and MLP and RBF (radial basis function) trained sequentially. We introduce a MLP-RBF cooperative neural network. Our last study is a multi-MLP neural network. (orig.)

Advance High Temperature Inspection Capabilities for Small Modular Reactors: Part 1 - Ultrasonics

Energy Technology Data Exchange (ETDEWEB)

Bond, Leonard J. [Iowa State Univ., Ames, IA (United States); Bowler, John R. [Iowa State Univ., Ames, IA (United States)

2017-08-30

The project objective was to investigate the development non-destructive evaluation techniques for advanced small modular reactors (aSMR), where the research sought to provide key enabling inspection technologies needed to support the design and maintenance of reactor component performance. The project tasks for the development of inspection techniques to be applied to small modular reactor are being addressed through two related activities. The first is focused on high temperature ultrasonic transducers development (this report Part 1) and the second is focused on an advanced eddy current inspection capability (Part 2). For both inspection techniques the primary aim is to develop in-service inspection techniques that can be carried out under standby condition in a fast reactor at a temperature of approximately 250°C in the presence of liquid sodium. The piezoelectric material and the bonding between layers have been recognized as key factors fundamental for development of robust ultrasonic transducers. Dielectric constant characterization of bismuth scantanate-lead titanate ((1-x)BiScO₃-xPbTiO₃) (BS-PT) has shown a high Curie temperature in excess of 450°C , suitable for hot stand-by inspection in liquid metal reactors. High temperature pulse-echo contact measurements have been performed with BS-PT bonded to 12.5 mm thick 1018-low carbon steel plate from 20C up to 260 C. High temperature air-backed immersion transducers have been developed with BS-PT, high temperature epoxy and quarter wavlength nickel plate, needed for wetting ability in liquid sodium. Ultrasonic immersion measurements have been performed in water up to 92C and in silicone oil up to 140C. Physics based models have been validated with room temperature experimental data with benchmark artifical defects.

Dynamics and inherent safety features of small modular high temperature gas-cooled reactors

International Nuclear Information System (INIS)

Harrington, R.M.; Ball, S.J.; Cleveland, J.C.

1986-01-01

Investigations were made at Oak Ridge National Laboratory to characterize the dynamics and inherent safety features of various modular high temperature gas-cooled reactor (HTGR) designs. This work was sponsored by the US Nuclear Regulatory Commission's HTGR Safety Research program. The US Department of Energy (DOE) and the Gas Cooled Reactor Associates (GCRA) have sponsored studies of several modular HTGR concepts, each having it own unique advantageous economic and inherent safety features. The DOE design team has recently choses a 350-MW(t) annular core with prismatic, graphite matrix fuel for its reference plant. The various safety features of this plant and of the pebble-bed core designs similar to those currently being developed and operated in the Federal Republic of Germany (FRG) are described. A varity of postulated accident sequences involving combinations of loss of forced circulation of the helium primary coolant, loss of primary coolant pressurization, and loss of normal and backup heat sinks were studied and are discussed. Results demonstrate that each concept can withstand an uncontrolled heatup accident without reaching excessive peak fuel temperatures. Comparisons of calculated and measured response for a loss of forced circulation test on the FRG reactor, AVR, are also presented. 10 refs

Modularization and nuclear power. Report by the Technology Transfer Modularization Task Team

International Nuclear Information System (INIS)

1985-06-01

This report describes the results of the work performed by the Technology Transfer Task Team on Modularization. This work was performed as part of the Technology Transfer work being performed under Department of Energy Contract 54-7WM-335406, between December, 1984 and February, 1985. The purpose of this task team effort was to briefly survey the current use of modularization in the nuclear and non-nuclear industries and to assess and evaluate the techniques available for potential application to nuclear power. A key conclusion of the evaluation was that there was a need for a study to establish guidelines for the future development of Light Water Reactor, High Temperature Gas Reactor and Liquid Metal Reactor plants. The guidelines should identify how modularization can improve construction, maintenance, life extension and decommissioning

Site Suitability and Hazard Assessment Guide for Small Modular Reactors

Energy Technology Data Exchange (ETDEWEB)

Wayne Moe

2013-10-01

Commercial nuclear reactor projects in the U.S. have traditionally employed large light water reactors (LWR) to generate regional supplies of electricity. Although large LWRs have consistently dominated commercial nuclear markets both domestically and abroad, the concept of small modular reactors (SMRs) capable of producing between 30 MW(t) and 900 MW(t) to generating steam for electricity is not new. Nor is the idea of locating small nuclear reactors in close proximity to and in physical connection with industrial processes to provide a long-term source of thermal energy. Growing problems associated continued use of fossil fuels and enhancements in efficiency and safety because of recent advancements in reactor technology suggest that the likelihood of near-term SMR technology(s) deployment at multiple locations within the United States is growing. Many different types of SMR technology are viable for siting in the domestic commercial energy market. However, the potential application of a particular proprietary SMR design will vary according to the target heat end-use application and the site upon which it is proposed to be located. Reactor heat applications most commonly referenced in connection with the SMR market include electric power production, district heating, desalinization, and the supply of thermal energy to various processes that require high temperature over long time periods, or a combination thereof. Indeed, the modular construction, reliability and long operational life purported to be associated with some SMR concepts now being discussed may offer flexibility and benefits no other technology can offer. Effective siting is one of the many early challenges that face a proposed SMR installation project. Site-specific factors dealing with support to facility construction and operation, risks to the plant and the surrounding area, and the consequences subsequent to those risks must be fully identified, analyzed, and possibly mitigated before a license

Design and construction of multi research reactor

International Nuclear Information System (INIS)

1985-05-01

This is the report about design and construction of multi research reactor, which introduces the purpose and necessity of the project, business contents, plan of progress of project and budget for the project. There are three appendixes about status of research reactor in other country, a characteristic of research reactor, three charts about evaluation, process and budget for the multi research reactor and three drawings for the project.

AUS - the Australian modular scheme for reactor neutronics computations

International Nuclear Information System (INIS)

Robinson, G.S.

1975-12-01

A general description is given of the AUS modular scheme for reactor neutronics calculations. The scheme currently includes modules which provide the capacity for lattice calculations, 1D transport calculations, 1 and 2D diffusion calculations (with feedback-free kinetics), and burnup calculations. Details are provided of all system aspects of AUS, but individual modules are only outlined. A complete specification is given of that part of user input which controls the calculation sequence. The report also provides sufficient details of the supervisor program and of the interface data sets to enable additional modules to be incorporated in the scheme. (author)

Development of the DMS (2) (double MS: modular simplified and medium small reactor). Plant layout and construction for the DMS

International Nuclear Information System (INIS)

Shizuka Hirako; Yuusuke Shimizu; Shigeru Yokouchi; Yoshinori Iimura; Kumiaki Moriya; Takahiko Hida

2005-01-01

Nuclear power is expected to become the main source of electric power generation in Japan for reasons of energy security and prevention of CO 2 emissions. In addition, the recent slowdown of electric power demand and the liberalization of the electric power market are accelerating medium and small sized reactor development. Under these circumstances, DMS's (modular simplified and medium small reactors) have been developed as 400MWe class LWR's supported by the Japan Atomic Power Company. In the development of medium and small sized reactors, the most important point is how to overcome the scale demerits. To this end, we have pursued not only the simplification of systems and equipment but also the standardization of layout and construction. The main technical feature of DMS's is the adoption of a natural circulation reactor with short length fuel. Short length fuel enables the reduction of RPV height as well as construction volume of the PCV and building volume. A natural circulation reactor has considerable rationalizing effects such as the elimination of re-circulation pumps and their drive power source. By applying simplified systems and equipment, a rationalized layout and construction method are adopted. To improve the construct ability by means of modular construction methods, steel containment is applied. The PCV size is reduced to 17m in diameter and 24m in height by applying a dish-shaped drywell and eccentric RPV arrangement. By applying a compact PCV and concentrated equipment arrangement in building, it can be confirmed that the ratio of building volume per unit power is equivalent to that of existing large sized ABWRs. Furthermore, a steel plate reinforced concrete structure (SC structure) is applied to the building layout. The application of the compact PCV (steel containment) and the SC structure makes it easier to apply a large-scale module, such as an integrated steel containment and SC structure module, and an integrated multi-layer BM (building

Reference modular High Temperature Gas-Cooled Reactor Plant: Concept description report

International Nuclear Information System (INIS)

1986-10-01

This report provides a summary description of the Modular High Temperature Gas-Cooled Reactor (MHTGR) concept and interim results of assessments of costs, safety, constructibility, operability, maintainability, and availability. Conceptual design of this concept was initiated in October 1985 and is scheduled for completion in 1987. Participating industrial contractors are Bechtel National, Inc. (BNI), Stone and Webster Engineering Corporation (SWEC), GA Technologies, Inc. (GA), General Electric Co. (GE), and Combustion Engineering, Inc

Reference modular High Temperature Gas-Cooled Reactor Plant: Concept description report

Energy Technology Data Exchange (ETDEWEB)

1986-10-01

This report provides a summary description of the Modular High Temperature Gas-Cooled Reactor (MHTGR) concept and interim results of assessments of costs, safety, constructibility, operability, maintainability, and availability. Conceptual design of this concept was initiated in October 1985 and is scheduled for completion in 1987. Participating industrial contractors are Bechtel National, Inc. (BNI), Stone and Webster Engineering Corporation (SWEC), GA Technologies, Inc. (GA), General Electric Co. (GE), and Combustion Engineering, Inc. (C-E).

CFD applications in the Pebble Bed Modular Reactor Project: A decade of progress

International Nuclear Information System (INIS)

Janse van Rensburg, J.J.; Kleingeld, M.

2011-01-01

Highlights: → This paper evaluates the evolution of Gas Cooled Reactor CFD analysis over the last decade. → It discusses the influence of advances in hardware and software on the evolution of capabilities. → The advances in mesh generation and the physics that can be included is also discussed. → The focus was on the capabilities rather than improving the assumptions and correlations. - Abstract: Of all the systems and components that have to be designed for a nuclear plant, the Reactor Unit is the most significant since it is at the very heart of the plant. At Pebble Bed Modular Reactor (Pty) Ltd. (PBMR), the design of the Reactor Unit is conducted with the aid of extensive analysis work. Due to the rapid computational improvements, the analysis capabilities have had to evolve rather significantly over the last decade. This paper evaluates the evolution of RU Computational Fluid Dynamics (CFD) analysis in particular and presents a historical timeline of the analyses conducted at PBMR. The influence of advances in the hardware and software applications on the evolution of the analysis capabilities is also discussed. When evaluating the evolution of analysis, it is important to look not only at the advances in mesh generation and the representation of the geometry, but also at the improvements regarding the physics that were included in the models. The discussion evaluates the improvements from the pre-conceptual analyses, the concept design, the basic design and finally, the detail design. It is however important to note that the focus of this research was on establishing a methodology for the integrated CFD analysis of High Temperature Reactors. It is recognized however that results from this research can currently only be used to investigate and understand trends and behaviors rather than absolute values. It was therefore required to also launch an extensive V and V program of which the focus was to verify the approach and validate the methodology that

Multi purpose research reactor

International Nuclear Information System (INIS)

Raina, V.K.; Sasidharan, K.; Sengupta, Samiran; Singh, Tej

2006-01-01

At present Dhruva and Cirus reactors provide the majority of research reactor based facilities to cater to the various needs of a vast pool of researchers in the field of material sciences, physics, chemistry, bio sciences, research and development work for nuclear power plants and production of radio isotopes. With a view to further consolidate and expand the scope of research and development in nuclear and allied sciences, a new 20 MWt multi purpose research reactor is being designed. This paper describes some of the design features and safety aspects of this reactor

A Small Modular Reactor Core Design using FCM Fuel and BISO BP particles

Energy Technology Data Exchange (ETDEWEB)

Choi, Jae Yeon; Hwang, Dae Hee; Yoo, Ho Seong; Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of)

2016-10-15

The objective of this work is to design a PWR small modular reactor which employs the advanced fuel technology of FCM particle fuels including BISO burnable poisons and advanced cladding of SiC in order to improve the fuel economy and safety by increasing fuel burnup and temperature, and by reducing hydrogen generation under accidents. Recently, many countries including USA have launched projects to develop the accident tolerant fuels (ATF) which can cope with the accidents such as LOCA (Loss of Coolant Accident). In general, the ATF fuels are required to meet the PWR operational, safety, and fuel cycle constraints which include enhanced burnup, lower or no generation of hydrogen, lower operating temperatures, and enhanced retention of fission products. Another stream of research and development in nuclear society is to develop advanced small modular reactors in order to improve inherent passive safety and to reduce the risk of large capital investment. In this work, a small PWR modular reactor core was neutronically designed and analyzed. The SMR core employs new 13x13 fuel assemblies which are loaded with thick FCM fuel rods in which TRISO fuel particles AO and also the first cycle has the AOs which are within the typical design limit. Also, this figure shows that the evolutions of AO for the cycles 6 and 7 are nearly the same. we considered the SiC cladding for reduction of hydrogen generation under accidents. From the results of core design and analysis, it is shown that the core has long cycle length of 732 -1191 EFPDs, high discharge burnup of 101-105 MWD/kg, low power peaking factors, low axial offsets, negative MTCs, and large shutdown margins except for BOC of the first cycle. So, it can be concluded that the new SMR core is neutronically feasible.

A Small Modular Reactor Core Design using FCM Fuel and BISO BP particles

International Nuclear Information System (INIS)

Choi, Jae Yeon; Hwang, Dae Hee; Yoo, Ho Seong; Hong, Ser Gi

2016-01-01

The objective of this work is to design a PWR small modular reactor which employs the advanced fuel technology of FCM particle fuels including BISO burnable poisons and advanced cladding of SiC in order to improve the fuel economy and safety by increasing fuel burnup and temperature, and by reducing hydrogen generation under accidents. Recently, many countries including USA have launched projects to develop the accident tolerant fuels (ATF) which can cope with the accidents such as LOCA (Loss of Coolant Accident). In general, the ATF fuels are required to meet the PWR operational, safety, and fuel cycle constraints which include enhanced burnup, lower or no generation of hydrogen, lower operating temperatures, and enhanced retention of fission products. Another stream of research and development in nuclear society is to develop advanced small modular reactors in order to improve inherent passive safety and to reduce the risk of large capital investment. In this work, a small PWR modular reactor core was neutronically designed and analyzed. The SMR core employs new 13x13 fuel assemblies which are loaded with thick FCM fuel rods in which TRISO fuel particles AO and also the first cycle has the AOs which are within the typical design limit. Also, this figure shows that the evolutions of AO for the cycles 6 and 7 are nearly the same. we considered the SiC cladding for reduction of hydrogen generation under accidents. From the results of core design and analysis, it is shown that the core has long cycle length of 732 -1191 EFPDs, high discharge burnup of 101-105 MWD/kg, low power peaking factors, low axial offsets, negative MTCs, and large shutdown margins except for BOC of the first cycle. So, it can be concluded that the new SMR core is neutronically feasible

NEPTUNE: a modular scheme for the calculation of light water reactors

International Nuclear Information System (INIS)

Kavenoky, A.

1975-01-01

The NEPTUNE modular scheme has been developed to provide the physicist and the design engineer with a single system of codes for the calculation of light water reactors. The APOLLO code is included in NEPTUNE for the multigroup transport treatment of cells, groups of cells and complete fuel assemblies; few groups cross section libraries are automatically transmitted to the reactor multidimensional diffusion modules. In the reactor phase, 1D and 2D diffusion calculations can be performed by use of the finite difference method; 2D and 3D calculations are done respectively by the BILAN and TRIDENT modules using the finite element method. For the depletion calculation coarse and refined computations are offered. NEPTUNE is characterized by two special features for the data processing: the OTOMAT system which provides a virtual memory simulation and the intervention Monitor which allow to disconnect the computation modules and the control modules [fr

Nuclear security regulatory framework analysis for small modular reactors in Canada and abroad

Energy Technology Data Exchange (ETDEWEB)

Farah, A., E-mail: amjad.farah@uoit.ca [University of Ontario Institute of Technology, Oshawa, ON (Canada)

2015-07-01

Small Modular Reactors (SMRs) are gaining global attention as a potential solution for future power plants due to claims of flexibility and cost effectiveness, while maintaining or increasing safety and security. With the change of design and the potential deployment in remote areas, however, challenges arise from a regulatory standpoint, to meet the safety and security regulations while maintaining economic feasibility. This work comprises of a review of the nuclear security regulatory frameworks in place for SMRs in Canada, USA and the IAEA; how they compare to each other, and to those of large reactors. The goal is to gauge what needs to be adjusted in order to address the changes in design between the two reactor sizes. Some key challenges concern the type of reactor, transportation of reactor components and fuel to remote areas, reduced security staff, and increased complexity of emergency planning and evacuation procedures. (author)

Nuclear security regulatory framework analysis for small modular reactors in Canada and abroad

International Nuclear Information System (INIS)

Farah, A.

2015-01-01

Small Modular Reactors (SMRs) are gaining global attention as a potential solution for future power plants due to claims of flexibility and cost effectiveness, while maintaining or increasing safety and security. With the change of design and the potential deployment in remote areas, however, challenges arise from a regulatory standpoint, to meet the safety and security regulations while maintaining economic feasibility. This work comprises of a review of the nuclear security regulatory frameworks in place for SMRs in Canada, USA and the IAEA; how they compare to each other, and to those of large reactors. The goal is to gauge what needs to be adjusted in order to address the changes in design between the two reactor sizes. Some key challenges concern the type of reactor, transportation of reactor components and fuel to remote areas, reduced security staff, and increased complexity of emergency planning and evacuation procedures. (author)

Multi-purpose nuclear heat source for advanced gas-cooled reactor plants

International Nuclear Information System (INIS)

McDonald, C.F.

1993-01-01

Nuclear power has the potential to be the ultimate green technology in that it could eliminate the need for burning fossil fuels with their polluting combustion products and greenhouse gases. This view is shared by many technologists, but it may be a generation before the public becomes convinced, and that will involve overcoming many safety, institutional, financial, and technical impediments. This paper addresses only the latter topic; a major theme being that for nuclear power to truly be a green technology and significantly benefit society, it must meet the needs of the full energy spectrum. Specifically, it must satisfy energy needs beyond just the electricity generating sector by today's nuclear plants. By virtue of its high temperature capability, the Modular High Temperature Gas-Cooled Reactor (MHTGR) is the only type of reactor that has the potential to meet the wide range of energy needs that will emerge in the future. This paper discusses the nuclear heat source that gives the MHTGR multi-purpose capability, which is recognized today, but will not be implemented until early in the next century

Passive Safety Features for Small Modular Reactors

International Nuclear Information System (INIS)

Ingersoll, Daniel T.

2010-01-01

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

A review of reactor physics uncertainties and validation requirements for the modular high-temperature gas-cooled reactor

International Nuclear Information System (INIS)

Baxter, A.M.; Lane, R.K.; Hettergott, E.; Lefler, W.

1991-01-01

The important, safety-related, physics parameters for the low-enriched Modular High-Temperature gas-Cooled Reactor (MHTGR) such as control rod worth, shutdown margins, temperature coefficients, and reactivity worths, are considered, and estimates are presented of the uncertainties in the calculated values of these parameters. The basis for the uncertainty estimate in several of the important calculated parameters is reviewed, including the available experimental data used in obtaining these estimates. Based on this review, the additional experimental data needed to complete the validation of the methods used to calculate these parameters is presented. The role of benchmark calculations in validating MHTGR reactor physics data is also considered. (author). 10 refs, 5 figs, 3 tabs

Considerations in the development of safety requirements for innovative reactors: Application to modular high temperature gas cooled reactors

International Nuclear Information System (INIS)

2003-08-01

Member States of the IAEA have frequently requested this organization to assess, at the conceptual stage, the safety of the design of nuclear reactors that rely on a variety of technologies and are of a high degree of innovation. However, to date, for advanced and innovative reactors and for reactors with characteristics that are different from those of existing light water reactors, widely accepted design standards and rules do not exist. This TECDOC is an outcome of the efforts deployed by the IAEA to develop a general approach for assessing the safety of the design of advanced and innovative reactors, and of all reactors in general including research reactors, with characteristics that differ from those of light water reactors. This publication puts forward a method for safety assessment that is based on the well established and accepted principle of defence in depth. The need to develop a general approach for assessing the safety of the design of reactors that applies to all kinds of advanced reactors was emphasized by the request to the IAEA by South Africa to review the safety of the South African pebble bed modular reactor. This reactor, as other modular high temperature gas cooled reactors (MHTGRs), adopts very specific design features such as the use of coated particle fuel. The characteristics of the fuel deeply affect the design and the safety of the plant, thereby posing several challenges to traditional safety assessment methods and to the application of existing safety requirements that have been developed primarily for water reactors. In this TECDOC, the MHTGR has been selected as a case study to demonstrate the viability of the method proposed. The approach presented is based on an extended interpretation of the concept of defence in depth and its link with the general safety objectives and fundamental safety functions as set out in 'Safety of Nuclear Power Plants: Design', IAEA Safety Standards No. NS-R.1, issued by the IAEA in 2000. The objective

Economic evaluation of small modular nuclear reactors and the complications of regulatory fee structures

International Nuclear Information System (INIS)

Vegel, Benjamin; Quinn, Jason C.

2017-01-01

Carbon emission concerns and volatility in fossil fuel resources have renewed world-wide interest in nuclear energy as a solution to growing energy demands. Several large nuclear reactors are currently under construction in the United States, representing the first new construction in over 30 years. Small Modular Reactors (SMRs) have been in design for many years and offer potential technical and economic advantages compared with traditionally larger reactors. Current SMR capital and operational expenses have a wide range of uncertainty. This work evaluates the potential for SMRs in the US, develops a robust techno-economic assessment of SMRs, and leverages the model to evaluate US regulatory fees structures. Modeling includes capital expenses of a factory facility and capital and operational expenses with multiple scenarios explored through a component-level capital cost model. Policy regarding the licensing and regulation of SMRs is under development with proposed annual US regulatory fees evaluated through the developed techno-economic model. Results show regulatory fees are a potential barrier to the economic viability of SMRs with an alternate fee structure proposed and evaluated. The proposed fee structure is based on the re-distribution of fees for all nuclear reactors under a single structure based on reactor thermal power rating. - Highlights: • Potential demand for new small modular nuclear power in the US is established. • Capital costs are broken down on component level and include factory production. • US regulatory fees structures are evaluated, results show potential barrier. • An additional fee structure is proposed and compared with current US fee structures.

CFD Analysis of the Primary Cooling System for the Small Modular Natural Circulation Lead Cooled Fast Reactor SNRLFR-100

Directory of Open Access Journals (Sweden)

Pengcheng Zhao

2016-01-01

Full Text Available Small modular reactor (SMR has drawn wide attention in the past decades, and Lead cooled fast reactor (LFR is one of the most promising advanced reactors which are able to meet the safety economic goals of Gen-IV nuclear energy systems. A small modular natural circulation lead cooled fast reactor-100 MWth (SNRLFR-100 is being developed by University of Science and Technology of China (USTC. In the present work, a 3D CFD model, primary heat exchanger model, fuel pin model, and point kinetic model were established based on some reasonable simplifications and assumptions, the steady-state natural circulation characteristics of SNCLFR-100 primary cooling system were discussed and illustrated, and some reasonable suggestions were proposed for the reactor’s thermal-hydraulic and structural design. Moreover, in order to have a first evaluation of the system behavior in accident conditions, an unprotected loss of heat sink (ULOHS transient simulation at beginning of the reactor cycle (BOC has been analyzed and discussed based on the steady-state simulation results. The key temperatures of the reactor core are all under the safety limits at transient state; the reactor has excellent thermal-hydraulic performance.

Concepts for space nuclear multi-mode reactors

International Nuclear Information System (INIS)

Myrabo, L.; Botts, T.E.; Powell, J.R.

1983-01-01

A number of nuclear multi-mode reactor power plants are conceptualized for use with solid core, fixed particle bed and rotating particle bed reactors. Multi-mode systems generate high peak electrical power in the open cycle mode, with MHD generator or turbogenerator converters and cryogenically stored coolants. Low level stationkeeping power and auxiliary reactor cooling (i.e., for the removal of reactor afterheat) are provided in a closed cycle mode. Depending on reactor design, heat transfer to the low power converters can be accomplished by heat pipes, liquid metal coolants or high pressure gas coolants. Candidate low power conversion cycles include Brayton turbogenerator, Rankine turbogenerator, thermoelectric and thermionic approaches. A methodology is suggested for estimating the system mass of multi-mode nuclear power plants as a function of peak electric power level and required mission run time. The masses of closed cycle nuclear and open cycle chemical power systems are briefly examined to identify the regime of superiority for nuclear multi-mode systems. Key research and technology issues for such power plants are also identified

Design characteristics for pressurized water small modular nuclear power reactors with focus on safety

Energy Technology Data Exchange (ETDEWEB)

Kani, Iraj Mahmoudzadeh [Tehran Univ. (Iran, Islamic Republic of). Civil Faculty; Zandieh, Mehdi [Tehran Univ. (Iran, Islamic Republic of). Civil Faculty; International Univ. of Imam Khomeini (Iran, Islamic Republic of). Architecture Faculty; Abadi, Saeed Kheirollahi Hossein [International Univ. of Imam Khomeini (Iran, Islamic Republic of). Architecture Faculty

2016-05-15

Small Modular Reactors (SMRs) are a technology, attracting attention. Light water SMR possess an upgraded design case and emphasize the significance of integral models. Beside of these advantages, SMRs has faced numerous challenges, e.g. licensing, cost/investment, safety and security observation, social and environmental issues in building new plants.

Application of S-CO{sub 2} Cycle for Small Modular Reactor coupled with Desalination System

Energy Technology Data Exchange (ETDEWEB)

Lee, Won Woong; Bae, Seong Jun; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of)

2016-10-15

The Korean small modular reactor, SMART (System-integrated Modular Advanced ReacTor, 100MWe), is designed to achieve enhanced safety and improved economics through reliable passive safety systems, a system simplification and component modularization. SMART can generate electricity and provide water by seawater desalination. However, due to the desalination aspect of SMART, the total amount of net electricity generation is decreased from 100MWe to 90MWe. The authors suggest in this presentation that the reduction of electricity generation can be replenished by applying S-CO{sub 2} power cycle technology. The S-CO{sub 2} Brayton cycle, which is recently receiving significant attention as the next generation power conversion system, has some benefits such as high cycle efficiency, simple configuration, compactness and so on. In this study, the cycle performance analysis of the S-CO{sub 2} cycles for SMART with desalination system is conducted. The simple recuperated S-CO{sub 2} cycle is revised for coupling with desalination system. The three revised layout are proposed for the cycle performance comparison. In this results of the 3rd revised layout, the cycle efficiency reached 37.8%, which is higher than the efficiency of current SMART with the conventional power conversion system 30%.

Economic competitiveness of small modular reactors versus coal and combined cycle plants

International Nuclear Information System (INIS)

Alonso, Gustavo; Bilbao, Sama; Valle, Edmundo del

2016-01-01

Small modular reactors (SMRs) may be an option to cover the electricity needs of isolated regions, distributed generation grids and countries with small electrical grids. Previous analyses show that the overnight capital cost for SMRs is between 4500 US$/kW and 5350 US$/kW, which is between a 6% and a 26% higher than the average cost of a current large nuclear reactor. This study analyzes the economic competitiveness of small modular reactors against thermal plants using coal and natural gas combined cycle plants. To assess the economic competitiveness of SMRs, three overnight capital costs are considered 4500 US$/kW, 5000 US$/kW and 5350 US$/kW along with three discount rates for each overnight cost considered, these are 3, 7, and 10%. To compare with natural gas combined cycle (CC) units, four different gas prices are considered, these are 4.74 US$/GJ (5 US$/mmBTU), 9.48 US$/GJ (10 US$/mmBTU), 14.22 US$/GJ (15 US$/mmBTU), and 18.96 US$/GJ (20 US$/mmBTU). To compare against coal, two different coal prices are considered 80 and 120 US$/ton of coal. The carbon tax considered, for both CC and coal, is 30 US$/ton CO_2. The results show what scenarios make SMRs competitive against coal and/or combined cycle plants. In addition, because the price of electricity is a key component to guarantee the feasibility of a new project, this analysis calculates the price of electricity for the economically viable deployment of SMRs in all the above scenarios. In particular, this study shows that a minimum price of electricity of 175 US$/MWh is needed to guarantee the feasibility of a new SMR, if its overnight capital cost is 5350 US$/kWe and the discount rate is 10%. Another result is that when the price of electricity is around 100 US$/MWh then the discount rate must be around 7% or less to provide appropriate financial conditions to make SMRs economically feasible. - Highlights: • Small modular reactor (SMR) are economically assessed. • SMR are compared against gas and coal

Numerical benchmark for the deep-burn modular helium-cooled reactor (DB-MHR)

International Nuclear Information System (INIS)

Taiwo, T. A.; Kim, T. K.; Buiron, L.; Varaine, F.

2006-01-01

Numerical benchmark problems for the deep-burn concept based on the prismatic modular helium-cooled reactor design (a Very High Temperature Reactor (VHTR)) are specified for joint analysis by U.S. national laboratories and industry and the French CEA. The results obtained with deterministic and Monte Carlo codes have been inter-compared and used to confirm the underlying feature of the DB-MHR concept (high transuranics consumption). The results are also used to evaluate the impact of differences in code methodologies and nuclear data files on the code predictions for DB-MHR core physics parameters. The code packages of the participating organizations (ANL and CEA) are found to give very similar results. (authors)

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

International Nuclear Information System (INIS)

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

2007-01-01

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

An exergoeconomic assessment of waste heat recovery from a Gas Turbine-Modular Helium Reactor using two transcritical CO_2 cycles

International Nuclear Information System (INIS)

Wang, Xurong; Dai, Yiping

2016-01-01

Highlights: • A Gas Turbine-Modular Helium Reactor is coupled with two transcritical CO_2 cycles. • Exergoeconomic analysis and optimization of the combined cycle was performed. • The energy efficiency of the cogeneration system was enhanced by 7.92% at 850 °C. • The overall exergoeconomic factor of the system could up to 55.2%. - Abstract: A comprehensive study is performed on an energy conversion system which combines a gas turbine-modular helium reactor (GT-MHR) and two transcritical CO_2 cycles (tCO_2). The aim of this study is to assess the energy, exergy and economic behavior of the proposed system, considering five indicators: the energy efficiency, the exergy efficiency, the total exergy destruction cost rate, the overall exergoeconomic factor and the total cost rate. A parametric study is also conducted to evaluate the influence of key decision variables on the GT-MHR/tCO_2 performance. Finally, the combined cycle is optimized to minimize the total cost rate. The results show that the energy efficiency of GT-MHR/tCO_2 cycle is 7.92% higher than that of the simple GT-MHR cycle at 850 °C. The largest exergy destruction rate takes place in the reactor, and after that in the helium turbine and the recuperator. The components in tCO_2 cycles have less exergy destruction. When the optimization is conducted based on the exergoeconomics, the overall exergoeconomic factor, the total cost rate and the total exergy destruction cost rate are 55.2%, 20,752 $/h and 9292 $/h, respectively.

Modeling and Application of Pneumatic Conveying for Spherical Fuel Element in Pebble-Bed Modular High-Temperature Gas-Cooled Reactor

International Nuclear Information System (INIS)

Zhou Shuyong; Wang Junsan; Wang Yuding; Cai Ruizhong; Zhang Xuan; Cao Jianting

2014-01-01

The fuel handling system is an important system for on-load refueling in pebble-bed modular high-temperature gas-cooled reactor. A dynamic model of pneumatic conveying for spherical fuel element in fuel handling system was established to describe the pneumatically conveying process. The motion characteristics of fuel elements in pipeline and the effect of fuel elements on gas velocity were studied using the model. The results show that the theoretical analyses are consistent with the experimental. The research has been used in developing full scope simulator for pebble-bed modular high-temperature gas-cooled reactor, also provides references for the design and optimization of the fuel handling system. (author)

From field to factory-Taking advantage of shop manufacturing for the pebble bed modular reactor

International Nuclear Information System (INIS)

Wallace, Edward; Matzie, Regis; Heiderd, Roger; Maddalena, John

2006-01-01

The move of nuclear plant construction from the field to the factory for small, advanced pebble bed modular reactor (PBMR) designs has significant benefits compared to traditional light water reactor (LWR) field oriented designs. The use of modular factory construction techniques has a growing economic benefit over time through well-established process learning applications. This paper addresses the basic PBMR design objectives and commercialization model that drive this approach; provides a brief technical description of the PBMR design and layout with representative CAD views and discusses derived figures of merit highlighting the relative simplicity of PBMR compared to a modern LWR. The discussion emphasizes that more of PBMR can be built in the factory due to the simple design of a direct helium Brayton cycle compared to an indirect LWR steam cycle with its associated equipment. For the PBMR design there are fewer and less cumbersome auxiliary and safety systems with their attendant support requirements. Additionally, the labor force economic efficiency for nuclear projects is better in the factory than in the field, including consideration of labor costs and nuclear quality programs. Industrial learning is better in the factory because of the more controlled environment, mechanization optimization opportunities and because of the more stable labor force compared to the field. Supply chain benefits are more readily achievable with strategic contracts for module suppliers. Although building a nuclear power plant is not a typical high volume manufacturing process, for the PBMR-type of plant, with its high degree of standardization and relatively small, simplified design, the shift to factory work has a significant impact on overall project cost due to earlier identification and better coordination of parallel construction paths. This is in stark contrast to the construction of a large LWR in the past. Finally, the PBMR modular plant concept continues at the

Development history of the gas turbine modular high temperature reactor

International Nuclear Information System (INIS)

Brey, H.L.

2001-01-01

The development of the high temperature gas cooled reactor (HTGR) as an environmentally agreeable and efficient power source to support the generation of electricity and achieve a broad range of high temperature industrial applications has been an evolutionary process spanning over four decades. This process has included ongoing major development in both the HTGR as a nuclear energy source and associated power conversion systems from the steam cycle to the gas turbine. This paper follows the development process progressively through individual plant designs from early research of the 1950s to the present focus on the gas turbine modular HTGR. (author)

Dynamic and Control Analysis of Modular Multi-Parallel Rectifiers (MMR)

DEFF Research Database (Denmark)

Zare, Firuz; Ghosh, Arindam; Davari, Pooya

2017-01-01

This paper presents dynamic analysis of a Modular Multi-Parallel Rectifier (MMR) based on state-space modelling and analysis. The proposed topology is suitable for high power application which can reduce line current harmonics emissions significantly. However, a proper controller is required...... to share and control current through each rectifier. Mathematical analysis and preliminary simulations have been carried out to verify the proposed controller under different operating conditions....

Studies on a Stellarator reactor of the Helias type: The modular coil system

International Nuclear Information System (INIS)

Harmeyer, E.; Kisslinger, J.; Rau, F.; Wobig, H.

1993-02-01

Helias Stellarator Reactors (HSR) are considered, focussing on the superconducting modular coil system which generates the magnetic field, aiming to clarify critical issues of such systems. The development of the coil system is presented and the properties of the vacuum magnetic field are discussed. Electromagnetic forces and the resulting mechanical stresses and strains inside the coils and the surrounding structure are calculated. Parameter studies are made varying the major radius R 0 between 18 m and 24 m in order to investigate the engineering parameters for the superconducting coil system. The total mass and the fusion power output of HSR are compared with values evaluated for tokamak reactors. (orig.). 36 figs

Activity-based costing as a method for assessing the economics of modularization

DEFF Research Database (Denmark)

Thyssen, Jesper; Israelsen, Poul; Jørgensen, Brian

2006-01-01

The paper accounts for an Activity-Based Costing (ABC) analysis supporting decision-making concerning product modularity. The ABC analysis carried out is communicated to decision-makers by telling how much higher the variable cost of the multi-purpose module can be compared to the average variable...... cost for the product-unique modules that it substitutes to break even in total cost. The analysis provides the platform for stating three general rules of cost efficiency of modularization, which in combination identify the highest profit potential of product modularization. Finally the analysis points...... to problems of using ABC in costing modularity, i.e. handling of R&D costs and identification of product profitability upon an enhanced modularization....

Fusion reactor remote maintenance study. Final report

International Nuclear Information System (INIS)

Sniderman, M.

1979-04-01

An analysis of a major maintenance operation, the remote replacement of a modular sector of a tokamak reactor, was performed in substantial detail. Specific assumptions were developed which included concepts from various existing designs so that the operation which was studied includes some design features generic to any fusion reactor design. Based on the work performed in this study, the principal conclusions are: (1) It appears feasible to design a tokamak fusion reactor plant with availability comparable to existing fossil and fission plants, but this will require diligence and comprehensive planning during the complete design phase. (2) Since the total fusion program is paced by the success of each device, maintenance considerations must be incorporated into each device during design, even if the device is an experimental unit. (3) Innovative approaches, such as automatic computer controlled operations, should be developed so that large step reductions in planned maintenance times can be achieved

Preliminary analysis in support to the experimental activities on the mixing process in the pressurizer of a small modular reactor integrated primary system

Energy Technology Data Exchange (ETDEWEB)

Nascimento, Samira R.V.; Lira, Carlos A.B.O.; Bezerra, Jair L.; Silva, Mario A.B.; Silva, Willdauany C.F., E-mail: samiraruana@gmail.com [Universidade Federal de Pernambuco (DEN/UFPE), Recife, PE (Brazil). Departamento de Energia Nuclear; Lapa, Celso M.F., E-mail: lapa@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil); Lima, Fernando R.A., E-mail: falima@crcn.gov.br [Centro Regional de Ciencias Nucleares (CRCN/CNEN-NE), Recife, PE (Brazil); Otero, Maria E.M.; Hernandez, Carlos R.G., E-mail: mmontesi@instec.cu [Department of Nuclear Engineering, InSTEC/CUBA, Higher Institute of Technology and Applied Science, La Habana (Cuba)

2015-07-01

Nowadays, there is a renewed interest in the development of advanced/innovative small and medium sized modular reactors (SMRs). The SMRs are variants of the Generation IV systems and usually have attractive characteristics of simplicity, enhanced safety and require limited financial resources. The concept of the integrated primary system reactor (IPSR) is characterized by the inclusion of the entire primary system within a single pressure vessel, including the steam generator and pressurizer. The pressurizer is located within the reactor vessel top, this configuration involves changes on the techniques and is necessary investigate the boron mixing. The present work represents a contribution to the design of an experimental facility planned to provide data relevant for the mixing phenomena in the pressurizer of a compact modular reactor. In particular, in order to evaluate the boron concentration in the surge orifices to simulate the in-surge and out-surge in a facility, scaled 1:200, respect to the ¼ of the pressurizer. The facility behavior studied from one inlet and one outlet of the test section with represent one in-surge e one out-surge the pressurizer of a small modular reactor integrated primary system. (author)

Nuclear Safeguards Considerations For The Pebble Bed Modular Reactor (PBMR)

Energy Technology Data Exchange (ETDEWEB)

Phillip Casey Durst; David Beddingfield; Brian Boyer; Robert Bean; Michael Collins; Michael Ehinger; David Hanks; David L. Moses; Lee Refalo

2009-10-01

High temperature reactors (HTRs) have been considered since the 1940s, and have been constructed and demonstrated in the United Kingdom (Dragon), United States (Peach Bottom and Fort Saint Vrain), Japan (HTTR), Germany (AVR and THTR-300), and have been the subject of conceptual studies in Russia (VGM). The attraction to these reactors is that they can use a variety of reactor fuels, including abundant thorium, which upon reprocessing of the spent fuel can produce fissile U-233. Hence, they could extend the stocks of available uranium, provided the fuel is reprocessed. Another attractive attribute is that HTRs typically operate at a much higher temperature than conventional light water reactors (LWRs), because of the use of pyrolytic carbon and silicon carbide coated (TRISO) fuel particles embedded in ceramic graphite. Rather than simply discharge most of the unused heat from the working fluid in the power plant to the environment, engineers have been designing reactors for 40 years to recover this heat and make it available for district heating or chemical conversion plants. Demonstrating high-temperature nuclear energy conversion was the purpose behind Fort Saint Vrain in the United States, THTR-300 in Germany, HTTR in Japan, and HTR-10 and HTR-PM, being built in China. This resulted in nuclear reactors at least 30% or more thermodynamically efficient than conventional LWRs, especially if the waste heat can be effectively utilized in chemical processing plants. A modern variant of high temperature reactors is the Pebble Bed Modular Reactor (PBMR). Originally developed in the United States and Germany, it is now being redesigned and marketed by the Republic of South Africa and China. The team examined historical high temperature and high temperature gas reactors (HTR and HTGR) and reviewed safeguards considerations for this reactor. The following is a preliminary report on this topic prepared under the ASA-100 Advanced Safeguards Project in support of the NNSA Next

Capital cost evaluation of liquid metal reactor by plant type - comparison of modular type with monolithic type -

International Nuclear Information System (INIS)

Mun, K. H.; Seok, S. D.; Song, K. D.; Kim, I. C.

1999-01-01

A preliminary economic comparison study was performed for KALIMER(Korea Advanced LIquid MEtal Reactor)between a modular plant type with 8 150MWe modules and a 1200MWe monolithic plant type. In both cases of FOAK (First-Of-A-Kind) Plant and NOAK (Nth-Of-A-Kind) Plant, the result says that the economics of monolithic plant is superior to its modular plant. In case of NOAK plant comparison, however, the cost difference is not significant. It means that modular plant can compete with monolithic plant in capital cost if it makes efforts of cost reduction and technical progress on the assumption that the same type of NOAK plant will be constructed continuously

Saturated Adaptive Output-Feedback Power-Level Control for Modular High Temperature Gas-Cooled Reactors

Directory of Open Access Journals (Sweden)

Zhe Dong

2014-11-01

Full Text Available Small modular reactors (SMRs are those nuclear fission reactors with electrical output powers of less than 300 MWe. Due to its inherent safety features, the modular high temperature gas-cooled reactor (MHTGR has been seen as one of the best candidates for building SMR-based nuclear plants with high safety-level and economical competitive power. Power-level control is crucial in providing grid-appropriation for all types of SMRs. Usually, there exists nonlinearity, parameter uncertainty and control input saturation in the SMR-based plant dynamics. Motivated by this, a novel saturated adaptive output-feedback power-level control of the MHTGR is proposed in this paper. This newly-built control law has the virtues of having relatively neat form, of being strong adaptive to parameter uncertainty and of being able to compensate control input saturation, which are given by constructing Lyapunov functions based upon the shifted-ectropies of neutron kinetics and reactor thermal-hydraulics, giving an online tuning algorithm for the controller parameters and proposing a control input saturation compensator respectively. It is proved theoretically that input-to-state stability (ISS can be guaranteed for the corresponding closed-loop system. In order to verify the theoretical results, this new control strategy is then applied to the large-range power maneuvering control for the MHTGR of the HTR-PM plant. Numerical simulation results show not only the relationship between regulating performance and control input saturation bound but also the feasibility of applying this saturated adaptive control law practically.

Energy Multiplier Module (EM{sup 2}) - advanced small modular reactor for electricity generation

Energy Technology Data Exchange (ETDEWEB)

Bertch, T.; Schleicher, R.; Choi, H.; Rawls, J., E-mail: timothy.bertch@ga.com [General Atomics, San Diego, California (United States)

2013-07-01

In order to provide cost effective nuclear energy in other than large reactor, large grid applications, fission technology needs to make further advances. 'Convert and burn' fast reactors offer long life cores, improved fuel utilization, reduced waste and other benefits while achieving cost effective energy production in a smaller reactor. General Atomics' Energy Multiplier Module (EM{sup 2}), a helium-cooled compact fast reactor that augments its fissile fuel load with either depleted uranium (DU) or used nuclear fuel (UNF). The convert and burn in-situ provides 250 MWe with a 30 year core life. High temperature provides a simple, high efficiency direct cycle gas turbine which along with modular construction, fewer systems, road shipment and minimum on site construction support cost effectiveness. Additional advantages in fuel cycle, non-proliferation and siting flexibility and its ability to meet all safety requirements make for an attractive power source, especially in remote and small grid regions. (author)

Implications of small modular reactors for climate change mitigation

International Nuclear Information System (INIS)

Iyer, Gokul; Hultman, Nathan; Fetter, Steve; Kim, Son H.

2014-01-01

Achieving climate policy targets will require large-scale deployment of low-carbon energy technologies, including nuclear power. The small modular reactor (SMR) is viewed as a possible solution to the problems of energy security as well as climate change. In this paper, we use an integrated assessment model (IAM) to investigate the evolution of a global energy portfolio with SMRs under a stringent climate policy. Technology selection in the model is based on costs; we use results from previous expert elicitation studies of SMR costs. We find that the costs of achieving a 2 °C target are lower with SMRs than without. The costs are higher when large reactors do not compete for market share compared to a world in which they can compete freely. When both SMRs and large reactors compete for market share, reduction in mitigation cost is achieved only under advanced assumptions about SMR technology costs and future cost improvements. While the availability of SMRs could lower mitigation costs by a moderate amount, actual realization of these benefits would depend on the rapid up-scaling of SMRs in the near term. Such rapid deployment could be limited by several social, institutional and behavioral obstacles. - Highlights: • Costs of achieving a 2 °C target are lower with SMRs than without. • Costs are higher when large reactors do not compete for market share. • Under competition, cost is reduced only with advanced SMR technology. • Realization of benefits will depend on rapid near term up-scaling of SMRs

Selection of design basis event for modular high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Sato, Hiroyuki; Nakagawa, Shigeaki; Ohashi, Hirofumi

2016-06-01

Japan Atomic Energy Agency (JAEA) has been investigating safety requirements and basic approach of safety guidelines for modular High Temperature Gas-cooled Reactor (HTGR) aiming to increase internarial contribution for nuclear safety by developing an international HTGR safety standard under International Atomic Energy Agency. In this study, we investigate a deterministic approach to select design basis events utilizing information obtained from probabilistic approach. In addition, selections of design basis events are conducted for commercial HTGR designed by JAEA. As a result, an approach for selecting design basis event considering multiple failures of safety systems is established which has not been considered as design basis in the safety guideline for existing nuclear facility. Furthermore, selection of design basis events for commercial HTGR has completed. This report provides an approach and procedure for selecting design basis events of modular HTGR as well as selected events for the commercial HTGR, GTHTR300. (author)

Numerical Simulation of Particle Flow Motion in a Two-Dimensional Modular Pebble-Bed Reactor with Discrete Element Method

Directory of Open Access Journals (Sweden)

Guodong Liu

2013-01-01

Full Text Available Modular pebble-bed nuclear reactor (MPBNR technology is promising due to its attractive features such as high fuel performance and inherent safety. Particle motion of fuel and graphite pebbles is highly associated with the performance of pebbled-bed modular nuclear reactor. To understand the mechanism of pebble’s motion in the reactor, we numerically studied the influence of number ratio of fuel and graphite pebbles, funnel angle of the reactor, height of guide ring on the distribution of pebble position, and velocity by means of discrete element method (DEM in a two-dimensional MPBNR. Velocity distributions at different areas of the reactor as well as mixing characteristics of fuel and graphite pebbles were investigated. Both fuel and graphite pebbles moved downward, and a uniform motion was formed in the column zone, while pebbles motion in the cone zone was accelerated due to the decrease of the cross sectional flow area. The number ratio of fuel and graphite pebbles and the height of guide ring had a minor influence on the velocity distribution of pebbles, while the variation of funnel angle had an obvious impact on the velocity distribution. Simulated results agreed well with the work in the literature.

The multi region molten-salt reactor concept

International Nuclear Information System (INIS)

Gyula, Csom; Sandor, Feher; Szieberth, M.; Szabolcs, Czifrus

2003-01-01

The molten-salt reactor (MSR) concept is one of the most promising systems for the realisation of transmutation. The objective is the development of a transmutation technique along with a device implementing it, which yield higher transmutation efficiencies than that of the known procedures. The procedure is the multi-step transmutation, in which the transformation is carried out in several consecutive steps of different neutron flux and spectrum. In order to implement this, a multi-region transmutation device, i.e. nuclear reactor or sub-critical system is proposed, in which several separate flow-through irradiation rooms are formed with various neutron spectra and fluxes. The paper presents calculations that were performed for a special 5-region version of the multi-region molten-salt reactor. (author)

Numerical simulation of severe water ingress accidents in a modular high temperature gas cooled reactor

International Nuclear Information System (INIS)

Zhang Zuoyi; Scherer, W.

1996-01-01

This report analyzes reverse water ingress accidents in the SIEMENS 200 MW Modular Pebble-Bed High Temperature Gas Cooled Reactor (HTR-MODULE) under the assumption of no active safety protection systems in order to find the safety margins of the current HTR-MODULE design and to realize a catastrophe-free nuclear technology. A water, steam and helium multi-phase cavity model is developed and implemented in the DSNP simulation system. The DSNP system is then used to simulate the primary and secondary circuit of a HTR-MODULE power plant. Comparisons of the model with experiments and with TINTE calculations serve as validation of the simulation. The analysis of the primary circuit tries to answer the question how fast the water enters the reactor core. It was found that the maximum H 2 O concentration increase in the reactor core is smaller than 0.3 kg/(m 3 s). The liquid water vaporization in the steam generator and H 2 O transport from the steam generator to the reactor core reduce the ingress velocity of the H 2 O into the reactor core. In order to answer the question how much water enters the primary circuit, the full cavitation of the feed water pumps is analyzed. It is found that if the secondary circuit is depressurized enough, the feed water pumps will be inherently stopped by the full cavitation. This limits the water to be pumped from the deaerator to the steam generator. A comprehensive simulation of the MODUL-HTR power plant then shows that the H 2 O inventory in the primary circuit can be limited to about 3000 kg. The nuclear reactivity increase caused by the water ingress leads to a fast power excursion, which, however, is inherently counterbalanced by negative feedback effects. Concerning the integrity of the fuel elements, the safety relevant temperature limit of 1600 C was not reached in any case. (orig.) [de

Integrated design approach of the pebble BeD modular reactor using models

International Nuclear Information System (INIS)

Venter, Pieter J.; Mitchell, Mark N.

2007-01-01

The pebble bed modular reactor (PBMR) is the first pebble bed reactor that will be utilised in a high temperature direct Brayton cycle configuration. This implies that there are a number of unique features in the PBMR that extend from the German experience base. One of the challenges in the design of the PBMR is developing an understanding of the expected behaviour of the reactor through analyses and simulations and managing the integrated design process between the designers, the physicists and the analysts. This integrated design process is managed through model-based development work. Three-dimensional CAD models are constructed of the components and parts in the reactor. From the CAD models, CFD models, neutronic models, shielding models, FEM models and other thermodynamic models are derived. These models range from very simple models to extremely detailed and complex models. The models are used in legacy software as well as commercial off-the-shelf software. The different models are also used in code-to-code comparisons to verify the results. This paper will briefly discuss the different models and the interaction between the models, and how the models are used in the iterative design process that is used in the development of the reactor at PBMR

The modular high-temperature gas-cooled reactor: A cost/risk competitive nuclear option

International Nuclear Information System (INIS)

Gotschall, H.L.

1994-01-01

The business risks of nuclear plant ownership are identified as a constraint on the expanded use of nuclear power. Such risks stem from the exacting demands placed on owner/operator organizations of current plants to demonstrate ongoing compliance with safety regulations and the resulting high costs for operation and maintenance. This paper describes the Modular High-Temperature Gas-Cooled Reactor (MHTGR) design, competitive economics, and approach to reducing the business risks of nuclear plant ownership

Parametric systems analysis of the Modular Stellarator Reactor (MSR)

International Nuclear Information System (INIS)

Miller, R.L.; Krakowski, R.A.; Bathke, C.G.

1982-05-01

The close coupling in the stellarator/torsatron/heliotron (S/T/H) between coil design (peak field, current density, forces), magnetics topology (transform, shear, well depth), and plasma performance (equilibrium, stability, transport, beta) complicates the reactor assessment more so than for most magnetic confinement systems. In order to provide an additional degree of resolution of this problem for the Modular Stellarator Reactor (MSR), a parametric systems model has been developed and applied. This model reduces key issues associted ith plasma performance, first-wall/blanket/shield (FW/B/S), and coil design to a simple relationship between beta, system geometry, and a number of indicators of overall plant performance. The results of this analysis can then be used to guide more detailed, multidimensional plasma, magnetics, and coil design efforts towards technically and economically viable operating regimes. In general, it is shown that beta values > 0.08 may be needed if the MSR approach is to be substantially competitive with other approaches to magnetic fusion in terms of system power density, mass utilization, and cost for total power output around 4.0 GWt; lower powers will require even higher betas

Concept of a HTR modular plant for generation of process heat in a chemical plant

International Nuclear Information System (INIS)

1991-07-01

This final report summarizes the results of a preliminary study on behalf of Buna AG and Leunawerke AG. With regard to the individual situations the study investigated the conditions for modular HTR-2 reactors to cover on-site process heat and electric power demands. HTR-2 reactor erection and operation were analyzed for their economic efficiency compared with fossil-fuel power plants. Considering the prospective product lines, the technical and economic conditions were developed in close cooperation with Buna AG and Leunawerke AG. The study focused on the technical integration of modular HTR reactors into plants with regard to safety concepts, on planning, acceptance and erection concepts which largely exclude uncalculable scheduling and financial risks, and on comparative economic analyses with regard to fossil-fuel power plants. (orig.) [de

Inherent controllability in modular ALMRs

International Nuclear Information System (INIS)

Sackett, J.I.; Sevy, R.H.; Wei, T.Y.C.

1989-01-01

As part of recent development efforts on advanced reactor designs ANL has proposed the IFR (Integral Fast Reactor) concept. The IFR concept is currently being applied to modular sized reactors which would be built in multiple power paks together with an integrated fuel cycle facility. It has been amply demonstrated that the concept as applied to the modular designs has significant advantages in regard to ATWS transients. Attention is now being focussed on determining whether or not those advantages deriving from the traits of the IFR can be translated to the operational/DBA (design basis accident) class of transients. 5 refs., 3 figs., 3 tabs

Advanced Small Modular Reactor Economics Status Report

Energy Technology Data Exchange (ETDEWEB)

Harrison, Thomas J [ORNL

2014-10-01

This report describes the data collection work performed for an advanced small modular reactor (AdvSMR) economics analysis activity at the Oak Ridge National Laboratory. The methodology development and analytical results are described in separate, stand-alone documents as listed in the references. The economics analysis effort for the AdvSMR program combines the technical and fuel cycle aspects of advanced (non-light water reactor [LWR]) reactors with the market and production aspects of SMRs. This requires the collection, analysis, and synthesis of multiple unrelated and potentially high-uncertainty data sets from a wide range of data sources. Further, the nature of both economic and nuclear technology analysis requires at least a minor attempt at prediction and prognostication, and the far-term horizon for deployment of advanced nuclear systems introduces more uncertainty. Energy market uncertainty, especially the electricity market, is the result of the integration of commodity prices, demand fluctuation, and generation competition, as easily seen in deregulated markets. Depending on current or projected values for any of these factors, the economic attractiveness of any power plant construction project can change yearly or quarterly. For long-lead construction projects such as nuclear power plants, this uncertainty generates an implied and inherent risk for potential nuclear power plant owners and operators. The uncertainty in nuclear reactor and fuel cycle costs is in some respects better understood and quantified than the energy market uncertainty. The LWR-based fuel cycle has a long commercial history to use as its basis for cost estimation, and the current activities in LWR construction provide a reliable baseline for estimates for similar efforts. However, for advanced systems, the estimates and their associated uncertainties are based on forward-looking assumptions for performance after the system has been built and has achieved commercial operation

A modular optical sensor

Science.gov (United States)

Conklin, John Albert

This dissertation presents the design of a modular, fiber-optic sensor and the results obtained from testing the modular sensor. The modular fiber-optic sensor is constructed in such manner that the sensor diaphragm can be replaced with different configurations to detect numerous physical phenomena. Additionally, different fiber-optic detection systems can be attached to the sensor. Initially, the modular sensor was developed to be used by university of students to investigate realistic optical sensors and detection systems to prepare for advance studies of micro-optical mechanical systems (MOMS). The design accomplishes this by doing two things. First, the design significantly lowers the costs associated with studying optical sensors by modularizing the sensor design. Second, the sensor broadens the number of physical phenomena that students can apply optical sensing techniques to in a fiber optics sensor course. The dissertation is divided into seven chapters covering the historical development of fiber-optic sensors, a theoretical overview of fiber-optic sensors, the design, fabrication, and the testing of the modular sensor developed in the course of this work. Chapter 1 discusses, in detail, how this dissertation is organized and states the purpose of the dissertation. Chapter 2 presents an historical overview of the development of optical fibers, optical pressure sensors, and fibers, optical pressure sensors, and optical microphones. Chapter 3 reviews the theory of multi-fiber optic detection systems, optical microphones, and pressure sensors. Chapter 4 presents the design details of the modular, optical sensor. Chapter 5 delves into how the modular sensor is fabricated and how the detection systems are constructed. Chapter 6 presents the data collected from the microphone and pressure sensor configurations of the modular sensor. Finally, Chapter 7 discusses the data collected and draws conclusions about the design based on the data collected. Chapter 7 also

Small Modular Reactors: Nuclear Energy Market Potential for Near-term Deployment

International Nuclear Information System (INIS)

Lokhov, Alexey; Sozoniuk, Vladislav; Rothwell, Geoffrey; ); Cometto, Marco; Paillere, Henri; ); Crozat, Matt; Genoa, Paul; Joon Kim, Tae; McGough, Mike; Ingersoll, Dan; Rickman, Robin; Stout, Dan; Halnon, Greg; Chenais, Jacques; Briffod, Francois-Xavier; Perrier, Sylvain; Shahrokhi, Farshid; Kaufer, Barry; Wasylyk, Andrew; Shropshire, David; ); Danrong, Song; Swinburn, Richard

2016-01-01

Recent interest in small modular reactors (SMRs) is being driven by a desire to reduce the total capital costs associated with nuclear power plants and to provide power to small grid systems. According to estimates available today, if all the competitive advantages of SMRs were realised, including serial production, optimised supply chains and smaller financing costs, SMRs could be expected to have lower absolute and specific (per-kWe) construction costs than large reactors. Although the economic parameters of SMRs are not yet fully determined, a potential market exists for this technology, particularly in energy mixes with large shares of renewables. This report assesses the size of the market for SMRs that are currently being developed and that have the potential to broaden the ways of deploying nuclear power in different parts of the world. The study focuses on light water SMRs that are expected to be constructed in the coming decades and that strongly rely on serial, factory-based production of reactor modules. In a high-case scenario, up to 21 GWe of SMRs could be added globally by 2035, representing approximately 3% of total installed nuclear capacity. (authors)

A HUMAN AUTOMATION INTERACTION CONCEPT FOR A SMALL MODULAR REACTOR CONTROL ROOM

Energy Technology Data Exchange (ETDEWEB)

Le Blanc, Katya; Spielman, Zach; Hill, Rachael

2017-06-01

Many advanced nuclear power plant (NPP) designs incorporate higher degrees of automation than the existing fleet of NPPs. Automation is being introduced or proposed in NPPs through a wide variety of systems and technologies, such as advanced displays, computer-based procedures, advanced alarm systems, and computerized operator support systems. Additionally, many new reactor concepts, both full scale and small modular reactors, are proposing increased automation and reduced staffing as part of their concept of operations. However, research consistently finds that there is a fundamental tradeoff between system performance with increased automation and reduced human performance. There is a need to address the question of how to achieve high performance and efficiency of high levels of automation without degrading human performance. One example of a new NPP concept that will utilize greater degrees of automation is the SMR concept from NuScale Power. The NuScale Power design requires 12 modular units to be operated in one single control room, which leads to a need for higher degrees of automation in the control room. Idaho National Laboratory (INL) researchers and NuScale Power human factors and operations staff are working on a collaborative project to address the human performance challenges of increased automation and to determine the principles that lead to optimal performance in highly automated systems. This paper will describe this concept in detail and will describe an experimental test of the concept. The benefits and challenges of the approach will be discussed.

MOSRA-SRAC. Lattice calculation module of the modular code system for nuclear reactor analyses MOSRA

International Nuclear Information System (INIS)

Okumura, Keisuke

2015-10-01

MOSRA-SRAC is a lattice calculation module of the Modular code System for nuclear Reactor Analyses (MOSRA). This module performs the neutron transport calculation for various types of fuel elements including existing light water reactors, research reactors, etc. based on the collision probability method with a set of the 200-group cross-sections generated from the Japanese Evaluated Nuclear Data Library JENDL-4.0. It has also a function of the isotope generation and depletion calculation for up to 234 nuclides in each fuel material in the lattice. In these ways, MOSRA-SRAC prepares the burn-up dependent effective microscopic and macroscopic cross-section data to be used in core calculations. A CD-ROM is attached as an appendix. (J.P.N.)

RHEIN, Modular System for Reactor Design Calculation

International Nuclear Information System (INIS)

Reiche, Christian; Barz, Hansulrich; Kunzmann, Bernd; Seifert, Eberhard; Wand, Hartmut

1990-01-01

1 - Description of program or function: RHEIN is a modular reactor code system for neutron physics calculations. It consists of a small number of system codes for execution control, data management, and handling support, as well as of the physical calculation routines. The execution is controlled by input data containing mathematical and physical parameters and simple commands for routine calls and data manipulations. The calculation routines are in tune with one another and the system takes care of the data transfer between them. Cross-section libraries with self shielding parameters are added to the system. 2 - Method of solution: The calculation routines can be used for solving the following physics problems: - Calculation of cross-section sets for infinite mediums, taking into account chord length. - Zero-dimensional spectrum calculation in diffusion, P1, or B1 approximation. - One-dimensional calculation in diffusion, P1, or collision probability approximation. - Two-dimensional diffusion calculation. - Cell calculation by THERMOS. - Zone-wise homogenized group collapsing within zero, one, or two-dimensional models. - Normalization, summarizing, etc. - Output of cross-section sets to off systems Sn and Monte-Carlo calculations

Multi-purpose reactor

International Nuclear Information System (INIS)

1991-05-01

The Multi-Purpose-Reactor (MPR), is a pool-type reactor with an open water surface and variable core arrangement. Its main feature is plant safety and reliability. Its power is 22MW t h, cooled by light water and moderated by beryllium. It has platetype fuel elements (MTR type, approx. 20%. enriched uranium) clad in aluminium. Its cobalt (Co 60 ) production capacity is 50000 Ci/yr, 200 Ci/gr. The distribution of the reactor core and associated control and safety systems is essentially based on the following design criteria: - upwards cooling flow, to waive the need for cooling flow inversion in case the reactor is cooled by natural convection if confronted with a loss of pumping power, and in order to establish a superior heat transfer potential (a higher coolant saturation temperature); - easy access to the reactor core from top of pool level with the reactor operating at full power, in order to facilitate actual implementation of experiments. Consequently, mechanisms associated to control and safety rods s,re located underneath the reactor tank; - free access of reactor personnel to top of pool level with the reactor operating at full power. This aids in the training of personnel and the actual carrying out of experiments, hence: - a vast water column was placed over the core to act as radiation shielding; - the core's external area is cooled by a downwards flow which leads to a decay tank beyond the pool (for N 16 to decay); - a small downwards flow was directed to stream downwards from above the reactor core in order to drag along any possibly active element; and - a stagnant hot layer system was placed at top of pool level so as to minimize the upwards coolant flow rising towards pool level

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

International Nuclear Information System (INIS)

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

2003-01-01

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

Extending the Candu Nuclear Reactor Concept: The Multi-Spectrum Nuclear Reactor

International Nuclear Information System (INIS)

Allen, Francis; Bonin, Hugues

2008-01-01

The aim of this work is to examine the multi-spectrum nuclear reactor concept as an alternative to fast reactors and accelerator-driven systems for breeding fissile material and reducing the radiotoxicity of spent nuclear fuel. The design characteristics of the CANDU TM nuclear power reactor are shown to provide a basis for a novel approach to this concept. (authors)

Extending the Candu Nuclear Reactor Concept: The Multi-Spectrum Nuclear Reactor

Energy Technology Data Exchange (ETDEWEB)

Allen, Francis [Director General Nuclear Safety, 280 Slater St, Ottawa, K1A OK2 (Canada); Bonin, Hugues [Royal Military College of Canada, 11 General Crerar Cres, Kingston, K7K 7B4 (Canada)

2008-07-01

The aim of this work is to examine the multi-spectrum nuclear reactor concept as an alternative to fast reactors and accelerator-driven systems for breeding fissile material and reducing the radiotoxicity of spent nuclear fuel. The design characteristics of the CANDU{sup TM} nuclear power reactor are shown to provide a basis for a novel approach to this concept. (authors)

Large modular pebble-bed reactors with passive safety properties as a contribution for catastrophe-free nuclear technology. Flexibility in design and application

International Nuclear Information System (INIS)

Eladly, H.

1996-01-01

Worldwide investigations are carried out for different reactor concepts, in order to realize nuclear energy production in modular power plants. In that concept several small or middle sized reactors are joined together in a modular way to form one power plant. The size of MODUL-reactors is designed in such a way, that exclusively inherent safety properties perform the control of accidents without active technical proceedings. In order to achieve this, the reactor should be relatively small. On the other hand, it should be relatively large for economic and competitive reasons. The range of possible development of the modular pebble-bed reactor for raising the power output are discussed in this study. Based on the MODUL 200 MW concept, the design of the 'Great-Modul-Medul' reactor (GMM) with a power output of 500 MWth is introduced, in which the loading modus MEDUL is applied with repeated circulation of the spheres through the core. A 'Great-Modul-OTTO' GMO with a power output of 400 MWth is designed with only one pass of the pebbles (OTTO). In comparison to the GMM, that has the advantage of being simpler in construction and in the method of operation. Furthermore, another simplification is studied consisting of the combination (PO) of 'Peu a Peu' and 'OTTO' loading modus. All designed cases show a favourable flexibility when changing the application of the reactor from steam cycle to gas turbine cycle or to seawater desalination. The study outlines, that the inherently determined limitation of the excess temperature in case of a loss coolant accident and the ability for controling the water ingress reactivity are maintained for all variants being considered. (orig.) [de

A Conceptual Study on a Supercritical CO_2-cooled Micro Modular Reactor

International Nuclear Information System (INIS)

Yu, Hwanyeal; Hartanto, Donny; Kim, Yonghee

2014-01-01

A Micro Modular Reactor (MMR) using Supercritical-CO_2 (S-CO_2) as coolant has been investigated from the neutronics perspective. The MMR is designed to be transportable so it can reach the remote areas. The thermal power of the reactor is 36.2 M Wth. The size of the active core is limited to 1.2 m length and 93.16 cm width. The size of whole core is 2.8 m length and 166.9 cm width. The reactor lifetime design target is 20 years. To maximize the fuel volume fraction in the core, high density uranium nitride UN"1"5 was used. The PbO/MgO reflector was also utilized to improve the neutron economy. The S-CO_2 is chosen as the coolant because it offers a higher thermal efficiency. In this study, neutronics calculations and depletion using McCARD Monte Carlo code has been done to determine the lifetime and behavior of the core. Several important safety parameters such as Control Rod worth, Doppler reactivity coefficients and coolant void reactivity coefficient have also been analyzed. (author)

Hypothetical air ingress scenarios in advanced modular high temperature gas cooled reactors

International Nuclear Information System (INIS)

Kroeger, P.G.

1988-01-01

Considering an extremely hypothetical scenario of complete cross duct failure and unlimited air supply into the reactor vessel of a modular high temperature gas cooled ractor, it is found that the potential air inflow remains limited due to the high friction pressure drop through the active core. All incoming air will be oxidized to CO and some local external burning would be temporarily possible in such a scenario. The accident would have to continue with unlimited air supply for hundreds of hours before the core structural integrity would be jeopardized

Safety aspects of forced flow cooldown transients in modular high temperature gas-cooled reactors

International Nuclear Information System (INIS)

Kroeger, P.G.

1992-01-01

During some of the design basis accidents in Modular High Temperature Gas Cooled Reactors (MHTGRs) the main Heat Transport System (HTS) and the Shutdown Cooling System (SCS), are assumed to have failed. Decay heat is then removed by the passive Reactor Cavity Cooling System (RCCS) only. If either forced flow cooling system becomes available during such a transient, its restart could significantly reduce the down-time. This paper uses the THATCH code to examine whether such restart, during a period of elevated core temperatures, can be accomplished within safe limits for fuel and metal component temperatures. If the reactor is scrammed, either system can apparently be restarted at any time, without exceeding any safe limits. However, under unscrammed conditions a restart of forced cooling can lead to recriticality, with fuel and metal temperatures significantly exceeding the safety limits

Closed-loop digital control of nuclear reactors characterized by spatial dynamics

International Nuclear Information System (INIS)

Bernard, J.A.; Henry, A.F.; Lanning, D.D.; Meyer, J.E.

1991-03-01

This report describes the theoretical development and the evaluation via both simulation and, to a lesser degree, experiment of a digital method for the closed-loop control of power and temperature in reactors characterized by spatial dynamics. The major conclusions of the research are that (1) the sophistication of advanced reactor physics and thermal-hydraulic nodal methods is now such that accurate, real-time models of spatially-dependent, heterogeneous reactor cores can be run on present-generation minicomputers; (2) operation of both present-day commercial reactors as well as the multi-modular reactors now being considered for construction in the United States could be significantly improved by incorporating model-generated information on in-core conditions in a digital controller; and (3) digital controllers for spatially-dependent reactors should have a hierarchical or multi-tiered structure consisting of supervisory algorithms that preclude challenges to the safety system, global control laws designed to provide an optimal response to temperature and power perturbations, and local control laws that maintain parameters such as the margin to departure from nucleate boiling within specification. The technology described is appropriate to present-day pressurized water reactors and to the proposed multi-modular designs. The end-product of this research was a (near) real-time analytic plant-estimation code that was given the acronym POPSICLE for POwer Plant SImulator and ControlLEr. POPSICLE's core neutronics model is based on a quasi-static transient solution of the analytic nodal diffusion equations. 126 refs., 159 figs., 17 tabs

Closed-loop digital control of nuclear reactors characterized by spatial dynamics

Energy Technology Data Exchange (ETDEWEB)

Bernard, J.A. (Massachusetts Inst. of Tech., Cambridge, MA (USA). Nuclear Reactor Lab.); Henry, A.F.; Lanning, D.D.; Meyer, J.E. (Massachusetts Inst. of Tech., Cambridge, MA (USA). Dept. of Nuclear Engineering)

1991-03-01

This report describes the theoretical development and the evaluation via both simulation and, to a lesser degree, experiment of a digital method for the closed-loop control of power and temperature in reactors characterized by spatial dynamics. The major conclusions of the research are that (1) the sophistication of advanced reactor physics and thermal-hydraulic nodal methods is now such that accurate, real-time models of spatially-dependent, heterogeneous reactor cores can be run on present-generation minicomputers; (2) operation of both present-day commercial reactors as well as the multi-modular reactors now being considered for construction in the United States could be significantly improved by incorporating model-generated information on in-core conditions in a digital controller; and (3) digital controllers for spatially-dependent reactors should have a hierarchical or multi-tiered structure consisting of supervisory algorithms that preclude challenges to the safety system, global control laws designed to provide an optimal response to temperature and power perturbations, and local control laws that maintain parameters such as the margin to departure from nucleate boiling within specification. The technology described is appropriate to present-day pressurized water reactors and to the proposed multi-modular designs. The end-product of this research was a (near) real-time analytic plant-estimation code that was given the acronym POPSICLE for POwer Plant SImulator and ControlLEr. POPSICLE's core neutronics model is based on a quasi-static transient solution of the analytic nodal diffusion equations. 126 refs., 159 figs., 17 tabs.

Investigation of nonplanar modular coil systems for stellarator fusion reactors

International Nuclear Information System (INIS)

Harmeyer, E.

1988-12-01

Steady-state stellarators constitute an important option for a future fusion reactor. The helical magnetic field required for plasma confinement can be produced by means of a set of modular nonplanar coils. In order to achieve optimum power density of the plasma, the magnetic flux density inside the torus is made as high as possible. State-of-the-art estimates allow values of the magnetic flux density on axis of B 0 = 4-7 T. The present report is concerned with investigations on modular nonplanar stellarator coil systems. Coil systems with poloidal periodicity l=2 and a coil system of the W VII-AS type with superposed l=0, 1, 2, 3 terms are treated. Furthermore, the parameters are simultaneously varied while keeping constant the ratios of certain magnitudes. In the parameter space of the geometric values and coil number the following quantities are evaluated: maximum magnetic flux density in the coil domain, stored magnetic energy of the coil system, magnetic force density distribution or magnetic forces, and mechanical stress distribution in the coils. Numerical methods are applied in the programme systems used for these calculations. The aim of the study is to determine an optimum regime for the above parameters. The numerical results are compared with those of analytical approximation solutions. (orig.)

Coupling the modular helium reactor to hydrogen production processes

International Nuclear Information System (INIS)

Richards, M.B.; Shenoy, A.S.; Schultz, K.R.

2004-01-01

Steam reforming of natural gas (methane) currently produces the bulk of hydrogen gas used in the world today. Because this process depletes natural gas resources and generates the greenhouse gas carbon dioxide as a by-product, there is a growing interest in using process heat and/or electricity generated by nuclear reactors to generate hydrogen by splitting water. Process heat from a high temperature nuclear reactor can be used directly to drive a set of chemical reactions, with the net result of splitting water into hydrogen and oxygen. For example, process heat at temperatures in the range 850 deg C to 950 deg C can drive the sulphur-iodine (S-I) thermochemical process to produce hydrogen with high efficiency. The S-I process produces highly pure hydrogen and oxygen, with formation, decomposition, regeneration, and recycle of the intermediate chemical reagents. Electricity can also 1)e used directly to split water, using conventional, low-temperature electrolysis (LTE). Hydrogen can also be produced with hybrid processes that use both process heat and electricity to generate hydrogen. An example of a hybrid process is high-temperature electrolysis (HTE), in which process heat is used to generate steam, which is then supplied to an electrolyzer to generate hydrogen. This process is of interest because the efficiency of electrolysis increases with temperature. Because of its high temperature capability, advanced stage of development relative to other high-temperature reactor concepts, and passive-safety features, the modular helium reactor (MHR) is well suited for producing hydrogen using nuclear energy. In this paper we investigate the coupling of the MHR to the S-I process, LTE, and HTE. These concepts are referred to as the H2-MHR. (author)

Presentation summary: Gas Turbine - Modular Helium Reactor (GT-MHR)

International Nuclear Information System (INIS)

2001-01-01

Numerous prototypes and demonstration plants have been constructed and operated beginning with the Dragon plant in the early 1960s. The MHTGR was the U.S. developed modular plant and underwent pre application review by NRC. The GT-MHR represents a further refinement on this concept with the steam cycle being replaced by a closed loop gas turbine (Brayton) cycle. Modular gas reactors and the GT-MHR represent a fundamental shift in reactor design and safety philosophy. The reactor system is contained in a 3 vessel, side-by-side arrangement. The reactor and a shutdown cooling system are in one vessel, and the gas turbine based power conversion system, including the generator, in a second parallel vessel. A more detailed look at the system shows the compact arrangement of gas turbine, compressors, recuperator, heat exchanges, and generator. Fueled blocks are stacked in three concentric rings with inert graphite blocks making up the inner and outer reflectors. Operating control rods are located outside the active core while startup control rods and channels for reserve shutdown pellets are located near the core center. Ceramic coated fuel is the key to the GT-MHR's safety and economics. A kernel of Uranium oxycarbide (or UO 2 ) is placed in a porous carbon buffer and then encapsulated in multiple layers of pyrolytic carbon and silicon carbide. These micro pressure vessels withstand internal pressures of up to 2,000 psi and temperatures of nearly 2,000 C providing extremely resilient containment of fission products under both normal operating and accident conditions. The fuel particles are blended in carbon pitch, forming fuel rods, and then loaded into holes within large graphite fuel elements. Fuel elements are stacked to form the core. Fuel particle testing in has repeatedly demonstrated the high temperature resilience of coated particle fuel to temperature approaching 2,000 C. As an conservative design goal, GT-MHR has been sized to keep maximum fuel temperatures

A severe accident analysis for the system-integrated modular advanced reactor

International Nuclear Information System (INIS)

Jung, Gunhyo; Jae, Moosung

2015-01-01

The System-Integrated Modular Advanced Reactor (SMART) that has been recently designed in KOREA and has acquired standard design certification from the nuclear power regulatory body (NSSC) is an integral type reactor with 330MW thermal power. It is a small sized reactor in which the core, steam generator, pressurizer, and reactor coolant pump that are in existing pressurized light water reactors are designed to be within a pressure vessel without any separate pipe connection. In addition, this reactor has much different design characteristics from existing pressurized light water reactors such as the adoption of a passive residual heat removal system and a cavity flooding system. Therefore, the safety of the SMART against severe accidents should be checked through severe accident analysis reflecting the design characteristics of the SMART. For severe accident analysis, an analysis model has been developed reflecting the design information presented in the standard design safety analysis report. The severe accident analysis model has been developed using the MELCOR code that is widely used to evaluate pressurized LWR severe accidents. The steady state accident analysis model for the SMART has been simulated. According to the analysis results, the developed model reflecting the design of the SMART is found to be appropriate. Severe accident analysis has been performed for the representative accident scenarios that lead to core damage to check the appropriateness of the severe accident management plan for the SMART. The SMART has been shown to be safe enough to prevent severe accidents by utilizing severe accident management systems such as a containment spray system, a passive hydrogen recombiner, and a cavity flooding system. In addition, the SMART is judged to have been technically improved remarkably compared to existing PWRs. The SMART has been designed to have a larger reactor coolant inventory compared to its core's thermal power, a large surface area in

A Small Modular Reactor Design for Multiple Energy Applications: HTR50S

Energy Technology Data Exchange (ETDEWEB)

Yan, X.; Tachibana, Y.; Ohashi, H.; Sato, H.; Tazawa, Y.; Kunitomi, K. [Japan Atomic Energy Agency, Ibaraki (Japan)

2013-06-15

HTR50S is a small modular reactor system based on HTGR. It is designed for a triad of applications to be implemented in successive stages. In the first stage, a base plant for heat and power is constructed of the fuel proven in JAEA's 950 .deg. C, 30MWt test reactor HTTR and a conventional steam turbine to minimize development risk. While the outlet temperature is lowered to 750 .deg. C for the steam turbine, thermal power is raised to 50MWt by enabling 40% greater power density in 20% taller core than the HTTR. However the fuel temperature limit and reactor pressure vessel diameter are kept. In second stage, a new fuel that is currently under development at JAEA will allow the core outlet temperature to be raised to 900 .deg. C for the purpose of demonstrating more efficient gas turbine power generation and high temperature heat supply. The third stage adds a demonstration of nuclear-heated hydrogen production by a thermochemical process. A licensing approach to coupling high temperature industrial process to nuclear reactor will be developed. The low initial risk and the high longer-term potential for performance expansion attract development of the HTR50S as a multipurpose industrial or distributed energy source.

An Overview of the Safety Case for Small Modular Reactors

International Nuclear Information System (INIS)

Ingersoll, Daniel T.

2011-01-01

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

Multi-echelon Repair Inventory Systems: Select Issues in Modular Electronic Equipment

OpenAIRE

Amik Garg; S. G. Deshmukh

2010-01-01

Flow of modules/printed circuit boards (PCBs) in a multi-echelon repair inventory system pertaining to modular electronic equipment for large maintenance organisations having large inventory in range and depth, like defence has been critically examined using a case study. Desirable features of the proposed system are identified and a general framework suggested for examining its feasibility and implementation in an organisation. An analytical model with an objective to reduce number of echelo...

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

International Nuclear Information System (INIS)

Araujo, Nathália N.; Su, Jian

2017-01-01

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

CFD Analysis for Hot Spot Fuel Temperature of Deep-Burn Modular Helium Reactor

International Nuclear Information System (INIS)

Tak, Nam Il; Jo, Chang Keun; Jun, Ji Su; Kim, Min Hwan; Venneri, Francesco

2009-01-01

As an alternative concept of a conventional transmutation using fast reactors, a deep-burn modular helium reactor (DB-MHR) concept has been proposed by General Atomics (GA). Kim and Venneri published an optimization study on the DB-MHR core in terms of nuclear design. The authors concluded that more concrete evaluations are necessary including thermo-fluid and safety analysis. The present paper describes the evaluation of the hot spot fuel temperature of the fuel assembly in the 600MWth DB-MHR core under full operating power conditions. Two types of fuel shuffling scheme (radial and axial hybrid shuffling and axial-only shuffling) are investigated. For accurate thermo-fluid analysis, the computational fluid dynamics (CFD) analysis has been performed on a 1/12 fuel assembly using the CFX code

A Multi-Year Plan for Research, Development, and Prototype Testing of Standard Modular Hydropower Technology

Energy Technology Data Exchange (ETDEWEB)

Smith, Brennan T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Welch, Tim [U.S. Department of Energy (DOE), Washington, DC (United States).Office of Energy Efficiency and Renewable Energy (EERE); Witt, Adam M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Stewart, Kevin M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Lee, Kyutae [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); DeNeale, Scott T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bevelhimer, Mark S. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Burress, Timothy A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pracheil, Brenda M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pries, Jason L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); O' Connor, Patrick W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Curd, Shelaine L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ekici, Kivanc [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Univ. of Tennessee, Knoxville, TN (United States); Papanicolaou, Thanos [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Univ. of Tennessee, Knoxville, TN (United States); Tsakiris, Achilleas [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Univ. of Tennessee, Knoxville, TN (United States); Kutz, Benjamin [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Univ. of Tennessee, Knoxville, TN (United States); Bishop, Norm [Knight Piesold, Denver, CO (United States); McKeown, Alisha [McKeown and Associates, Moberly, MO (United States); Rabon, Daniel [U.S. Department of Energy (DOE), Washington, DC (United States).Office of Energy Efficiency and Renewable Energy (EERE); Zimmerman, Gregory P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Uria Martinez, Rocio [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); McManamay, Ryan A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-02-01

The Multi-Year Plan for Research, Development, and Prototype Testing of Standard Modular Hydropower Technology (MYRP) presents a strategy for specifying, designing, testing, and demonstrating the efficacy of standard modular hydropower (SMH) as an environmentally compatible and cost-optimized renewable electricity generation technology. The MYRP provides the context, background, and vision for testing the SMH hypothesis: if standardization, modularity, and preservation of stream functionality become essential and fully realized features of hydropower technology, project design, and regulatory processes, they will enable previously unrealized levels of new project development with increased acceptance, reduced costs, increased predictability of outcomes, and increased value to stakeholders. To achieve success in this effort, the MYRP outlines a framework of stakeholder-validated criteria, models, design tools, testing facilities, and assessment protocols that will facilitate the development of next-generation hydropower technologies.

Preliminary analysis of combined cycle of modular high-temperature gas cooled reactor

International Nuclear Information System (INIS)

Baogang, Z.; Xiaoyong, Y.; Jie, W.; Gang, Z.; Qian, S.

2015-01-01

Modular high-temperature gas cooled reactor (HTGR) is known as one of the most advanced nuclear reactors because of its inherent safety and high efficiency. The power conversion system of HTGR can be steam turbine based on Rankine cycle or gas turbine based on Brayton cycle respectively. The steam turbine system is mature and the gas turbine system has high efficiency but under development. The Brayton-Rankine combined cycle is an effective way to further promote the efficiency. This paper investigated the performance of combined cycle from the viewpoint of thermodynamics. The effect of non-dimensional parameters on combined cycle’s efficiency, such as temperature ratio, compression ratio, efficiency of compressor, efficiency of turbine, was analyzed. Furthermore, the optimal parameters to achieve highest efficiency was also given by this analysis under engineering constraints. The conclusions could be helpful to the design and development of combined cycle of HTGR. (author)

A Conceptual Study on a Supercritical CO2-cooled KAIST Micro Modular Reactor

International Nuclear Information System (INIS)

Yu, HwanYeal; Hartanto, Donny; Kim, Yonghee

2015-01-01

Small Modular Reactors (SMRs) are nuclear reactors that are completely built in a factory and shipped to the designated site for installation. As such, the SMR is especially advantageous as a flexible and cost-effective energy source for remote and isolated areas. Furthermore, the concept requires a relatively low capital cost, which makes it attractive for developing countries with limited electricity grid. In addition, the SMR concepts also generate more interest after the Fukushima accident since it can easily be designed with a passive decay heat removal system. One of the major advantages of a water-cooled SMR is its relatively small core size. Nonetheless, in spite of its small core size, the volume and area required for its steam-cycle power conversion unit is still significant. In this study, neutronics feasibility of a fully compact and transportable KAIST micro-modular reactor (MMR) was demonstrated. Rated thermal power of the core was 36.2 MWth with total weight of about 39.6 tons. The core was loaded with 15.5 w/o uranium mono-nitride U15N fuels in order to achieve a targeted lifetime of 20 EFPYs. To achieve targeted lifetime, initial excess reactivity of the core should be quite high, around 4,707 pcm. To reduce the high excess reactivity to about 2,500 pcm, a replaceable burnable absorber was utilized in the design. As a result, the MMR has a 20-year lifetime with a relatively small burnup reactivity swing. Several important safety parameters of the KAIST MMR core were also determined in this study. The Doppler reactivity coefficients and CVRs were demonstrated to negative. Worth of the primary control drums and secondary control rod were much higher than initial excess reactivity

Preliminary Design of KAIST Micro Modular Reactor with Dry Air Cooling

Energy Technology Data Exchange (ETDEWEB)

Baik, Seung Joon; Bae, Seong Jun; Kim, Seong Gu; Lee, Jeong Ik [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2014-05-15

KAIST research team recently proposed a Micro Modular Reactor (MMR) concept which integrates power conversion unit (PCU) with the reactor core in a single module. Using supercritical CO{sub 2} as a working fluid of cycle can achieve physically compact size due to small turbomachinery and heat exchangers. The objective of this project is to develop a concept that can operate at isolated area. The design focuses especially on the operation in the inland area where cooling water is insufficient. Thus, in this paper the potential for dry air cooling of the proposed reactor will be examined by sizing the cooling system with preliminary approach. The KAIST MMR is a recently proposed concept of futuristic SMR. The MMR size is being determined to be transportable with land transportation. Special attention is given to the MMR design on the dry cooling, which the cooling system does not depend on water. With appropriately designed air cooling heat exchanger, the MMR can operate autonomously. Two types of air cooling methods are suggested. One is using fan and the other is utilizing cooling tower for the air flow. With fan type air cooling method it consumes about 0.6% of generated electricity from the nuclear reactor. Cooling tower occupies an area of 227 m{sup 2} and 59.6 m in height. This design is just a preliminary estimation of the dry cooling method, and therefore more detailed and optimal design will be followed in the next phase.

Final Report: Self-Consolidating Concrete Construction for Modular Units

Energy Technology Data Exchange (ETDEWEB)

Gentry, Russell [Georgia Inst. of Technology, Atlanta, GA (United States); Kahn, Lawrence [Georgia Inst. of Technology, Atlanta, GA (United States); Kurtis, Kimberly [Georgia Inst. of Technology, Atlanta, GA (United States); Petrovic, Bojan [Georgia Inst. of Technology, Atlanta, GA (United States); Loreto, Giovanni [Georgia Inst. of Technology, Atlanta, GA (United States); Van Wyk, Jurie [Westinghouse Inc., Cranberry Township, PA (United States); Canterero-Leal, Carlos [Westinghouse Inc., Cranberry Township, PA (United States)

2016-07-29

This report focuses on work completed on DE-NE0000667, Self-Consolidating Concrete for Modular Units, in connection with the Department of Energy Nuclear Energy Enabling Technologies (DOE-NEET) program. This project was completed in the School of Civil and Environmental Engineering at the Georgia Institute of Technology, with Westinghouse Corporation as the industrial partner. The primary objective of this project was to develop self-consolidating concrete (also termed “self-compacting concrete” or SCC) mixtures so that concrete placement can be made into steel plate composite (SC) modular structures without the need for continuous concrete placement. As part of the research, SCC mixtures were developed and validated to ensure sufficient shear capacity across cold-joints, while minimizing shrinkage and temperature increase during curing to enhance concrete bonding with the steel plate construction found in modular units. The SCC mixtures developed were able to carry shearing forces across the cold-joint boundaries. This “self-roughening” was achieved by adding a tailored fraction of lightweight aggregate (LWA) to the concrete mix, some of which raised to the surface during curing, forming a rough surface on which subsequent concrete placements were made. The self-roughening behavior was validated through three sets of structural tests. Shear friction on small-scale specimens with cold joints was assessed using varying fractions of LWA and with varying amounts of external steel plate reinforcement. The results show that the shear friction coefficient, to be used with the provisions of ACI 318-14, Section 22.9, can be taken as 1.35. Mid-scale beam tests were completed to assess the cold-joint capacity in both in-plane and out-of-plane bending. The results showed that the self-roughened joints performed as well as monolithic joints. The final assessment was a full-scale test using a steel composite module supplied by Westinghouse and similar in construction to

Final Report: Self-Consolidating Concrete Construction for Modular Units

International Nuclear Information System (INIS)

Gentry, Russell; Kahn, Lawrence; Kurtis, Kimberly; Petrovic, Bojan; Loreto, Giovanni; Van Wyk, Jurie; Canterero-Leal, Carlos

2016-01-01

This report focuses on work completed on DE-NE0000667, Self-Consolidating Concrete for Modular Units, in connection with the Department of Energy Nuclear Energy Enabling Technologies (DOE-NEET) program. This project was completed in the School of Civil and Environmental Engineering at the Georgia Institute of Technology, with Westinghouse Corporation as the industrial partner. The primary objective of this project was to develop self-consolidating concrete (also termed ''self-compacting concrete'' or SCC) mixtures so that concrete placement can be made into steel plate composite (SC) modular structures without the need for continuous concrete placement. As part of the research, SCC mixtures were developed and validated to ensure sufficient shear capacity across cold-joints, while minimizing shrinkage and temperature increase during curing to enhance concrete bonding with the steel plate construction found in modular units. The SCC mixtures developed were able to carry shearing forces across the cold-joint boundaries. This ''self-roughening'' was achieved by adding a tailored fraction of lightweight aggregate (LWA) to the concrete mix, some of which raised to the surface during curing, forming a rough surface on which subsequent concrete placements were made. The self-roughening behavior was validated through three sets of structural tests. Shear friction on small-scale specimens with cold joints was assessed using varying fractions of LWA and with varying amounts of external steel plate reinforcement. The results show that the shear friction coefficient, to be used with the provisions of ACI 318-14, Section 22.9, can be taken as 1.35. Mid-scale beam tests were completed to assess the cold-joint capacity in both in-plane and out-of-plane bending. The results showed that the self-roughened joints performed as well as monolithic joints. The final assessment was a full-scale test using a steel composite module supplied by

Sensitivity studies of air ingress accidents in modular HTGRs

International Nuclear Information System (INIS)

Ball, Syd; Richards, Matt; Shepelev, Sergey

2008-01-01

Postulated air ingress accidents, while of very low probability in a modular high-temperature gas-cooled reactor (HTGR), are of considerable interest to the plant designer, operator, and regulator because of the possibility that the core could sustain significant damage under some circumstances. Sensitivity analyses are described that cover a wide spectrum of conditions affecting outcomes of the postulated accident sequences, for both prismatic and pebble-bed core designs. The major factors affecting potential core damage are the size and location of primary system leaks, flow path resistances, the core temperature distribution, and the long-term availability of oxygen in the incoming gas from a confinement building. Typically, all the incoming oxygen entering the core area is consumed within the reactor vessel, so it is more a matter of where, not whether, oxidation occurs. An air ingress model with example scenarios and means for mitigating damage are described. Representative designs of modular HTGRs included here are a 400-MW(th) pebble-bed reactor (PBR), and a 600-MW(th) prismatic-core modular reactor (PMR) design such as the gas-turbine modular helium reactor (GT-MHR)

Progress and challenges in the development and qualification of multi-level multi-physics coupled methodologies for reactor analysis

International Nuclear Information System (INIS)

Ivanov, K.; Avramova, M.

2007-01-01

Current trends in nuclear power generation and regulation as well as the design of next generation reactor concepts along with the continuing computer technology progress stimulate the development, qualification and application of multi-physics multi-scale coupled code systems. The efforts have been focused on extending the analysis capabilities by coupling models, which simulate different phenomena or system components, as well as on refining the scale and level of detail of the coupling. This paper reviews the progress made in this area and outlines the remaining challenges. The discussion is illustrated with examples based on neutronics/thermohydraulics coupling in the reactor core modeling. In both fields recent advances and developments are towards more physics-based high-fidelity simulations, which require implementation of improved and flexible coupling methodologies. First, the progresses in coupling of different physics codes along with the advances in multi-level techniques for coupled code simulations are discussed. Second, the issues related to the consistent qualification of coupled multi-physics and multi-scale code systems for design and safety evaluation are presented. The increased importance of uncertainty and sensitivity analysis are discussed along with approaches to propagate the uncertainty quantification between the codes. The incoming OECD LWR Uncertainty Analysis in Modeling (UAM) benchmark is the first international activity to address this issue and it is described in the paper. Finally, the remaining challenges with multi-physics coupling are outlined. (authors)

Progress and challenges in the development and qualification of multi-level multi-physics coupled methodologies for reactor analysis

Energy Technology Data Exchange (ETDEWEB)

Ivanov, K.; Avramova, M. [Pennsylvania State Univ., University Park, PA (United States)

2007-07-01

Current trends in nuclear power generation and regulation as well as the design of next generation reactor concepts along with the continuing computer technology progress stimulate the development, qualification and application of multi-physics multi-scale coupled code systems. The efforts have been focused on extending the analysis capabilities by coupling models, which simulate different phenomena or system components, as well as on refining the scale and level of detail of the coupling. This paper reviews the progress made in this area and outlines the remaining challenges. The discussion is illustrated with examples based on neutronics/thermohydraulics coupling in the reactor core modeling. In both fields recent advances and developments are towards more physics-based high-fidelity simulations, which require implementation of improved and flexible coupling methodologies. First, the progresses in coupling of different physics codes along with the advances in multi-level techniques for coupled code simulations are discussed. Second, the issues related to the consistent qualification of coupled multi-physics and multi-scale code systems for design and safety evaluation are presented. The increased importance of uncertainty and sensitivity analysis are discussed along with approaches to propagate the uncertainty quantification between the codes. The incoming OECD LWR Uncertainty Analysis in Modeling (UAM) benchmark is the first international activity to address this issue and it is described in the paper. Finally, the remaining challenges with multi-physics coupling are outlined. (authors)

Hydrogen Process Coupling to Modular Helium Reactors

International Nuclear Information System (INIS)

Shenoy, Arkal; Richards, Matt; Buckingham, Robert

2009-01-01

The U.S. Department of Energy (DOE) has selected the helium-cooled High Temperature Gas-Cooled Reactor (HTGR) as the concept to be used for the Next Generation Nuclear Plant (NGNP), because it is the most advanced Generation IV concept with the capability to provide process heat at sufficiently high temperatures for production of hydrogen with high thermal efficiency. Concurrently with the NGNP program, the Nuclear Hydrogen Initiative (NHI) was established to develop hydrogen production technologies that are compatible with advanced nuclear systems and do not produce greenhouse gases. The current DOE schedule for the NGNP Project calls for startup of the NGNP plant by 2021. The General Atomics (GA) NGNP pre-conceptual design is based on the GA Gas Turbine Modular Helium Reactor (GT-MHR), which utilizes a direct Brayton cycle Power Conversion System (PCS) to produce electricity with a thermal efficiency of 48%. The nuclear heat source for the NGNP consists of a single 600-MW(t) MHR module with two primary coolant loops for transport of the high-temperature helium exiting the reactor core to a direct cycle PCS for electricity generation and to an Intermediate Heat Exchanger (IHX) for hydrogen production. The GA NGNP concept is designed to demonstrate hydrogen production using both the thermochemical sulfur-iodine (SI) process and high-temperature electrolysis (HTE). The two primary coolant loops can be operated independently or in parallel. The reactor design is essentially the same as that for the GT-MHR, but includes the additional primary coolant loop to transport heat to the IHX and other modifications to allow operation with a reactor outlet helium temperature of 950 .deg. C (vs. 850 .deg. C for the GT-MHR). The IHX transfers a nominal 65 MW(t) to the secondary heat transport loop that provides the high-temperature heat required by the SI-based and HTE-based hydrogen production facilities. Two commercial nuclear hydrogen plant variations were evaluated with

Pebble bed modular reactor fuel enrichment discrimination using delayed neutrons - HTR2008-58133

International Nuclear Information System (INIS)

Skoda, R.; Rataj, J.; Uhera, J.

2008-01-01

The Pebble Bed Modular Reactor (PBMR) is a helium-cooled, graphite-moderated high temperature nuclear power reactor which utilise fuel in form of spheres that are randomly loaded and continuously circulated through the core until they reach their prescribed end-of-life burn-up limit. When the reactor is started up for the first time, the lower-enriched start-up fuel is used, mixed with graphite spheres, to bring the core to criticality. As the core criticality is established and the start-up fuel is burned-in, the graphite spheres are progressively removed and replaced with more start-up fuel. Once it becomes necessary for maintaining power output, the higher enriched equilibrium fuel is introduced to the reactor and the start-up fuel is removed. During the initial run of the reactor it is important to discriminate between the irradiated startup fuel and the irradiated equilibrium fuel to ensure that only the equilibrium fuel is returned to the reactor. There is therefore a need for an on-line enrichment discrimination device that can discriminate between irradiated start-up fuel spheres and irradiated equilibrium fuel spheres. The device must also not be confused by the presence of any remaining graphite spheres. Due to it's on-line nature the device must accomplish the discrimination within tight time limits. Theoretical calculations and experiments show that Fuel Enrichment Discrimination based on delayed neutrons detection is possible. The paper presents calculations and experiments showing viability of the method. (authors)

Development and Deployment Strategy for a Small Advanced Light Water Reactor

International Nuclear Information System (INIS)

Modro, S. Michael; Reith, Raymond; Babka, Pierre

2002-01-01

This paper discusses development and deployment strategies for the modular Multi-Application Small Light Water Reactor (MASLWR). Modularity, small size, capability to transport whole modules including containment on road or by rail, simplicity and safety of this reactor allows innovative deployment strategies for a variety of applications. A larger plant may be constructed of many independent power generation units. The multi-module plant is intended to be operated as a base-load plant. Each reactor is to be operated at full load. However, in response to changes in power demand individual units can brought on line or shut down. A larger plant can be built in small increments to match the power demand balancing capital commitments with revenues from sales of electricity. Also, an unplanned shutdown of a reactor only affects a relatively small portion of the total plant capacity. Simplification of MASLWR design and extensive use of modularization coupled with factory fabrication will result in improved productivity of fieldwork and improved quality achieved in a factory environment. The initial MASLWR design concept development has been completed under the U.S. DOE (Department of Energy) Nuclear Energy Research Initiative (NERI) project. This paper discusses a strategy for developing and deploying a MASLWR plant by 2015. This schedule is realistic because the plant design relies on existing industrial experience and manufacturing capabilities. The development strategy consists of the following elements: concept confirmation through testing (under the NERI program a scaled integral test facility has been constructed and initial testing performed), design concept optimization, and design certification based on prototype testing. (authors)

Hypothetical accident scenario analyses for a 250-MW(t) modular high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Harrington, R.M.; Ball, S.J.; Cleveland, J.C.

1985-11-01

This paper describes calculations performed to characterize the inherent safety of a 250-MW(t), 100-MW(e), pebble bed modular high temperature gas-cooled reactor (HTGR) design with vertical in-line arrangement (i.e., upflow core with steam generators directly above the core). A variety of postulated accident sequences involving combinations of loss of forced primary coolant (helium) circulation, loss of primary coolant pressurization, and loss of heat sink were studied and were discussed

Service Modularity and Architecture

DEFF Research Database (Denmark)

Brax, Saara A.; Bask, Anu; Hsuan, Juliana

2017-01-01

, platform-based and mass-customized service business models, comparative research designs, customer perspectives and service experience, performance in context of modular services, empirical evidence of benefits and challenges, architectural innovation in services, modularization in multi-provider contexts......Purpose: Services are highly important in a world economy which has increasingly become service driven. There is a growing need to better understand the possibilities for, and requirements of, designing modular service architectures. The purpose of this paper is to elaborate on the roots...... of the emerging research stream on service modularity, provide a concise overview of existing work on the subject, and outline an agenda for future research on service modularity and architecture. The articles in the special issue offer four diverse sets of research on service modularity and architecture. Design...

Advanced reactors transition fiscal year 1995 multi-year program plan WBS 7.3

International Nuclear Information System (INIS)

Loika, E.F.

1994-01-01

This document describes in detail the work to be accomplished in FY-1995 and the out years for the Advanced Reactors Transition (WBS 7.3). This document describes specific milestones and funding profiles. Based upon the Fiscal Year 1995 Multi-Year Program Plan, DOE will provide authorization to perform the work outlined in the FY 1995 MYPP. Following direction given by the US Department of Energy (DOE) on December 15, 1993, Advanced Reactors Transition (ART), previously known as Advanced Reactors, will provide the planning and perform the necessary activities for placing the Fast Flux Test Facility (FFTF) in a radiologically and industrially safe shutdown condition. The DOE goal is to accomplish the shutdown in approximately five years. The Advanced Reactors Transition Multi-Year Program Plan, and the supporting documents; i.e., the FFTF Shutdown Program Plan and the FFTF Shutdown Project Resource Loaded Schedule (RLS), are defined for the life of the Program. During the transition period to achieve the Shutdown end-state, the facilities and systems will continue to be maintained in a safe and environmentally sound condition. Additionally, facilities that were associated with the Office of Nuclear Energy (NE) Programs, and are no longer required to support the Liquid Metal Reactor Program will be deactivated and transferred to an alternate sponsor or the Decontamination and Decommissioning (D and D) Program for final disposition, as appropriate

The effects of stainless steel radial reflector on core reactivity for small modular reactor

Energy Technology Data Exchange (ETDEWEB)

Kang, Jung Kil, E-mail: jkkang@email.kings.ac.kr; Hah, Chang Joo, E-mail: changhah@kings.ac.kr [KINGS, 658-91, Haemaji-ro, Seosaeng-myeon, Ulju-gun, Ulsan, 689-882 (Korea, Republic of); Cho, Sung Ju, E-mail: sungju@knfc.co.kr; Seong, Ki Bong, E-mail: kbseong@knfc.co.kr [KNFC, Daedeok-daero 989beon-gil, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of)

2016-01-22

Commercial PWR core is surrounded by a radial reflector, which consists of a baffle and water. Radial reflector is designed to reflect neutron back into the core region to improve the neutron efficiency of the reactor and to protect the reactor vessels from the embrittling effects caused by irradiation during power operation. Reflector also helps to flatten the neutron flux and power distributions in the reactor core. The conceptual nuclear design for boron-free small modular reactor (SMR) under development in Korea requires to have the cycle length of 4∼5 years, rated power of 180 MWth and enrichment less than 5 w/o. The aim of this paper is to analyze the effects of stainless steel radial reflector on the performance of the SMR using UO{sub 2} fuels. Three types of reflectors such as water, water/stainless steel 304 mixture and stainless steel 304 are selected to investigate the effect on core reactivity. Additionally, the thickness of stainless steel and double layer reflector type are also investigated. CASMO-4/SIMULATE-3 code system is used for this analysis. The results of analysis show that single layer stainless steel reflector is the most efficient reflector.

Study of various Brayton cycle designs for small modular sodium-cooled fast reactor

International Nuclear Information System (INIS)

Ahn, Yoonhan; Lee, Jeong Ik

2014-01-01

Highlights: • Application of closed Brayton cycle for small and medium sized SFRs is reviewed. • S-CO 2 , helium and nitrogen cycle designs for small modular SFR applications are analyzed and compared in terms of cycle efficiency, component performance and physical size. • Several new layouts for each Brayton cycle are suggested to simplify the turbomachinery designs. • S-CO 2 cycle design shows the best efficiency and compact size compared to other Brayton cycles. - Abstract: Many previous sodium cooled fast reactors (SFRs) adopted steam Rankine cycle as the power conversion system. However, the concern of sodium water reaction has been one of the major design issues of a SFR system. As an alternative to the steam Rankine cycle, several closed Brayton cycles including supercritical CO 2 cycle, helium cycle and nitrogen cycle have been suggested recently. In this paper, these alternative gas Brayton cycles will be compared to each other in terms of cycle performance and physical size for small modular SFR application. Several new layouts are suggested for each fluid while considering the turbomachinery design and the total system volume

Proposal of 'modular heliotron'

International Nuclear Information System (INIS)

Yamazaki, Kozo.

1993-11-01

A new modular helical configuration named 'Modular Heliotron' with clean and efficient helical magnetic divertor is proposed as an extension of the present conventional design of the continuous helical coil system. The sectored helical coils on one plane of the torus and the sectored returning vertical field coils on the other plane are combined. This coil system produces magnetic surfaces nearly equivalent to those of the l=2 helical system with one-pair poloidal coils, and overcomes the defects of construction and maintenance difficulties of the continuous coil systems. This concept satisfies the compatibility between the coil modularity and the sufficient divertor-space utilization, different from previous modular coil designs. The allowable length of the gap between each modular coil is clarified to keep good magnetic surfaces. Typical examples of the reactor coil configuration are described as an extension of the LHD (Large Helical Device) configuration. (author)

Proposal of 'Modular Heliotron'

International Nuclear Information System (INIS)

Yamazaki, Kozo

1994-01-01

A new modular helical system named 'Modular Heliotron' with clean and efficient helical magnetic divertor is proposed as an extension of the present conventional design of the continuous helical coil system. The sectored helical coils on one plane of the torus and the sectored returning vertical field coils on the other plane are combined. This coil system produces magnetic surfaces nearly equivalent to those of the l=2 helical system with one-pair poloidal coils, and overcomes the defects of construction and maintenance difficulties of the continuous coil systems. This concept satisfies the compatibility between the coil modularity and the sufficient divertor-space utilization, different from previous modular coil designs. The allowable length of the gap between each modular coil is clarified to keep good magnetic surfaces. Typical examples of the reactor coil configuration are described as an extension of the LHD (Large Helical Device) configuration. (author)

Considerations on the design of a helium circulator for a high temperature modular reactor system

International Nuclear Information System (INIS)

Dumm, K.; Donaldson, J.

1988-01-01

A modular helium cooled, high temperature reactor system with a thermal output of 200 MW per reactor has been developed by the KWU group for cogeneration of electricity and process steam. The flow of the reactor coolant - Helium at 60 bars and 250/700 deg. C is maintained by one circulator per reactor. The circulator is driven by a variable speed Siemens asynchronous motor and is submerged in the helium primary system. For operational reasons high reliability and availability of the circulator is required. The operational requirements for the circulator design are presented in this paper. The actual design has been carried out in close cooperation with the designer and manufacturer of all submerged circulators operating in AGR plants in Great Britain, James Howden Co. Renfrew, Scotland. Design solutions received so far and mainly based on sufficiently proven components - such as oil bath lubricated bearing systems - will be described. Special attention will be paid on the necessary test work; especially for the prototype to confirm the lay out. (author). 9 figs

Safety and licensing of MHTGR [Modular High Temperature Gas Cooled Reactor

International Nuclear Information System (INIS)

Silady, F.A.; Millunzi, A.C.; Kelley, A.P. Jr.; Cunliffe, J.

1987-07-01

The Modular High Temperature Gas Cooled Reactor (MHTGR) design meets stringent top-level regulatory and user safety requirements that require that the normal and off-normal operation of the plant not disturb the public's day-to-day activities. Quantitative, top-level regulatory criteria have been specified from US NRC and EPA sources to guide the design. The user/utility group has further specified that these criteria be met at the plant boundary. The focus of the safety approach has then been centered on retaining the radionuclide inventory within the fuel by removing core heat, controlling chemical attack, and by controlling heat generation. The MHTGR is shown to passively meet the stringent requirements with margin. No operator action is required and the plant is insensitive to operator error

Effects of Levels of Automation for Advanced Small Modular Reactors: Impacts on Performance, Workload, and Situation Awareness

Energy Technology Data Exchange (ETDEWEB)

Johanna Oxstrand; Katya Le Blanc

2014-07-01

The Human-Automation Collaboration (HAC) research effort is a part of the Department of Energy (DOE) sponsored Advanced Small Modular Reactor (AdvSMR) program conducted at Idaho National Laboratory (INL). The DOE AdvSMR program focuses on plant design and management, reduction of capital costs as well as plant operations and maintenance costs (O&M), and factory production costs benefits.

Cooling device for thermonuclear reactor and modular packing block for the wall realization of a such device

International Nuclear Information System (INIS)

Archer, J.; Stalport, G.; Besson, D.; Faron, R.; Coulon, M.

1988-01-01

The cooling device for a thermonuclear reactor wall is made by modular thermally conductive heat-resistant blocks (graphite by example), a prismatic head on one face of each block, the opposite face bearing against cooling tubes, a base to each block with an aperture and rods passing through the apertures reversibly fixing each row of blocks to a support [fr

Design Study of Modular Nuclear Power Plant with Small Long Life Gas Cooled Fast Reactors Utilizing MOX Fuel

Science.gov (United States)

Ilham, Muhammad; Su'ud, Zaki

2017-01-01

Growing energy needed due to increasing of the world’s population encourages development of technology and science of nuclear power plant in its safety and security. In this research, it will be explained about design study of modular fast reactor with helium gas cooling (GCFR) small long life reactor, which can be operated over 20 years. It had been conducted about neutronic design GCFR with Mixed Oxide (UO2-PuO2) fuel in range of 100-200 MWth NPPs of power and 50-60% of fuel fraction variation with cylindrical pin cell and cylindrical balance of reactor core geometry. Calculation method used SRAC-CITATION code. The obtained results are the effective multiplication factor and density value of core reactor power (with geometry optimalization) to obtain optimum design core reactor power, whereas the obtained of optimum core reactor power is 200 MWth with 55% of fuel fraction and 9-13% of percentages.

Design of a Modular Monolithic Implicit Solver for Multi-Physics Applications

Science.gov (United States)

Carton De Wiart, Corentin; Diosady, Laslo T.; Garai, Anirban; Burgess, Nicholas; Blonigan, Patrick; Ekelschot, Dirk; Murman, Scott M.

2018-01-01

The design of a modular multi-physics high-order space-time finite-element framework is presented together with its extension to allow monolithic coupling of different physics. One of the main objectives of the framework is to perform efficient high- fidelity simulations of capsule/parachute systems. This problem requires simulating multiple physics including, but not limited to, the compressible Navier-Stokes equations, the dynamics of a moving body with mesh deformations and adaptation, the linear shell equations, non-re effective boundary conditions and wall modeling. The solver is based on high-order space-time - finite element methods. Continuous, discontinuous and C1-discontinuous Galerkin methods are implemented, allowing one to discretize various physical models. Tangent and adjoint sensitivity analysis are also targeted in order to conduct gradient-based optimization, error estimation, mesh adaptation, and flow control, adding another layer of complexity to the framework. The decisions made to tackle these challenges are presented. The discussion focuses first on the "single-physics" solver and later on its extension to the monolithic coupling of different physics. The implementation of different physics modules, relevant to the capsule/parachute system, are also presented. Finally, examples of coupled computations are presented, paving the way to the simulation of the full capsule/parachute system.

Population Sensitivity Evaluation of Two Proposed Hampton Roads Area Sites for a Possible Small Modular Reactor

Energy Technology Data Exchange (ETDEWEB)

Belles, R. J. [ORNL; Omitaomu, O. A. [ORNL

2014-08-01

The overall objective of this research project is to use the OR-SAGE tool to support the US Department of Energy (DOE) Office of Nuclear Energy (NE) in evaluating future electrical generation deployment options for small modular reactors (SMRs) in areas with significant energy demand from the federal sector. Deployment of SMRs in zones with high federal energy use can provide a means of meeting federal clean energy goals.

A feasibility study of a linear laser heated solenoid fusion reactor. Final report

International Nuclear Information System (INIS)

Steinhauer, L.C.

1976-02-01

This report examines the feasibility of a laser heated solenoid as a fusion or fusion-fission reactor system. The objective of this study, was an assessment of the laser heated solenoid reactor concept in terms of its plasma physics, engineering design, and commercial feasibility. Within the study many pertinent reactor aspects were treated including: physics of the laser-plasma interaction; thermonuclear behavior of a slender plasma column; end-losses under reactor conditions; design of a modular first wall, a hybrid (both superconducting and normal) magnet, a large CO 2 laser system; reactor blanket; electrical storage elements; neutronics; radiation damage, and tritium processing. Self-consistent reactor configurations were developed for both pure fusion and fusion-fission designs, with the latter designed both to produce power and/or fissile fuels for conventional fission reactors. Appendix A is a bibliography with commentary of theoretical and experimental studies that have been directed at the laser heated solenoid

Utilization of thorium in a Gas Turbine – Modular Helium Reactor

International Nuclear Information System (INIS)

Şahin, Hacı Mehmet; Erol, Özgür; Acır, Adem

2012-01-01

Highlights: ► Performance parameters for the original fuel in GT-MHR depending on time were found. ► A proper plutonium–thorium mixture ratio was found using the original fuel results. ► Performance comparison of plutonium mixture and original fuel was made. ► Comparison showed that weapons grade plutonium mixture can be used in the reactor. - Abstract: Gas Turbine-Modular Helium Reactor (GT-MHR) is one of the new types of the reactors with high efficiency and increased safety features. The usage of different kinds of fissile material in this reactor can increase the life of it. Weapons-grade plutonium (WGrPu), which can be acquired from the old dismantled nuclear weapons, can be an option in a GT-MHR. In order to increase the sustainability of the WGrPu resources this fuel can be mixed with thorium, which is a fertile material that can be found in the nature and has resources three times more than uranium. In this study, possibility of utilization of the weapons-grade plutonium–thorium mixture was investigated and an optimum mixture ratio was determined. The behavior of this mixture and the original fuel was studied by using MCNP5 1.4, Monteburns 2.0 and Origen 2.2 tools. Calculations showed that, a GT-MHR type reactor, which is using the original TRISO fuel particle mixture of 20% enriched uranium + natural uranium (original fuel) has an effective multiplication factor (k eff ) of 1.270. Corresponding to this k eff value the weapons grade plutonium/thorium oxide mixture was found 19%/81%. By using Monteburns Code, the operation time, which describes the time passed until the reactor reaches a k eff value of 1.02, was found as 515 days for the original fuel and 1175 days for the weapons grade plutonium mixture. Furthermore, the burn-up values for the original fuel and WGrPu fuels were found as 47.69 and 119.27 GWd/MTU, respectively.

Perspectives on understanding and verifying the safety terrain of modular high temperature gas-cooled reactors

Energy Technology Data Exchange (ETDEWEB)

Carlson, Donald E., E-mail: donald@carlsonperin.net [11221 Empire Lane, Rockville, MD 20852 (United States); Ball, Sydney J., E-mail: beckysyd@comcast.net [100 Greywood Place, Oak Ridge, TN 37830 (United States)

2016-09-15

The passive safety characteristics of modular high temperature gas-cooled reactors (HTGRs) are conceptually well known and are largely supported by insights from past and ongoing research. This paper offers perspectives on selected issues in areas where further analysis and testing achievable within existing research and demonstration programs could help address residual uncertainties and better support the analysis of safety performance and the regulatory assessment of defense in depth. Areas considered include the evaluation of normal and anomalous core operating conditions and the analysis of accidents involving loss of forced cooling, coolant depressurization, air ingress, moisture ingress, and reactivity events. In addition to discussing associated uncertainties and potential measures to address them, this paper also proposes supplemental “safety terrain” studies that would use realistic assessments of postulated extreme event sequences to establish a more comprehensive understanding of the inherent behaviors and ultimate safety capabilities of modular HTGRs.

Perspectives on Understanding and Verifying the Safety Terrain of Modular High Temperature Gas-Cooled Reactors

International Nuclear Information System (INIS)

Carlson, Donald E.

2014-01-01

The inherent safety characteristics of modular high temperature gas-cooled reactors (HTGRs) are conceptually well known and are largely supported by insights from past and ongoing research. This paper offers perspectives on selected issues in areas where further analysis and testing achievable within existing research and demonstration programs could help address residual uncertainties and better support the analysis of safety performance and the regulatory assessment of defense in depth. Areas considered include the evaluation of normal and anomalous core operating conditions and the analysis of accidents involving coolant depressurization, air ingress, moisture ingress, and reactivity insertion. In addition to discussing associated uncertainties and potential measures to address them, the paper also proposes supplemental “safety terrain” studies that would use realistic assessments of postulated extreme event sequences to establish a more comprehensive understanding of the inherent behaviors and ultimate safety capabilities of modular HTGRs. (author)

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.

Study of indicators aggregation techniques for the selection of a new nuclear reactor for Mexico; Estudio de tecnicas de agregacion de indicadores para la seleccion de un nuevo reactor nuclear para Mexico

Energy Technology Data Exchange (ETDEWEB)

Barragan M, A.M.; Martin del Campo M, C. [UNAM, Facultad de Ingenieria, 04510 Mexico D.F. (Mexico)]. e-mail: ale_bar_m@yahoo.com.mx

2007-07-01

A study on several aggregation techniques that can be used as multi criteria analysis methods, like important part of the methodology developed for the selection of a nuclear reactor for Mexico is described. In an arbitrary way three reactors were selected to be compared, these they are the AP1000 (Advance Passive from 1000 MWe), the PBMR (Pebble Bed Modular Reactor) and the GT-MHR (Gas Turbine Modular Helium). The evaluation approaches were classified in three categories: Economic, Socio-political and of safety and environment. In each category they were defined the more important evaluation indicators and then it was built a matrix with those values of each reactor. The four studied aggregation methods are described: Normalization, Linear deliberation, Fuzzy Logic and AHP (Analytic Hierarchy Process). The well-known aggregation mechanisms are those that are obtained of the lineal deliberation and of the normalization, which have demonstrated to give good results before the simplicity of their use. The fuzzy logic has the advantage that it allows to manage qualitative and quantitative information simultaneously without the aggregation problems that are presented since in a conventional system the semantic pattern on that is based, it is provided by the theory of the diffuse groups that has demonstrated in other areas of the knowledge a better approach to the reality, when admitting that the nature has shades and that the decisions take in function of a wide range of possibilities and of approaches in contradictory occasions or in conflict, all equally worth. The Analytic Hierarchical Process (AHP) that consists in formalizing the intuitive understanding of a multi criteria complex problem, by means of the construction of a hierarchical model that allows the decision agent to structure the problem in visual form, giving him the form of a hierarchy of attributes (global objective of the problem, approaches and alternative). Finally, using the matrix of initiators

Efficacy and safety of a modular multi-modal exercise program in prostate cancer patients with bone metastases: a randomized controlled trial

International Nuclear Information System (INIS)

Galvão, Daniel A; Groom, Geoff; Newton, Robert U; Taaffe, Dennis R; Cormie, Prue; Spry, Nigel; Chambers, Suzanne K; Peddle-McIntyre, Carolyn; Baker, Michael; Denham, James; Joseph, David

2011-01-01

The presence of bone metastases has excluded participation of prostate cancer patients in exercise intervention studies to date and is also a relative contraindication to supervised exercise in the community setting because of concerns of fragility fracture. However, this group of patients often have developed significant muscle atrophy and functional impairments from prior and continuing androgen deprivation that is exacerbated by subsequent and more intensive interventions such as chemotherapy. The aim of this study is to determine the efficacy and safety of a modular multi-modal exercise program in prostate cancer patients with bone metastases. Multi-site randomized controlled trial in Western Australia and New South Wales to examine the efficacy and safety of a modular multi-modal physical exercise program in 90 prostate cancer survivors with bone metastases. Participants will be randomized to (1) modular multi-modal exercise intervention group or (2) usual medical care group. The modular multi-modal exercise group will receive a 3-month supervised exercise program based on bone lesion location/extent. Measurements for primary and secondary endpoints will take place at baseline, 3 months (end of the intervention) and 6 months follow-up. Delaying or preventing skeletal complication and improving physical function for men with bone metastases would provide clinically meaningful benefits to patients. However, exercise programs must be designed and executed with careful consideration of the skeletal complications associated with bone metastatic disease and cumulative toxicities from androgen deprivation such as osteoporosis and increased risk of fractures. The results from this study will form the basis for the development of a specific exercise prescription in this patient group in order to alleviate disease burden, counteract the adverse treatment related side-effects and enhance quality of life. ACTRN: http://www.anzctr.org.au/ACTRN12611001158954.aspx

AREVA Modular Steam Cycle – High Temperature Gas-Cooled Reactor Development Progress

International Nuclear Information System (INIS)

Lommers, L.; Shahrokhi, F.; Southworth, F.; Mayer, J. III

2014-01-01

The AREVA Steam Cycle – High Temperature Gas-Cooled Reactor (SCHTGR) is a modular graphite-moderated gas-cooled reactor currently being developed to support a wide variety of applications including industrial process heat, high efficiency electricity generation, and cogeneration. It produces high temperature superheated steam which makes it a good match for many markets currently dependent on fossil fuels for process heat. Moreover, the intrinsic safety characteristics of the SC-HTGR make it uniquely qualified for collocation with large industrial process heat users which is necessary for serving these markets. The NGNP Industry Alliance has selected the AREVA SC-HTGR as the basis for future development work to support commercial HTGR deployment. This paper provides a concise description of the SC-HTGR concept, followed by a summary of recent development activities. Since this concept was introduced, ongoing design activities have focused primarily on confirming key system capabilities and the suitability for potential future markets. These evaluations continue to confirm the suitability of the SC-HTGR for a variety of potential applications that are currently dependent on fossil fuels. (author)

Modular head assembly and method of retrofitting existing nuclear reactor facilities

International Nuclear Information System (INIS)

Malandra, L.J.; Ledue, R.J.; Hankinson, M.F.; Kowalski, E.F.

1987-01-01

A method is described of retrofitting existing nuclear reactor facilities so as to form a modular closure head assembly for a nuclear reactor pressure vessel, where the existing nuclear reactor facilities comprise control rod drive mechanism cooling systems which include vertically extending elbow air ducts inter-connecting vertically spaced upper and lower air manifolds. The elbow air ducts extend radially beyond the peripheral envelope of the closure head, comprising the steps of: removing the upper air manifold; removing the vertically extending elbow air ducts; capping the air ports of the lower air manifold which ports were previously fluidically connecting the lower air manifold to the vertically extending elbow air ducts; disposing vertically upwardly extending air exhaust ducts above the lower air manifold in such an manner that the air exhaust ducts are disposed within the peripheral envelope of the closure head; fluidically connecting exhaust fans to the upper regions of the air exhaust ducts; fluidically connecting the lower regions of the air exhaust ducts the lower air manifold; permanently securing lift rods to the closure head at positions disposed radially outwardly of the lower air manifold; attaching a seismic support platform to the lift rods; proving fluidic passage of the vertically extending air exhaust ducts through the seismic support platform; attaching a missile shield plate to the lift rods; and proving fluidic passage of the vertically extending air exhaust ducts through the missile shield plate

The Modular High-Temperature Gas-Cooled Reactor (MHTGR) in the US

International Nuclear Information System (INIS)

Neylan, A.J.; Graf, D.V.; Millunzi, A.C.

1987-08-01

The MHTGR is an advanced nuclear reactor concept being developed in the USA under a cooperative program involving the US Government, the nuclear industry, and the utilities. As its objective, this program is developing a safe, reliable, and economic nuclear power option for the USA and the other nations of the world to consider in meeting their individual nationalistic electrical generation or process heat needs by the turn of the century. The design is based on a concept of modularization that can meet the various power needs by combining any number of 350 MW(t) reactor modules in parallel with a selected number of turbine plants in a variety of arrangements. Basic HTGR features of ceramic fuel, helium coolant, and graphite are sized and configured to provide a low power density core with passive safety features such that no operator action or external source of power is needed for the plant to meet 10CFR100 or Protective Action Guidelines limits at the 425 m site boundary. This precludes the necessity to plan for the evacuation or sheltering of the public during any licensing basis event. The safe behavior of the reactor plant is not dependent upon operator action and it is insensitive to operator error. The Conceptual Design is presently being vigorously reviewed by the US Nuclear Regulatory Commission (NRC). A safety evaluation report and a licensability statement are scheduled for issuance by the NRC in January 1988. 2 refs., 5 figs., 1 tab

Modelling an industrial anaerobic granular reactor using a multi-scale approach

DEFF Research Database (Denmark)

Feldman, Hannah; Flores Alsina, Xavier; Ramin, Pedram

2017-01-01

The objective of this paper is to show the results of an industrial project dealing with modelling of anaerobic digesters. A multi-scale mathematical approach is developed to describe reactor hydrodynamics, granule growth/distribution and microbial competition/inhibition for substrate/space within...... the biofilm. The main biochemical and physico-chemical processes in the model are based on the Anaerobic Digestion Model No 1 (ADM1) extended with the fate of phosphorus (P), sulfur (S) and ethanol (Et-OH). Wastewater dynamic conditions are reproduced and data frequency increased using the Benchmark...... simulations show the effects on the overall process performance when operational (pH) and loading (S:COD) conditions are modified. Lastly, the effect of intra-granular precipitation on the overall organic/inorganic distribution is assessed at: 1) different times; and, 2) reactor heights. Finally...

Development of an Analytic Nodal Diffusion Solver in Multi-groups for 3D Reactor Cores with Rectangular or Hexagonal Assemblies

Energy Technology Data Exchange (ETDEWEB)

Lozano, Juan Andres; Aragones, Jose Maria; Garcia-Herranz, Nuria [Universidad Politecnica de Madrid, 28006 Jose Gutierrez Abascal 2, Madrid (Spain)

2008-07-01

More accurate modelling of physical phenomena involved in present and future nuclear reactors requires a multi-scale and multi-physics approach. This challenge can be accomplished by the coupling of best-estimate core-physics, thermal-hydraulics and multi-physics solvers. In order to make viable that coupling, the current trends in reactor simulations are along the development of a new generation of tools based on user-friendly, modular, easily linkable, faster and more accurate codes to be integrated in common platforms. These premises are in the origin of the NURESIM Integrated Project within the 6. European Framework Program, which is envisaged to provide the initial step towards a Common European Standard Software Platform for nuclear reactors simulations. In the frame of this project and to reach the above-mentioned goals, a 3-D multigroup nodal solver for neutron diffusion calculations called ANDES (Analytic Nodal Diffusion Equation Solver) has been developed and tested in-depth in this Thesis. ANDES solves the steady-state and time-dependent neutron diffusion equation in three-dimensions and any number of energy groups, utilizing the Analytic Coarse-Mesh Finite-Difference (ACMFD) scheme to yield the nodal coupling equations. It can be applied to both Cartesian and triangular-Z geometries, so that simulations of LWR as well as VVER, HTR and fast reactors can be performed. The solver has been implemented in a fully encapsulated way, enabling it as a module to be readily integrated in other codes and platforms. In fact, it can be used either as a stand-alone nodal code or as a solver to accelerate the convergence of whole core pin-by-pin code systems. Verification of performance has shown that ANDES is a code with high order definition for whole core realistic nodal simulations. In this paper, the methodology developed and involved in ANDES is presented. (authors)

Final Stage Development of Reactor Console Simulator

International Nuclear Information System (INIS)

Mohamad Idris Taib; Ridzuan Abdul Mutalib; Zareen Khan Abdul Jalil Khan; Mohd Khairulezwan Abdul Manan; Mohd Sabri Minhat; Nurfarhana Ayuni Joha

2013-01-01

The Reactor Console Simulator PUSPATI TRIGA Reactor was developed since end of 2011 and now in the final stage of development. It is will be an interactive tool for operator training and teaching of PUSPATI TRIGA Reactor. Behavior and characteristic for reactor console and reactor itself can be evaluated and understand. This Simulator will be used as complement for actual present reactor console. Implementation of human system interface (HSI) is using computer screens, keyboard and mouse. Multiple screens are used to match the physical of present reactor console. LabVIEW software are using for user interface and mathematical calculation. Polynomial equation based on control rods calibration data as well as operation parameters record was used to calculate and estimated reactor console parameters. The capabilities in user interface, reactor physics and thermal-hydraulics can be expanded and explored to simulation as well as modeling for New Reactor Console, Research Reactor and Nuclear Power Plant. (author)

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)

Activity Based Costing as a method for assessing the economics of modularization - a case study and beyond

DEFF Research Database (Denmark)

Thyssen, Peter; Israelsen, Poul; Jørgensen, Brian

The paper accounts for an Activity Based Costing (ABC) analysis supporting decision-making concerning product modularity. The ABC analysis carried out is communicated to decision-makers by telling how much higher the variable cost of the multi-purpose module can be compared to the average variable...... cost for the product-unique modules that it substitutes to break even in total cost. The analysis provides the platform for stating three general rules of cost efficiency of modularization, which in combination identify the highest profit potential of product modularisation. Finally the analysis points...... to problems of using ABC in costing modularity, i.e. handling of R&D costs and identification of product-profitability upon an enhanced modularization....

Development of the Mathematics of Learning Curve Models for Evaluating Small Modular Reactor Economics

Energy Technology Data Exchange (ETDEWEB)

Harrison, T. J. [ORNL

2014-02-01

The cost of nuclear power is a straightforward yet complicated topic. It is straightforward in that the cost of nuclear power is a function of the cost to build the nuclear power plant, the cost to operate and maintain it, and the cost to provide fuel for it. It is complicated in that some of those costs are not necessarily known, introducing uncertainty into the analysis. For large light water reactor (LWR)-based nuclear power plants, the uncertainty is mainly contained within the cost of construction. The typical costs of operations and maintenance (O&M), as well as fuel, are well known based on the current fleet of LWRs. However, the last currently operating reactor to come online was Watts Bar 1 in May 1996; thus, the expected construction costs for gigawatt (GW)-class reactors in the United States are based on information nearly two decades old. Extrapolating construction, O&M, and fuel costs from GW-class LWRs to LWR-based small modular reactors (SMRs) introduces even more complication. The per-installed-kilowatt construction costs for SMRs are likely to be higher than those for the GW-class reactors based on the property of the economy of scale. Generally speaking, the economy of scale is the tendency for overall costs to increase slower than the overall production capacity. For power plants, this means that doubling the power production capacity would be expected to cost less than twice as much. Applying this property in the opposite direction, halving the power production capacity would be expected to cost more than half as much. This can potentially make the SMRs less competitive in the electricity market against the GW-class reactors, as well as against other power sources such as natural gas and subsidized renewables. One factor that can potentially aid the SMRs in achieving economic competitiveness is an economy of numbers, as opposed to the economy of scale, associated with learning curves. The basic concept of the learning curve is that the more a

NCSU reactor sharing program. Final technical report

International Nuclear Information System (INIS)

Perez, P.B.

1997-01-01

The Nuclear Reactor Program at North Carolina State University provides the PULSTAR Research Reactor and associated facilities to eligible institutions with support, in part, from the Department of Energy Reactor Sharing Program. Participation in the NCSU Reactor Sharing Program continues to increase steadily with visitors ranging from advance high school physics and chemistry students to Ph.D. level research from neighboring universities. This report is the Final Technical Report for the DOE award reference number DE-FG05-95NE38136 which covers the period September 30, 1995 through September 30, 1996

350 MW(t) MHTGR preassembly and modularization

International Nuclear Information System (INIS)

Venkatesh, M.C.; Jones, G.; Dilling, D.A.; Parker, W.J.

1991-05-01

The Modular High Temperature Gas Cooled Reactor (MHTGR) provides a safe and economical nuclear power option for the world's electrical generation needs by the turn of the century. The proposed MHTGR plant is composed of four 350 MW(t) prismatic core reactor modules, coupled to a 2(2 x 1) turbine generator producing a net plant electrical output of 538 MW(e). Each of the four reactor module is located in a below-ground level concrete silo, and consists of a reactor vessel and a steam generator vessel interconnected by a cross duct vessel. The modules, along with the service buildings, are contained within a Nuclear Island (NI). The turbine generators and power generation facilities are in the non-nuclear Energy Conversion Area (ECA). The MHTGR design reduces cost and improves schedule by maximizing shop fabrication, minimizing field fit up of the Reactor Internals components and modularizing the NI ampersand ECA facilities. 3 refs., 6 figs., 2 tabs

Three-Dimensional Analysis of the Hot-Spot Fuel Temperature in Pebble Bed and Prismatic Modular Reactors

International Nuclear Information System (INIS)

In, W. K.; Lee, S. W.; Lim, H. S.; Lee, W. J.

2006-01-01

High temperature gas-cooled reactors(HTGR) have been reviewed as potential sources for future energy needs, particularly for a hydrogen production. Among the HTGRs, the pebble bed reactor(PBR) and a prismatic modular reactor(PMR) are considered as the nuclear heat source in Korea's nuclear hydrogen development and demonstration project. PBR uses coated fuel particles embedded in spherical graphite fuel pebbles. The fuel pebbles flow down through the core during an operation. PMR uses graphite fuel blocks which contain cylindrical fuel compacts consisting of the fuel particles. The fuel blocks also contain coolant passages and locations for absorber and control material. The maximum fuel temperature in the core hot spot is one of the important design parameters for both a PBR and a PMR. The objective of this study is to predict the hot-spot fuel temperature distributions in a PBR and a PMR at a steady state. The computational fluid dynamics(CFD) code, CFX-10 is used to perform the three-dimensional analysis. The latest design data was used here based on the reference reactor designs, PBMR400 and GTMHR60

A Conceptual Study of a Supercritical CO2-Cooled Micro Modular Reactor

Directory of Open Access Journals (Sweden)

Hwanyeal Yu

2015-12-01

Full Text Available A neutronics conceptual study of a supercritical CO2-cooled micro modular reactor (MMR has been performed in this work. The suggested MMR is an extremely compact and truck-transportable nuclear reactor. The thermal power of the MMR is 36.2 MWth and it is designed to have a 20-year lifetime without refueling. A salient feature of the MMR is that all the components including the generator are integrated in a small reactor vessel. For a minimal volume and long lifetime of the MMR core, a fast neutron spectrum is utilized in this work. To enhance neutron economy and maximize the fuel volume fraction in the core, a high-density uranium mono-nitride U15N fuel is used in the fast-spectrum MMR. Unlike the conventional supercritical CO2-cooled fast reactors, a replaceable fixed absorber (RFA is introduced in a unique way to minimize the excess reactivity and the power peaking factor of the core. For a compact core design, the drum-type control absorber is adopted as the primary reactivity control mechanism. In this study, the neutronics analyses and depletions have been performed by using the continuous energy Monte Carlo Serpent code with the evaluated nuclear data file ENDF/B-VII.1 Library. The MMR core is characterized in view of several important safety parameters such as control system worth, fuel temperature coefficient (FTC and coolant void reactivity (CVR, etc. In addition, a preliminary thermal-hydraulic analysis has also been performed for the hottest channel of the Korea Advanced Institute of Science and Technology (KAIST MMR.

Study of indicators aggregation techniques for the selection of a new nuclear reactor for Mexico

International Nuclear Information System (INIS)

Barragan M, A.M.; Martin del Campo M, C.

2007-01-01

A study on several aggregation techniques that can be used as multi criteria analysis methods, like important part of the methodology developed for the selection of a nuclear reactor for Mexico is described. In an arbitrary way three reactors were selected to be compared, these they are the AP1000 (Advance Passive from 1000 MWe), the PBMR (Pebble Bed Modular Reactor) and the GT-MHR (Gas Turbine Modular Helium). The evaluation approaches were classified in three categories: Economic, Socio-political and of safety and environment. In each category they were defined the more important evaluation indicators and then it was built a matrix with those values of each reactor. The four studied aggregation methods are described: Normalization, Linear deliberation, Fuzzy Logic and AHP (Analytic Hierarchy Process). The well-known aggregation mechanisms are those that are obtained of the lineal deliberation and of the normalization, which have demonstrated to give good results before the simplicity of their use. The fuzzy logic has the advantage that it allows to manage qualitative and quantitative information simultaneously without the aggregation problems that are presented since in a conventional system the semantic pattern on that is based, it is provided by the theory of the diffuse groups that has demonstrated in other areas of the knowledge a better approach to the reality, when admitting that the nature has shades and that the decisions take in function of a wide range of possibilities and of approaches in contradictory occasions or in conflict, all equally worth. The Analytic Hierarchical Process (AHP) that consists in formalizing the intuitive understanding of a multi criteria complex problem, by means of the construction of a hierarchical model that allows the decision agent to structure the problem in visual form, giving him the form of a hierarchy of attributes (global objective of the problem, approaches and alternative). Finally, using the matrix of initiators

Advanced Concepts for Pressure-Channel Reactors: Modularity, Performance and Safety

Science.gov (United States)

Duffey, Romney B.; Pioro, Igor L.; Kuran, Sermet

Based on an analysis of the development of advanced concepts for pressure-tube reactor technology, we adapt and adopt the pressure-tube reactor advantage of modularity, so that the subdivided core has the potential for optimization of the core, safety, fuel cycle and thermal performance independently, while retaining passive safety features. In addition, by adopting supercritical water-cooling, the logical developments from existing supercritical turbine technology and “steam” systems can be utilized. Supercritical and ultra-supercritical boilers and turbines have been operating for some time in coal-fired power plants. Using coolant outlet temperatures of about 625°C achieves operating plant thermal efficiencies in the order of 45-48%, using a direct turbine cycle. In addition, by using reheat channels, the plant has the potential to produce low-cost process heat, in amounts that are customer and market dependent. The use of reheat systems further increases the overall thermal efficiency to 55% and beyond. With the flexibility of a range of plant sizes suitable for both small (400 MWe) and large (1400 MWe) electric grids, and the ability for co-generation of electric power, process heat, and hydrogen, the concept is competitive. The choice of core power, reheat channel number and exit temperature are all set by customer and materials requirements. The pressure channel is a key technology that is needed to make use of supercritical water (SCW) in CANDU®1 reactors feasible. By optimizing the fuel bundle and fuel channel, convection and conduction assure heat removal using passive-moderator cooling. Potential for severe core damage can be almost eliminated, even without the necessity of activating the emergency-cooling systems. The small size of containment structure lends itself to a small footprint, impacts economics and building techniques. Design features related to Canadian concepts are discussed in this paper. The main conclusion is that development of

Current status and future development of modular high temperature gas cooled reactor technology

International Nuclear Information System (INIS)

2001-02-01

This report includes an examination of the international activities with regard to the development of the modular HTGR coupled to a gas turbine. The most significant of these gas turbine programmes include the pebble bed modular reactor (PBMR) being designed by ESKOM of South Africa and British Nuclear Fuels plc. (BNFL) of the United Kingdom, and the gas turbine-modular helium reactor (GT-MHR) by a consortium of General Atomics of the United States of America, MINATOM of the Russian Federation, Framatome of France and Fuji Electric of Japan. Details of the design, economics and plans for these plants are provided in Chapters 3 and 4, respectively. Test reactors to evaluate the safety and general performance of the HTGR and to support research and development activities including electricity generation via the gas turbine and validation of high temperature process heat applications are being commissioned in Japan and China. Construction of the high temperature engineering test reactor (HTTR) by the Japan Atomic Energy Research Institute (JAERI) at its Oarai Research Establishment has been completed with the plant currently in the low power physics testing phase of commissioning. Construction of the high temperature reactor (HTR-10) by the Institute of Nuclear Energy Technology (INET) in Beijing, China, is nearly complete with initial criticality expected in 2000. Chapter 5 provides a discussion of purpose, status and testing programmes for these two plants. In addition to the activities related to the above mentioned plants, Member States of the IWGGCR continue to support research associated with HTGR safety and performance as well as development of alternative designs for commercial applications. These activities are being addressed by national energy institutes and, in some projects, private industry, within China, France, Germany, Indonesia, Japan, the Netherlands, the Russian Federation, South Africa, United Kingdom and the USA. Chapter 6 includes details

ADVANTAGES, DISADVANTAGES, AND LESSONS LEARNED FROM MULTI-REACTOR DECOMMISSIONING PROJECTS

International Nuclear Information System (INIS)

Morton, M.R.; Nielson, R.R.; Trevino, R.A.

2003-01-01

This paper discusses the Reactor Interim Safe Storage (ISS) Project within the decommissioning projects at the Hanford Site and reviews the lessons learned from performing four large reactor decommissioning projects sequentially. The advantages and disadvantages of this multi-reactor decommissioning project are highlighted

Modular battery design for reliable, flexible and multi-technology energy storage systems

International Nuclear Information System (INIS)

Rothgang, Susanne; Baumhöfer, Thorsten; Hoek, Hauke van; Lange, Tobias; De Doncker, Rik W.; Sauer, Dirk Uwe

2015-01-01

Highlights: • Collection of existing battery topologies in electric vehicles. • Analysis of load profiles and the power consumption for electric vehicles. • Composition of battery packs and their passive components. • Modular, hybrid battery architecture with a dc-link. - Abstract: With large scale battery systems being more and more used in demanding applications regarding lifetime, performance and safety, it is of great importance to utilize not only cells with a high cyclic and calendric lifetime but also to optimize the whole system architecture. The aim of this work is therefore, to highlight the benefits of a modular system architecture allowing the use of hybrid battery systems combining high power and high energy cells in a multi-technology system. To achieve an optimized performance, efficiency and lifetime for an electric vehicle the complete drive train topology has to be taken into account instead of optimizing one of the components individually. Consequently, the topic will be analyzed from the system’s point of view, addressing in particular the modularization of the battery as well as the power electronics needed to do so. It will be shown that a highly flexible battery system can be realized by dc-to-dc converters between a modular, hybrid battery system and the drive inverter. By the dc-to-dc converters the battery output voltages and the inverter input voltages are decoupled. Hence, the battery’s topology can be chosen unrestrictedly within a wide range and easily be interconnected to a common dc-link of a different voltage. The benefits of this flexibility will be analyzed in detail showing especially how the lifetime of the battery system can be improved and the impact on system weight

The burnup capabilities of the Deep Burn Modular Helium Reactor analyzed by the Monte Carlo Continuous Energy Code MCB

Energy Technology Data Exchange (ETDEWEB)

Talamo, Alberto E-mail: alby@neutron.kth.se; Gudowski, Waclaw E-mail: wacek@neutron.kth.se; Venneri, Francesco E-mail: venneri@lanl.gov

2004-01-01

We have investigated the waste actinide burnup capabilities of a Gas Turbine Modular Helium Reactor (GT-MHR, similar to the reactor being designed by General Atomics and Minatom for surplus weapons plutonium destruction) with the Monte Carlo Continuous Energy Burnup Code MCB, an extension of MCNP developed at the Royal Institute of Technology in Stockholm and University of Mining and Metallurgy in Krakow. The GT-MHR is a gas-cooled, graphite-moderated reactor, which can be powered with a wide variety of fuels, like thorium, uranium or plutonium. In the present work, the GT-MHR is fueled with the transuranic actinides contained in Light Water Reactors (LWRs) spent fuel for the purpose of destroying them as completely as possible with minimum reliance on multiple reprocessing steps. After uranium extraction from the LWR spent fuel (UREX), the remaining waste actinides, including plutonium are partitioned into two distinct types of fuel for use in the GT-MHR: Driver Fuel (DF) and Transmutation Fuel (TF). The DF supplies the neutrons to maintain the fission chain reaction, whereas the TF emphasizes neutron capture to induce a deep burn transmutation and provide reactivity control by a negative feedback. When used in this mode, the GT-MHR is called Deep Burn Modular Helium Reactor (DB-MHR). Both fuels are contained in a structure of triple isotropic coated layers, TRISO coating, which has been proven to retain fission products up to 1600 deg. C and is expected to remain intact for hundreds of thousands of years after irradiation. Other benefits of this reactor consist of: a well-developed technology, both for the graphite-moderated core and the TRISO structure, a high energy conversion efficiency (about 50%), well established passive safety mechanism and a competitive cost. The destruction of more than 94% of {sup 239}Pu and the other geologically problematic actinide species makes this reactor a valid proposal for the reduction of nuclear waste and the prevention of

The burnup capabilities of the Deep Burn Modular Helium Reactor analyzed by the Monte Carlo Continuous Energy Code MCB

International Nuclear Information System (INIS)

Talamo, Alberto; Gudowski, Waclaw; Venneri, Francesco

2004-01-01

We have investigated the waste actinide burnup capabilities of a Gas Turbine Modular Helium Reactor (GT-MHR, similar to the reactor being designed by General Atomics and Minatom for surplus weapons plutonium destruction) with the Monte Carlo Continuous Energy Burnup Code MCB, an extension of MCNP developed at the Royal Institute of Technology in Stockholm and University of Mining and Metallurgy in Krakow. The GT-MHR is a gas-cooled, graphite-moderated reactor, which can be powered with a wide variety of fuels, like thorium, uranium or plutonium. In the present work, the GT-MHR is fueled with the transuranic actinides contained in Light Water Reactors (LWRs) spent fuel for the purpose of destroying them as completely as possible with minimum reliance on multiple reprocessing steps. After uranium extraction from the LWR spent fuel (UREX), the remaining waste actinides, including plutonium are partitioned into two distinct types of fuel for use in the GT-MHR: Driver Fuel (DF) and Transmutation Fuel (TF). The DF supplies the neutrons to maintain the fission chain reaction, whereas the TF emphasizes neutron capture to induce a deep burn transmutation and provide reactivity control by a negative feedback. When used in this mode, the GT-MHR is called Deep Burn Modular Helium Reactor (DB-MHR). Both fuels are contained in a structure of triple isotropic coated layers, TRISO coating, which has been proven to retain fission products up to 1600 deg. C and is expected to remain intact for hundreds of thousands of years after irradiation. Other benefits of this reactor consist of: a well-developed technology, both for the graphite-moderated core and the TRISO structure, a high energy conversion efficiency (about 50%), well established passive safety mechanism and a competitive cost. The destruction of more than 94% of 239 Pu and the other geologically problematic actinide species makes this reactor a valid proposal for the reduction of nuclear waste and the prevention of

Modular Robotic Wearable

DEFF Research Database (Denmark)

Lund, Henrik Hautop; Pagliarini, Luigi

2009-01-01

In this concept paper we trace the contours and define a new approach to robotic systems, composed of interactive robotic modules which are somehow worn on the body. We label such a field as Modular Robotic Wearable (MRW). We describe how, by using modular robotics for creating wearable....... Finally, by focusing on the intersection of the combination modular robotic systems, wearability, and bodymind we attempt to explore the theoretical characteristics of such approach and exploit the possible playware application fields....

Burning of spent fuel of an accelerator-driven modular HTGR in sub-critical condition

International Nuclear Information System (INIS)

Jing Xingqing; Yang Yongwei; Chang Hong; Wu Zongxin; Gu Yuxiang

2002-01-01

The modular high temperature gas cooled reactor (MHTGR) has good safety characteristics because of the use of coated particles in the fuel element. After the particles cool outside of the reactor for some time, the spent fuel can be re-utilized. The author describes a physics feasibility study for the burning of spent fuel from a 350 MW ring-shaped modular high temperature gas cooled reactor in an accelerator-driven sub-critical reactor. A conceptual design is given for the 30 MW accelerator-driven sub-critical reactor. The neutron transport in the sub-critical reactor was simulated using the MCNP code, and the burnup was calculated using the ORIGEN2 code. The results show that the accelerator-driven sub-critical gas-cooled reactor has reliable sub-criticality and low power density and that the spent fuel from a 350 MW ring-shaped modular high temperature gas cooled reactor can be burned to provide 20% more energy

Study on integrated TRU multi-recycling in sodium cooled fast reactor CDFR

International Nuclear Information System (INIS)

Hu Yun; Xu Mi; Wang Kan

2010-01-01

In view of recently proposed closed fuel cycle strategy which would recycle the integrated transuranics (TRU) from PWR spent fuel in the fast reactors, the neutronics characteristics of TRU recycled in China Demonstration Fast Reactor (CDFR) are studied in this paper. The results show that loading integrated TRU to substitute pure Pu as driver fuel will mainly make the influence on sodium void worth and negligible effects on other parameters, and hence TRU recycling in CDFR is feasible from viewpoint of core neutronics. If TRU is multi-recycled, the variation of TRU composition depends on fuel types and the ratio of TRU and U when recycling. It is indicated that, when TRU is multi-recycled in CDFR with MOX fuel, the minor actinides (MA) fraction in TRU will firstly decrease to ∼7.24% (minimum) within 8 TRU recycle times and then slowly increase to ∼7.7% after 20 TRU recycle times; while when TRU is multi-recycled in CDFR with metal fuel (TRU-U-10Zr), the MA fraction in TRU will gradually approach to an equilibrium state with the MA fraction of ∼3.8%, demonstrating better MA transmutation effect in metal fuel core. No matter 7.7 or 3.8%, they are both lower than ∼10% in PWR spent fuel with burnup of 45 GWd/tU, which presents satisfying effect of MA amount controlling for TRU multi-recycling strategy. On the other hand, the corresponding recycling parameters such as TRU heat release and neutron emission rate are also much lower in metal fuel than those in MOX fuel. Moreover, TRU recycled in metal fuel will bring greater fissile Pu isotopes equilibrium fraction due to better breeding capability of metal fuel. Finally, it could be summarized that integrated TRU multi-recycling in fast reactor can make contributions to both breeding and transmutation, and such strategy is a prospective closed fuel cycle manner to achieve the object of effective control of cumulated MA amount and sustainable development of nuclear energy.

Exploring the Deployment Potential of Small Modular Reactors

Science.gov (United States)

Abdulla, Ahmed Y.

This thesis reports the results of several investigations into the viability of an emergent technology. Due to the lack of data in such cases, and the sensitivity surrounding nuclear power, exploring the potential of small modular reactors (SMRs) proved challenging. Moreover, these reactors come in a wide range of sizes and can employ a number of technologies, which made investigating the category as a whole difficult. We started by looking at a subset of SMRs that were the most promising candidates for near to mid-term deployment: integral light water SMRs. We conducted a technically detailed elicitation of expert assessments of their capital costs and construction duration, focusing on five reactor deployment scenarios that involved a large reactor and two light water SMRs. Consistent with the uncertainty introduced by past cost overruns and construction delays, median estimates of the cost of new large plants varied by more than a factor of 2.5. Expert judgments about likely SMR costs displayed an even wider range. There was consensus that an SMR plant's construction duration would be shorter than a large reactor's. Experts identified more affordable unit cost, factory fabrication, and shorter construction schedules as factors that may make light water SMRs economically viable, though these reactors do not constitute a paradigm shift when it comes to nuclear power's safety and security. Using these expert assessments of cost and construction duration, we calculated levelized cost of electricity values for four of the five scenarios. For the large plant, median levelized cost estimates ranged from 56 to 120 per MWh. Median estimates of levelized cost ranged from 77 to 240 per MWh for a 45MWe SMR, and from 65 to 120 per MWh for a 225MWe unit. We concluded that controlling construction duration is important, though not as important a factor in the analysis as capital cost, and, given the price of electricity in some parts of the U.S., it is possible to construct an

Methods and strategies for future reactor safety goals

Science.gov (United States)

Arndt, Steven Andrew

There have been significant discussions over the past few years by the United States Nuclear Regulatory Commission (NRC), the Advisory Committee on Reactor Safeguards (ACRS), and others as to the adequacy of the NRC safety goals for use with the next generation of nuclear power reactors to be built in the United States. The NRC, in its safety goals policy statement, has provided general qualitative safety goals and basic quantitative health objectives (QHOs) for nuclear reactors in the United States. Risk metrics such as core damage frequency (CDF) and large early release frequency (LERF) have been used as surrogates for the QHOs. In its review of the new plant licensing policy the ACRS has looked at the safety goals, as has the NRC. A number of issues have been raised including what the Commission had in mind when it drafted the safety goals and QHOs, how risk from multiple reactors at a site should be combined for evaluation, how the combination of a new and old reactor at the same site should be evaluated, what the criteria for evaluating new reactors should be, and whether new reactors should be required to be safer than current generation reactors. As part of the development and application of the NRC safety goal policy statement the Commissioners laid out the expectations for the safety of a nuclear power plant but did not address the risk associated with current multi-unit sites, potential modular reactor sites, and hybrid sites that could contain current generation reactors, new passive reactors, and/or modular reactors. The NRC safety goals and the QHOs refer to a "nuclear power plant," but do not discuss whether a "plant" refers to only a single unit or all of the units on a site. There has been much discussion on this issue recently due to the development of modular reactors. Additionally, the risk of multiple reactor accidents on the same site has been largely ignored in the probabilistic risk assessments (PRAs) done to date, and in most risk

Software Project Management Plan for the Integrated Systems Code (ISC) of New Production Reactor -- Modular High Temperature Gas Reactor

International Nuclear Information System (INIS)

Taylor, D.

1990-11-01

The United States Department of Energy (DOE) has selected the Modular High Temperature Gas-Cooled Reactor (MHTGR) as one of the concepts for the New Production Reactor (NPR). DOE has also established several Technical Working Groups (TWG's) at the national laboratories to provide independent design confirmation of the NPR-MHTGR design. One of those TWG's is concerned with Thermal Fluid Flow (TFF) and analysis methods to provide independent design confirmation of the NPR-MHTGR. Analysis methods are also needed for operational safety evaluations, performance monitoring, sensitivity studies, and operator training. The TFF Program Plan includes, as one of its principal tasks, the development of a computer program (called the Integrated Systems Code, or ISC). This program will provide the needed long-term analysis capabilities for the MHTGR and its subsystems. This document presents the project management plan for development of the ISC. It includes the associated quality assurance tasks, and the schedule and resource requirements to complete these activities. The document conforms to the format of ANSI/IEEE Std. 1058.1-1987. 2 figs

Evaluation of design variants for improved inherent regulation of advanced small modular reactors - 15325

International Nuclear Information System (INIS)

Vilim, R.B.; Passerini, S.

2015-01-01

This paper examines design variants that can improve inherent regulation in Advanced Small Modular Reactors (ASMR). It looks at the nature of unprotected upsets and then develops appropriate design measures to ensure that no upset can override a capability for safe self-regulation. This work adopts a reference sodium fast reactor (SFR) design to serve as a baseline for operational and safety performance and for comparison with variants on this design. The effect of design measures on plant stability is then examined. It is found that compared to full-power operation, the stability margin is reduced under islanded-operation. Islanded-operation is more likely for an ASMR deployed in a small regional electric grid with high penetration of renewable energy sources. The stability of core power production is a function of the inlet temperature coefficient, coolant transport times, and temperature-front attenuation in heat exchangers. The interaction of these phenomena with the control system is described

Implementing a modular system of computer codes

International Nuclear Information System (INIS)

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

1983-07-01

A modular computation system has been developed for nuclear reactor core analysis. The codes can be applied repeatedly in blocks without extensive user input data, as needed for reactor history calculations. The primary control options over the calculational paths and task assignments within the codes are blocked separately from other instructions, admitting ready access by user input instruction or directions from automated procedures and promoting flexible and diverse applications at minimum application cost. Data interfacing is done under formal specifications with data files manipulated by an informed manager. This report emphasizes the system aspects and the development of useful capability, hopefully informative and useful to anyone developing a modular code system of much sophistication. Overall, this report in a general way summarizes the many factors and difficulties that are faced in making reactor core calculations, based on the experience of the authors. It provides the background on which work on HTGR reactor physics is being carried out

Symmetric modular torsatron

Science.gov (United States)

Rome, J.A.; Harris, J.H.

1984-01-01

A fusion reactor device is provided in which the magnetic fields for plasma confinement in a toroidal configuration is produced by a plurality of symmetrical modular coils arranged to form a symmetric modular torsatron referred to as a symmotron. Each of the identical modular coils is helically deformed and comprise one field period of the torsatron. Helical segments of each coil are connected by means of toroidally directed windbacks which may also provide part of the vertical field required for positioning the plasma. The stray fields of the windback segments may be compensated by toroidal coils. A variety of magnetic confinement flux surface configurations may be produced by proper modulation of the winding pitch of the helical segments of the coils, as in a conventional torsatron, winding the helix on a noncircular cross section and varying the poloidal and radial location of the windbacks and the compensating toroidal ring coils.

Investigation of different modifications of the modular high temperature reactor for cost reduction

International Nuclear Information System (INIS)

Heek, A. van.

1993-06-01

Regarding the specific costs reactor with low power output have disadvantages compared to large reactors. In this study it was investigated for the HTR Module how to reduce the specific investment costs and consequently the electricity production costs by modification of the reactor and the power plant design. To reduce the investment costs different design changes were investigated, e.g. possible alternatives for the arrangement of the reactors and the steam generators and a different building layout. The cost reduction gained by a more compact type of construction and a reduction of the number of large components remains below 10%. The use of a gas turbine instead of a steam turbine with steam generator would result in a more efficient cost reduction. The increase in the power output has a much bigger potential to reduce the electricity production costs. The best way to realize this is the changing of the core geometry from cylindric to annular. A module concept with a power maximized under certain boundary conditions (limitation of the core temperatures in case of accident, use of proven technology) was worked out in detail. The optimum core geometry and power density, determined by tolerable accident temperatures and core pressure loss, make a thermal power of 350 MW possible. The physical properties and the accident behaviour in case of pressure release and water ingress were investigated. The specific investment costs and the electricity production costs of a 4-Modular Plant for pure electricity production are reduced about 28% resp. 19% compared with the 200 MWth-Module (KWU/Interatom). (orig.) [de

Westinghouse Small Modular Reactor balance of plant and supporting systems design

Energy Technology Data Exchange (ETDEWEB)

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

2012-07-01

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

Westinghouse Small Modular Reactor balance of plant and supporting systems design

International Nuclear Information System (INIS)

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

2012-01-01

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

Pebble bed modular reactors versus other generation technologies. Costs and challenges for South Africa

International Nuclear Information System (INIS)

Grubert, Emily; Parks, Brian; Schneider, Erich; Sekar, Srinivas

2011-01-01

South Africa is Africa's major economy, with plans to double its electricity generation capacity by 2026. South Africa has spent almost two decades developing a nuclear reactor known as a Pebble Bed Modular Reactor (PBMR), which could provide substantial benefits to the electricity grid but was recently mothballed due to high costs. This work estimates the lifecycle financial costs of South African PBMRs, then compares these costs to those of five other generation options: coal, nuclear as pressurized water reactors (PWRs), wind, and solar as photovoltaics (PV) or concentrating solar power (CSP). Each technology is evaluated with low, base case, and high assumptions for capital costs, construction time, and interest rates. Decommissioning costs, project lifetime, capacity factors, and sensitivity to carbon price are also considered. PBMR could be cost competitive with coal under certain low cost conditions, even without a carbon price. However, international lending practices and other factors suggest that a high capital cost, high interest rate nuclear plant is likely to be competing with a low capital cost, low interest rate coal plant in a market where cost recovery is challenging. PBMR could potentially become more competitive if low rate international loans were available to nuclear projects or became unavailable to coal projects. (author)

Valuing modular nuclear power plants in finite time decision horizon

International Nuclear Information System (INIS)

Jain, Shashi; Roelofs, Ferry; Oosterlee, Cornelis W.

2013-01-01

Small and medium sized reactors, SMRs, (according to IAEA, ‘small’ refers to reactors with power less than 300 MWe, and ‘medium’ with power less than 700 MWe) are considered as an attractive option for investment in nuclear power plants. SMRs may benefit from flexibility of investment, reduced upfront expenditure, enhanced safety, and easy integration with small sized grids. Large reactors on the other hand have been an attractive option due to the economy of scale. In this paper we focus on the economic impact of flexibility due to modular construction of SMRs. We demonstrate, using real option analysis, the value of sequential modular SMRs. Numerical results under different considerations of decision time, uncertainty in electricity prices, and constraints on the construction of units, are reported for a single large unit and for modular SMRs. - Highlights: ► Real option value of modular construction in finite time decision horizon. ► Stochastic grid method is used to value the real option. ► Decisions in finite time can differ significantly from infinite decision time. ► Decisions depend on length of decision horizon and price volatilities

UWTOR-M, a stellarator power reactor utilizing modular coils

International Nuclear Information System (INIS)

Sviatoslavsky, I.N.; Van Sciver, S.W.; Kulcinski, G.L.

1981-10-01

We briefly describe the parametric considerations which led to the UWTOR-M reference design point. The design has 18 twisted coils utilizing a multipolarity of 3, a major radius of 24 m, a coil radius of 4.77 m and a plasma aspect ratio of 14. An assumed β of 5% was used. This configuration leads to a rotational transform on the edge of 1.125 giving favorable plasma physics conditions. The natural stellarator divertor is used for impurity control in conjunction with innovative high performance divertor targets. A unique blanket design is proposed which minimizes tritium inventory in the reactor. Finally, since maintainability is a prime consideration, we describe a scheme for servicing the first wall/blanket and other reactor components

Advanced Modular Power Approach to Affordable, Supportable Space Systems

Science.gov (United States)

Oeftering, Richard C.; Kimnach, Greg L.; Fincannon, James; Mckissock,, Barbara I.; Loyselle, Patricia L.; Wong, Edmond

2013-01-01

Recent studies of missions to the Moon, Mars and Near Earth Asteroids (NEA) indicate that these missions often involve several distinct separately launched vehicles that must ultimately be integrated together in-flight and operate as one unit. Therefore, it is important to see these vehicles as elements of a larger segmented spacecraft rather than separate spacecraft flying in formation. The evolution of large multi-vehicle exploration architecture creates the need (and opportunity) to establish a global power architecture that is common across all vehicles. The Advanced Exploration Systems (AES) Modular Power System (AMPS) project managed by NASA Glenn Research Center (GRC) is aimed at establishing the modular power system architecture that will enable power systems to be built from a common set of modular building blocks. The project is developing, demonstrating and evaluating key modular power technologies that are expected to minimize non-recurring development costs, reduce recurring integration costs, as well as, mission operational and support costs. Further, modular power is expected to enhance mission flexibility, vehicle reliability, scalability and overall mission supportability. The AMPS project not only supports multi-vehicle architectures but should enable multi-mission capability as well. The AMPS technology development involves near term demonstrations involving developmental prototype vehicles and field demonstrations. These operational demonstrations not only serve as a means of evaluating modular technology but also provide feedback to developers that assure that they progress toward truly flexible and operationally supportable modular power architecture.

Steam generator design considerations for modular HTGR plant

International Nuclear Information System (INIS)

McDonald, C.F.; DeFur, D.D.

1986-01-01

Studies are in progress to develop a standard High Temperature Gas-Cooled Reactor (HTGR) plant design that is amenable to serial production and is licensable. Based on the results of trade studies performed in the DOE-funded HTGR program, activities are being focused to emphasize a modular concept based on a 350 MW(t) annular reactor core with prismatic fuel elements. Utilization of a multiplicity of the standard module affords flexibility in power rating for utility electricity generation. The selected modular HTGR concept has the reactor core and heat transport systems housed in separate steel vessels. This paper highlights the steam generator design considerations for the reference plant, and includes a discussion of the major features of the heat exchanger concept and the technology base existing in the U.S

Analysis of radiation safety for Small Modular Reactor (SMR) on PWR-100 MWe type

Science.gov (United States)

Udiyani, P. M.; Husnayani, I.; Deswandri; Sunaryo, G. R.

2018-02-01

Indonesia as an archipelago country, including big, medium and small islands is suitable to construction of Small Medium/Modular reactors. Preliminary technology assessment on various SMR has been started, indeed the SMR is grouped into Light Water Reactor, Gas Cooled Reactor, and Solid Cooled Reactor and from its site it is group into Land Based reactor and Water Based Reactor. Fukushima accident made people doubt about the safety of Nuclear Power Plant (NPP), which impact on the public perception of the safety of nuclear power plants. The paper will describe the assessment of safety and radiation consequences on site for normal operation and Design Basis Accident postulation of SMR based on PWR-100 MWe in Bangka Island. Consequences of radiation for normal operation simulated for 3 units SMR. The source term was generated from an inventory by using ORIGEN-2 software and the consequence of routine calculated by PC-Cream and accident by PC Cosyma. The adopted methodology used was based on site-specific meteorological and spatial data. According to calculation by PC-CREAM 08 computer code, the highest individual dose in site area for adults is 5.34E-02 mSv/y in ESE direction within 1 km distance from stack. The result of calculation is that doses on public for normal operation below 1mSv/y. The calculation result from PC Cosyma, the highest individual dose is 1.92.E+00 mSv in ESE direction within 1km distance from stack. The total collective dose (all pathway) is 3.39E-01 manSv, with dominant supporting from cloud pathway. Results show that there are no evacuation countermeasure will be taken based on the regulation of emergency.

Preliminary S-CO_2 Compressor Design for Micro Modular Reactor

International Nuclear Information System (INIS)

Lee, Jekyoung; Cho, Seong Kuk; Kim, Seong Gu; Lee, Jeong Ik

2016-01-01

Due to economic benefit of S-CO_2 Brayton cycle which is came from high efficiency and compactness, active research is currently conducted by various research groups and various approaches are suggested to take benefits of S-CO_2 Brayton cycle. KAIST research team also has been working on advanced concept for application of S-CO_2 Brayton cycle to nuclear system and Micro Modular Reactor (MMR) concept was suggested. The preliminary compressor design of S-CO_2 compressor for MMR system was carried out to observe feasibility of compressor design. Preliminary S-CO_2 compressor design for MMR system was successfully conducted and some issues are discovered from the design study. From the previous work done by Cho, conceptual design for MMR system was provided. Thus, further preliminary design should be carried out to obtain feasible S-CO_2 compressor design for MMR system. KAIST_TMD which is turbomachinery in-house code for real gases including S-CO_2 is continuously updated and currently it has 3D geometry construction and design optimization capability

Coupling of Modular High-Temperature Gas-Cooled Reactor with Supercritical Rankine Cycle

Directory of Open Access Journals (Sweden)

Shutang Zhu

2008-01-01

Full Text Available This paper presents investigations on the possible combination of modular high-temperature gas-cooled reactor (MHTGR technology with the supercritical (SC steam turbine technology and the prospective deployments of the MHTGR SC power plant. Energy conversion efficiency of steam turbine cycle can be improved by increasing the main steam pressure and temperature. Investigations on SC water reactor (SCWR reveal that the development of SCWR power plants still needs further research and development. The MHTGR SC plant coupling the existing technologies of current MHTGR module design with operation experiences of SC FPP will achieve high cycle efficiency in addition to its inherent safety. The standard once-reheat SC steam turbine cycle and the once-reheat steam cycle with life-steam have been studied and corresponding parameters were computed. Efficiencies of thermodynamic processes of MHTGR SC plants were analyzed, while comparisons were made between an MHTGR SC plant and a designed advanced passive PWR - AP1000. It was shown that the net plant efficiency of an MHTGR SC plant can reach 45% or above, 30% higher than that of AP1000 (35% net efficiency. Furthermore, an MHTGR SC plant has higher environmental competitiveness without emission of greenhouse gases and other pollutants.

A Conceptual Study of Using an Isothermal Compressor on an S-CO{sub 2} Cooled KAIST Micro Modular Reactor (KAIST-MMR)

Energy Technology Data Exchange (ETDEWEB)

Heo, Jin Young; Ahn, Yoonhan; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of)

2016-05-15

Although the isothermal compressor technology is not fully matured for commercialization, its potential to reduce the compressor work and the cycle physical size due to the merging of the compressor and the precooler into one turbomachine can be utilized well especially in a concept such as the KAIST-MMR. Future works include optimization of the cycle layout and parameters to enhance performance, calculation of the heat exchanger sizes, and turbomachinery design of components. The development of small modular reactors (SMR) has been gathering attention due to factory manufacturing, but the power system required for a steam cycle in a water-cooled SMR is significantly large. A new concept called the KAIST Micro Modular Reactor (MMR) has been designed previously to make use of the supercritical CO{sub 2} (S-CO{sub 2}) cycle to significantly decrease the size and to improve its cycle efficiency. To further enhance the advantages of the KAIST-MMR concept, the conventional compressor is replaced with an isothermal compressor in the S-CO{sub 2} cycle.

The effect of water vapor in the reactor cavity in a MHTGR [Modular High Temperature Gas Cooled Reactor] on the radiation heat transfer

International Nuclear Information System (INIS)

Cappiello, M.W.

1991-01-01

Analyses have been completed to determine the effect of the presence of water vapor in the reactor cavity in a modular high temperature gas cooled reactor on the predicted radiation heat transfer from the vessel wall to the reactor cavity cooling system. The analysis involves the radiation heat transfer between two parallel plates with an absorbing and emitting medium present. Because the absorption in the water vapor is spectrally dependent, the solution is difficult even for simple geometries. A computer code was written to solve the problem using the Monte Carlo method. The code was validated against closed form solutions, and shows excellent agreement. In the analysis of the reactor problem, the results show that the reduction in heat transfer, and the consequent increase in the vessel wall temperature, can be significant. This effect can be cast in terms of a reduction in the wall surface emissivities from 0.8 to 0.59. Because of the insulating effect of the water vapor, increasing the gap distance between the vessel wall and the cooling system will cause the vessel wall temperature to increase further. Care should be taken in the design of the facility to minimize the gap distance and keep temperature increase within allowable limits. 3 refs., 6 figs., 4 tabs

Novel modular natural circulation BWR design and safety evaluation

International Nuclear Information System (INIS)

Ishii, Mamoru; Shi, Shanbin; Yang, Won Sik; Wu, Zeyun; Rassame, Somboon; Liu, Yang

2015-01-01

Highlights: • Introduction of BWR-type natural circulation small modular reactor preliminary design (NMR-50). • Design of long fuel cycle length for the NMR-50. • Design of double passive safety systems for the NMR-50. • RELAP5 analyses of design basis accidents for the NMR-50. - Abstract: The Purdue NMR (Novel Modular Reactor) represents a BWR-type small modular reactor with a significantly reduced reactor pressure vessel (RPV) height. Specifically, it has one third the height of a conventional BWR RPV with an electrical output of 50 MWe. The preliminary design of the NMR-50 including reactor, fuel cycle, and safety systems is described and discussed. The improved neutronics design of the NMR-50 extends the fuel cycle length up to 10 years. The NMR-50 is designed with double passive engineering safety system, which is intended to withstand a prolonged station black out with loss of ultimate heat sink accident such as experienced at Fukushima. In order to evaluate the safety features of the NMR-50, two representative design basis accidents, i.e. main steam line break (MSLB) and bottom drain line break (BDLB), are simulated by using the best-estimate thermal–hydraulic code RELAP5. The RPV water inventory, containment pressure, and the performance of engineering safety systems are investigated for about 33 h after the initiation of the accidents

Use of a multi-attribute utility theory for evaluating the best coolant material in transmutation reactors

International Nuclear Information System (INIS)

Yu, Dong Han; Han, Suk Joong; Kim, Do Hyung; Park, Won Suk

1998-12-01

In order to develop and design a good transmutation system, it is necessary first to select the best available coolant material for a reactor coolant system. Choosing the best coolant material may not be easy since there are several criteria associated with thermal performance, safety problem, cost problem, neutronic aspects. etc. The best option should be chosen based on the maximization of our needs in this situation. It is a challenging task. Decision theory can be employed to solve this type of problem. This report presents the feasibility study for evaluating the best coolant material in transmutation reactors based on the multi=attribute utility theory. The main problem presented here is how to logically evaluate candidate coolant materials under multiple criteria such as thermal performance, safety problem, cost problem, cost problem, neutronic aspects, etc. Since the current problem involves multiple criteria or attributes, first of all, the multi-attribute utility theory (MAUT) such as SMART and AHP has been extensively reviewed. Then, many candidate coolant material for transmutation reactors have been identified. The next step is to construct a value tree that express to reflect the relative importance of the attributes for overall evaluation. Finally, given these assignments, the final goal were obtained by manipulating these ranks through the value tree. The proposed approach is intended to help people be rational and logical in making decisions such complex tasks. (author). 8 refs., 7 tabs., 22 figs

Advanced Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Technical Exchange Meeting

Energy Technology Data Exchange (ETDEWEB)

Smith, Curtis [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2013-09-01

During FY13, the INL developed an advanced SMR PRA framework which has been described in the report Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Detailed Technical Framework Specification, INL/EXT-13-28974 (April 2013). In this framework, the various areas are considered: Probabilistic models to provide information specific to advanced SMRs Representation of specific SMR design issues such as having co-located modules and passive safety features Use of modern open-source and readily available analysis methods Internal and external events resulting in impacts to safety All-hazards considerations Methods to support the identification of design vulnerabilities Mechanistic and probabilistic data needs to support modeling and tools In order to describe this framework more fully and obtain feedback on the proposed approaches, the INL hosted a technical exchange meeting during August 2013. This report describes the outcomes of that meeting.

Seismic isolation of small modular reactors using metamaterials

Directory of Open Access Journals (Sweden)

Witarto Witarto

2018-04-01

Full Text Available Adaptation of metamaterials at micro- to nanometer scales to metastructures at much larger scales offers a new alternative for seismic isolation systems. These new isolation systems, known as periodic foundations, function both as a structural foundation to support gravitational weight of the superstructure and also as a seismic isolator to isolate the superstructure from incoming seismic waves. Here we describe the application of periodic foundations for the seismic protection of nuclear power plants, in particular small modular reactors (SMR. For this purpose, a large-scale shake table test on a one-dimensional (1D periodic foundation supporting an SMR building model was conducted. The 1D periodic foundation was designed and fabricated using reinforced concrete and synthetic rubber (polyurethane materials. The 1D periodic foundation structural system was tested under various input waves, which include white noise, stepped sine and seismic waves in the horizontal and vertical directions as well as in the torsional mode. The shake table test results show that the 1D periodic foundation can reduce the acceleration response (transmissibility of the SMR building up to 90%. In addition, the periodic foundation-isolated structure also exhibited smaller displacement than the non-isolated SMR building. This study indicates that the challenge faced in developing metastructures can be overcome and the periodic foundations can be applied to isolating vibration response of engineering structures.

Seismic isolation of small modular reactors using metamaterials

Science.gov (United States)

Witarto, Witarto; Wang, S. J.; Yang, C. Y.; Nie, Xin; Mo, Y. L.; Chang, K. C.; Tang, Yu; Kassawara, Robert

2018-04-01

Adaptation of metamaterials at micro- to nanometer scales to metastructures at much larger scales offers a new alternative for seismic isolation systems. These new isolation systems, known as periodic foundations, function both as a structural foundation to support gravitational weight of the superstructure and also as a seismic isolator to isolate the superstructure from incoming seismic waves. Here we describe the application of periodic foundations for the seismic protection of nuclear power plants, in particular small modular reactors (SMR). For this purpose, a large-scale shake table test on a one-dimensional (1D) periodic foundation supporting an SMR building model was conducted. The 1D periodic foundation was designed and fabricated using reinforced concrete and synthetic rubber (polyurethane) materials. The 1D periodic foundation structural system was tested under various input waves, which include white noise, stepped sine and seismic waves in the horizontal and vertical directions as well as in the torsional mode. The shake table test results show that the 1D periodic foundation can reduce the acceleration response (transmissibility) of the SMR building up to 90%. In addition, the periodic foundation-isolated structure also exhibited smaller displacement than the non-isolated SMR building. This study indicates that the challenge faced in developing metastructures can be overcome and the periodic foundations can be applied to isolating vibration response of engineering structures.

Sensitivity and Uncertainty Analysis of Coupled Reactor Physics Problems : Method Development for Multi-Physics in Reactors

NARCIS (Netherlands)

Perkó, Z.

2015-01-01

This thesis presents novel adjoint and spectral methods for the sensitivity and uncertainty (S&U) analysis of multi-physics problems encountered in the field of reactor physics. The first part focuses on the steady state of reactors and extends the adjoint sensitivity analysis methods well

TRANSFER-FUNCTIONS OF A LINEARIZED MULTI-REGION REACTOR

Energy Technology Data Exchange (ETDEWEB)

Higgins, Thomas J.

1963-09-15

The development of the transfer functions for a linearized multi-region reactor is studied, and an illustration is made of application of the corresponding theory by a numerical illustrative example. (auth)

Advanced Instrumentation and Control Methods for Small and Medium Reactors with IRIS Demonstration. Final Report. Volume 1

International Nuclear Information System (INIS)

Hines, J. Wesley; Upadhyaya, Belle R.; Doster, J. Michael; Edwards, Robert M.; Lewis, Kenneth D.; Turinsky, Paul; Coble, Jamie

2011-01-01

on meeting two of the eight needs outlined in the recently published 'Technology Roadmap on Instrumentation, Control, and Human-Machine Interface (ICHMI) to Support DOE Advanced Nuclear Energy Programs' which was created 'to provide a systematic path forward for the integration of new ICHMI technologies in both near-term and future nuclear power plants and the reinvigoration of the U.S. nuclear ICHMI community and capabilities.' The research consortium is led by The University of Tennessee (UT) and is focused on three interrelated topics: Topic 1 (simulator development and measurement sensitivity analysis) is led by Dr. Mike Doster with Dr. Paul Turinsky of North Carolina State University (NCSU). Topic 2 (multivariate autonomous control of modular reactors) is led by Dr. Belle Upadhyaya of the University of Tennessee (UT) and Dr. Robert Edwards of Penn State University (PSU). Topic 3 (monitoring, diagnostics, and prognostics system development) is led by Dr. Wes Hines of UT. Additionally, South Carolina State University (SCSU, Dr. Ken Lewis) participated in this research through summer interns, visiting faculty, and on-campus research projects identified throughout the grant period. Lastly, Westinghouse Science and Technology Center (Dr. Mario Carelli) was a no-cost collaborator and provided design information related to the IRIS demonstration platform and defining needs that may be common to other SMR designs. The results of this research are reported in a six-volume Final Report (including the Executive Summary, Volume 1). Volumes 2 through 6 of the report describe in detail the research and development under the topical areas. This volume serves to introduce the overall NERI-C project and to summarize the key results. Section 2 provides a summary of the significant contributions of this project. A list of all the publications under this project is also given in Section 2. Section 3 provides a brief summary of each of the five volumes (2-6) of the report. The

DEVELOPMENT AND DEPLOYMENT ASSESSMENT OF A MELT-DOWN PROOF MODULAR MICRO REACTOR (MDP-MMR)

Energy Technology Data Exchange (ETDEWEB)

Hawari, Ayman I.; Venneri, Francesco

2018-04-02

The objective of this project is to perform feasibility assessment and technology gap analysis and establish a development roadmap for an innovative and highly compact Micro Modular Reactor (MMR) concept that integrates power production, power conversion and electricity generation in a single unit. The MMR is envisioned to use fully ceramic micro-encapsulated (FCM) fuel, a particularly robust form of TRISO fuel, and to be gas-cooled (e.g., He or CO2) and capable of generating power in the range of 10 to 40 MW-thermal. It is designed to be absolutely melt-down proof (MDP) under all circumstances including complete loss of coolant scenarios with no possible release of radioactive material, to be factory produced, to have a cycle length of greater than 20 years, and to be highly proliferation resistant. In addition, it will be transportable, retrievable and suitable for use in remote areas. As such, the MDP-MMR will represent a versatile reactor concept that is suitable for use in various applications including electricity generation, process heat utilization and propulsion.

The application of a multi-physics tool kit to spatial reactor dynamics

International Nuclear Information System (INIS)

Clifford, I.; Jasak, H.

2009-01-01

Traditionally coupled field nuclear reactor analysis has been carried out using several loosely coupled solvers, each having been developed independently from the others. In the field of multi-physics, the current generation of object-oriented tool kits provides robust close coupling of multiple fields on a single framework. This paper describes the initial results obtained as part of continuing research in the use of the OpenFOAM multi-physics tool kit for reactor dynamics application development. An unstructured, three-dimensional, time-dependent multi-group diffusion code Diffusion FOAM has been developed using the OpenFOAM multi-physics tool kit as a basis. The code is based on the finite-volume methodology and uses a newly developed block-coupled sparse matrix solver for the coupled solution of the multi-group diffusion equations. A description of this code is given with particular emphasis on the newly developed block-coupled solver, along with a selection of results obtained thus far. The code has performed well, indicating that the OpenFOAM tool kit is suited to reactor dynamics applications. This work has shown that the neutronics and simplified thermal-hydraulics of a reactor May be represented and solved for using a common calculation platform, and opens up the possibility for research into robust close-coupling of neutron diffusion and thermal-fluid calculations. This work has further opened up the possibility for research in a number of other areas, including research into three-dimensional unstructured meshes for reactor dynamics applications. (authors)

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

International Nuclear Information System (INIS)

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

2015-01-01

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

U.S. Department of Energy instrumentation and controls technology research for advanced small modular reactors

International Nuclear Information System (INIS)

Wood, Richard Thomas

2013-01-01

Instrumentation, controls, and human-machine interfaces (ICHMI) are essential enabling technologies that strongly influence nuclear power plant performance and operational costs. The U.S. Department of Energy (DOE) has recognized that ICHMI research, development, and demonstration (RD and D) is needed to resolve the technical challenges that may compromise the effective and efficient utilization of modern ICHMI technology and consequently inhibit realization of the benefits offered by expanded utilization of nuclear power. Consequently, key DOE programs have substantial ICHMI RD and D elements to their respective research portfolio. This article describes current ICHMI research to support the development of advanced small modular reactors. (author)

The gas turbine modular helium reactor. An international project to develop a safe, efficient, flexible product

International Nuclear Information System (INIS)

Silberstein, A.J.

1998-01-01

As originally scheduled, the Conceptual Design Report of the 600 Mwt Gas Turbine Modular Helium Reactor has been issued in October 1997 by OKBM in Nizhny Novgorod, a keystone Russian Engineering Institute fully involved in the realization of this International Project. The plutonium burning, graphite moderated helium cooled reactor design results from the work done on the basis of General Atomics original concept combined with the goal of optimizing safety power and efficiency with multi contributions in specific fields from the Russian organizations: MINATOM, OKBM, VNIINM, Lutch, Kurchatov Institute, Seversk Chemical Combinat, Fuji Electric and FRAMATOME. The objective to concentrate the engineering work in Russia has met a full success due principally to the quality and experience of the people, to the international support and to the progressive integration of new techniques of communication, of project management culture and utilization of modern computerized design tools and methods. To day the best international standard of quality is reached in the engineering activity and expected to stay at this level for future developments, when including experimental facilities operation and components manufacturing activities, thanks to the diffusion of the common culture, acquired by the main actors during the conceptual design phase, that will be exported to Russian third parties. At this stage we are planning to start design verification and sensitive components and systems qualification, with the same original actors. The European Commission has already shown some significant interest through the MICHELANGELO Initiative in supporting the HTR concepts assessment and identification of the R and D needs. We are looking forward for further support from the International Community and particularly from European Institutions in the frame of the 5th PCRD to pursue the GT MHR R and D program. Furthermore we are looking for funding the building of a prototype in Russia

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

Energy Technology Data Exchange (ETDEWEB)

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

1994-02-01

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

Modular simulation of reefer container dynamics

DEFF Research Database (Denmark)

Sørensen, Kresten Kjær; Nielsen, Jens Frederik Dalsgaard; Stoustrup, Jakob

2014-01-01

for faults enabling preventive maintenance. In this paper the feasibility of using different simulation methods is assessed with the goal of identifying a fast but accurate method that works well in a multi-rate environment. A modular multi-rate simulation environment for a dynamical system consisting...

New to nuclear countries: considerations for adoption of Small Modular Reactors - a guide to future adopters

Energy Technology Data Exchange (ETDEWEB)

Paterson, A.; Ho, M.; Storr, G., E-mail: adi.paterson@ansto.gov.au [Australian Nuclear Science and Technology Organisation, Lucas Heights, NSW (Australia)

2014-07-01

Small Modular Reactors (SMRs) are under development in a number of countries. This class of reactors, with enhanced safety and security inherent to their design, can potentially offer advantages to countries adopting nuclear power for the first time. This includes countries considering expanding existing nuclear power capability using the benefits of Generation III+ and Generation IV technological advances. For example, public perceptions and engagement in relation to nuclear safety and security are important and need to be addressed. The regulation of nuclear power plants and the management of the nuclear fuel cycle are also important considerations. For some countries, an industrial strategy linked to participation in the nuclear fuel cycle could be a priority, associated with the development of a sophisticated workforce to support the design, construction, commissioning and operation of a fleet of reactors. Some countries will choose to be early adopters of SMRs. Others may prefer to wait until the technologies are more established. This paper will focus on the potential appeal of different SMR designs in relation to the considerations that new to nuclear countries must address and how this underpins effective decision making. (author)

Advanced High-Temperature Reactor for Production of Electricity and Hydrogen: Molten-Salt-Coolant, Graphite-Coated-Particle-Fuel

International Nuclear Information System (INIS)

Forsberg, C.W.

2002-01-01

The objective of the Advanced High-Temperature Reactor (AHTR) is to provide the very high temperatures necessary to enable low-cost (1) efficient thermochemical production of hydrogen and (2) efficient production of electricity. The proposed AHTR uses coated-particle graphite fuel similar to the fuel used in modular high-temperature gas-cooled reactors (MHTGRs), such as the General Atomics gas turbine-modular helium reactor (GT-MHR). However, unlike the MHTGRs, the AHTR uses a molten salt coolant with a pool configuration, similar to that of the PRISM liquid metal reactor. A multi-reheat helium Brayton (gas-turbine) cycle, with efficiencies >50%, is used to produce electricity. This approach (1) minimizes requirements for new technology development and (2) results in an advanced reactor concept that operates at essentially ambient pressures and at very high temperatures. The low-pressure molten-salt coolant, with its high heat capacity and natural circulation heat transfer capability, creates the potential for (1) exceptionally robust safety (including passive decay-heat removal) and (2) allows scaling to large reactor sizes [∼1000 Mw(e)] with passive safety systems to provide the potential for improved economics

Assessments of Water Ingress Accidents in a Modular High-Temperature Gas-Cooled Reactor

International Nuclear Information System (INIS)

Zhang Zuoyi; Dong Yujie; Scherer, Winfried

2005-01-01

Severe water ingress accidents in the 200-MW HTR-module were assessed to determine the safety margins of modular pebble-bed high-temperature gas-cooled reactors (HTR-module). The 200-MW HTR-module was designed by Siemens under the criteria that no active safety protection systems were necessary because of its inherent safe nature. For simulating the behavior of the HTR-module during severe water ingress accidents, a water, steam, and helium multiphase cavity model was developed and implemented in the dynamic simulator for nuclear power plants (DSNP) simulation system. Comparisons of the DSNP simulations incorporating these models with experiments and with calculations using the time-dependent neutronics and temperature dynamics code were made to validate the simulation. The analysis of the primary circuit showed that the maximum water concentration increase in the reactor core was 3 s). The water vaporization in the steam generator and characteristics of water transport from the steam generator to the reactor core would reduce the rate of water ingress into the reactor core. The analysis of a full cavitation of the feedwater pump showed that if the secondary circuit could be depressurized, the feedwater pump would be stopped by the full cavitation. This limits the water transported from the deaerator to the steam generator. A comprehensive simulation of the HTR-module power plant showed that the water inventory in the primary circuit was limited to ∼3000 kg. The nuclear reactivity increase caused by the water ingress would lead to a fast power excursion, which would be inherently counterbalanced by negative feedback effects. The integrity of the fuel elements, because the safety-relevant temperature limit of 1600 deg. C is not reached in any case, is not challenged

Work Domain Analysis of a Predecessor Sodium-cooled Reactor as Baseline for AdvSMR Operational Concepts

Energy Technology Data Exchange (ETDEWEB)

Ronald Farris; David Gertman; Jacques Hugo

2014-03-01

This report presents the results of the Work Domain Analysis for the Experimental Breeder Reactor (EBR-II). This is part of the phase of the research designed to incorporate Cognitive Work Analysis in the development of a framework for the formalization of an Operational Concept (OpsCon) for Advanced Small Modular Reactors (AdvSMRs). For a new AdvSMR design, information obtained through Cognitive Work Analysis, combined with human performance criteria, can and should be used in during the operational phase of a plant to assess the crew performance aspects associated with identified AdvSMR operational concepts. The main objective of this phase was to develop an analytical and descriptive framework that will help systems and human factors engineers to understand the design and operational requirements of the emerging generation of small, advanced, multi-modular reactors. Using EBR-II as a predecessor to emerging sodium-cooled reactor designs required the application of a method suitable to the structured and systematic analysis of the plant to assist in identifying key features of the work associated with it and to clarify the operational and other constraints. The analysis included the identification and description of operating scenarios that were considered characteristic of this type of nuclear power plant. This is an invaluable aspect of Operational Concept development since it typically reveals aspects of future plant configurations that will have an impact on operations. These include, for example, the effect of core design, different coolants, reactor-to-power conversion unit ratios, modular plant layout, modular versus central control rooms, plant siting, and many more. Multi-modular plants in particular are expected to have a significant impact on overall OpsCon in general, and human performance in particular. To support unconventional modes of operation, the modern control room of a multi-module plant would typically require advanced HSIs that would

South African safety assessment framework for the pebble bed modular reactor - HTR2008-58192

International Nuclear Information System (INIS)

Joubert, J.; Kohtz, N.; Coe, I.

2008-01-01

It is planned to construct a first of a kind Pebble Bed Modular Reactor (PBMR) in South Africa. A need has been recognized to accompany the licensing process for the PBMR with independent safety assessments to ensure that the safety case submitted by the applicant complies with the licensing requirements of the NNR. At the HTR 2006 Conference, the framework and major challenges on safety assessment that the South African National Nuclear Regulator (NNR) faces in developing and applying appropriate strategies and tools were presented. This paper discusses the current status of the various NNR assessment activities and describes how this will be considered in the NNR Final Report on the PBMR Safety Case. The traditional safety assessment process has been adapted to take into account the developmental nature of the project. By performing safety assessments, the designer and applicant must ensure that the design as proposed for construction and as-built meets the safety requirements defined by the regulatory framework. The regulator performs independent safety assessments, including independent analyses in areas deemed safety significant and potentially safety significant. The developmental nature of the project also led to the identification of a series of regulatory assessment activities preceding the formal assessment of the safety case. Besides an assessment of the resolution of Key Licensing Issues which have been defined in an early stage of the project and are discussed in /l/, these activities comprise the participation in an SAR Early Intervention Process, the execution of a regulatory HAZOP and the development of a regulatory assessment specification for the formal assessment of the safety case. This paper briefly describes these activities and their current status. During the last two years, significant progress was made with the development or adjustment of tools for the independent analysis by the regulator of the steady state core design, of the transient

Design Safety Considerations for Water Cooled Small Modular Reactors Incorporating Lessons Learned from the Fukushima Daiichi Accident

International Nuclear Information System (INIS)

2016-03-01

The global future deployment of advanced nuclear reactors for electricity generation depends primarily on the ability of nuclear industries, utilities and regulatory authorities to further enhance their reliability and economic competitiveness while satisfying stringent safety requirements. The IAEA has a project to help coordinate Member States efforts in the development and deployment of small and medium sized or small modular reactor (SMR) technology. This project aims simultaneously to facilitate SMR technology developers and potential SMR uses, particularly States embarking on a nuclear power programme, in identifying key enabling technologies and enhancing capacity building by resolving issues relevant to deployment, including nuclear reactor safety. The objective of this publication is to explore common practices for Member States, which will be an essential resource for future development and deployment of SMR technology. The accident at the Fukushima Daiichi nuclear power plant was caused by an unprecedented combination of natural events: a strong earthquake, beyond the design basis, followed by a series of tsunamis of heights exceeding the design basis tsunami considered in the flood analysis for the site. Consequently, all the operating nuclear power plants and advanced reactors under development, including SMRs, have been incorporating lessons learned from the accident to assure and enhance the performance of the engineered safety features in coping with such external events

Feasibility study on nuclear core design for soluble boron free small modular reactor

Energy Technology Data Exchange (ETDEWEB)

Rabir, Mohamad Hairie, E-mail: m-hairie@nuclearmalaysia.gov.my; Hah, Chang Joo; Ju, Cho Sung [Department of NPP Engineering, KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2015-04-29

A feasibility study on nuclear core design of soluble boron free (SBF) core for small size (150MWth) small modular reactor (SMR) was investigated. The purpose of this study was to design a once through cycle SMR core, where it can be used to supply electricity to a remote isolated area. PWR fuel assembly design with 17×17 arrangement, with 264 fuel rods per assembly was adopted as the basis design. The computer code CASMO-3/MASTER was used for the search of SBF core and fuel assembly analysis for SMR design. A low critical boron concentration (CBC) below 200 ppm core with 4.7 years once through cycle length was achieved using 57 fuel assemblies having 170 cm of active height. Core reactivity controlled using mainly 512 number of 4 wt% and 960 12 wt% Gd rods.

Evaluation of radiation heat transfer in porous medial: Application for a pebble bed modular reactor cooled by CO2 gas

Directory of Open Access Journals (Sweden)

Sidi-Ali Kamel

2013-01-01

Full Text Available This work analyses the contribution of radiation heat transfer in the cooling of a pebble bed modular reactor. The mathematical model, developed for a porous medium, is based on a set of equations applied to an annular geometry. Previous major works dealing with the subject have considered the forced convection mode and often did not take into account the radiation heat transfer. In this work, only free convection and radiation heat transfer are considered. This can occur during the removal of residual heat after shutdown or during an emergency situation. In order to derive the governing equations of radiation heat transfer, a steady-state in an isotropic and emissive porous medium (CO2 is considered. The obtained system of equations is written in a dimensionless form and then solved. In order to evaluate the effect of radiation heat transfer on the total heat removed, an analytical method for solving the system of equations is used. The results allow quantifying both radiation and free convection heat transfer. For the studied situation, they show that, in a pebble bed modular reactor, more than 70% of heat is removed by radiation heat transfer when CO2 is used as the coolant gas.

Modular cryogenic interconnects for multi-qubit devices

Energy Technology Data Exchange (ETDEWEB)

Colless, J. I.; Reilly, D. J., E-mail: david.reilly@sydney.edu.au [ARC Centre of Excellence for Engineered Quantum Systems, School of Physics, The University of Sydney, Sydney, NSW 2006 (Australia)

2014-11-15

We have developed a modular interconnect platform for the control and readout of multiple solid-state qubits at cryogenic temperatures. The setup provides 74 filtered dc-bias connections, 32 control and readout connections with −3 dB frequency above 5 GHz, and 4 microwave feed lines that allow low loss (less than 3 dB) transmission 10 GHz. The incorporation of a radio-frequency interposer enables the platform to be separated into two printed circuit boards, decoupling the simple board that is bonded to the qubit chip from the multilayer board that incorporates expensive connectors and components. This modular approach lifts the burden of duplicating complex interconnect circuits for every prototype device. We report the performance of this platform at milli-Kelvin temperatures, including signal transmission and crosstalk measurements.

Final Physics Report for the Engineering Test Reactor

International Nuclear Information System (INIS)

Wolfe, I. B.

1956-01-01

This report is a summary of the physics design work performed on the Engineering Test Reactor. The ETR presents computational difficulties not found in other reactors because of the large number of experimental holes in the core. The physics of the ETR depends strongly upon the contents of the in-core experimental facilities. In order to properly evaluate the reactor' taking into account the experiments in the core, multi-region, two-dimensional calculations are required. These calculations require the use of a large computer such as the Remington Rand Univac and are complex and expensive enough to warrant a five-stage program: 1. In the early stages of design, only preliminary two-dimensional calculations were performed .in order to obtain a rough idea of the general behavior of the reactor and its critical mass with tentative experiments in place. 2. A large amount of work was carried out in which the reactor was approximated as one with a uniform homogeneous core. With this model, detailed studies were carried out to investigate the feasibility and to obtain general design data on such points as the design and properties of the gray and black control rods, the design of the beryllium reflector, gamma and neutron heating, the use of burnable poisons, etc. In performing these calculations, use was made of the IBM 650 PROD code obtained from KAPL. 3. With stages 1 and 2 carried out, two-dimensional calculations of the core at start-up conditions were performed on the Univac computer. 4. Detailed two-dimensional calculations of the properties of the ETR with a proposed first set of experiments in place were carried out. 5. A series of nuclear tests were performed at the reactivity measurements facility at the MTR site in order to confirm the validity of the analytical techniques in physics analysis. In performing the two-dimensional Univac calculations, the MUG code developed by KAPL and the Cuthill code developed at the David Taylor Model Basin were utilized. In

Investigation of natural circulation instability and transients in passively safe novel modular reactor

Science.gov (United States)

Shi, Shanbin

The Purdue Novel Modular Reactor (NMR) is a new type small modular reactor (SMR) that belongs to the design of boiling water reactor (BWR). Specifically, the NMR is one third the height and area of a conventional BWR reactor pressure vessel (RPV) with an electric output of 50 MWe. The fuel cycle length of the NMR-50 is extended up to 10 years due to optimized neutronics design. The NMR-50 is designed with double passive engineering safety system. However, natural circulation BWRs (NCBWR) could experience certain operational difficulties due to flow instabilities that occur at low pressure and low power conditions. Static instabilities (i.e. flow excursion (Ledinegg) instability and flow pattern transition instability) and dynamic instabilities (i.e. density wave instability and flashing/condensation instability) pose a significant challenge in two-phase natural circulation systems. In order to experimentally study the natural circulation flow instability, a proper scaling methodology is needed to build a reduced-size test facility. The scaling analysis of the NMR uses a three-level scaling method, which was developed and applied for the design of the Purdue Multi-dimensional Integral Test Assembly (PUMA). Scaling criteria is derived from dimensionless field equations and constitutive equations. The scaling process is validated by the RELAP5 analysis for both steady state and startup transients. A new well-scaled natural circulation test facility is designed and constructed based on the scaling analysis of the NMR-50. The experimental facility is installed with different equipment to measure various thermal-hydraulic parameters such as pressure, temperature, mass flow rate and void fraction. Characterization tests are performed before the startup transient tests and quasi-steady tests to determine the loop flow resistance. The controlling system and data acquisition system are programmed with LabVIEW to realize the real-time control and data storage. The thermal

Low-enrichment and long-life Scalable LIquid Metal cooled small Modular (SLIMM-1.2) reactor

Energy Technology Data Exchange (ETDEWEB)

El-Genk, Mohamed S., E-mail: mgenk@unm.edu [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM (United States); Nuclear Engineering Department, University of New Mexico, Albuquerque, NM (United States); Mechanical Engineering Department, University of New Mexico, Albuquerque, NM (United States); Chemical and Biological Engineering Department, University of New Mexico, Albuquerque, NM (United States); Palomino, Luis M.; Schriener, Timothy M. [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM (United States); Nuclear Engineering Department, University of New Mexico, Albuquerque, NM (United States)

2017-05-15

Highlights: • Developed low enrichment and natural circulation cooled SLIMM-1.2 SMR for generating 10–100 MW{sub th}. • Neutronics analyses estimate operation life and temperature reactivity feedback. • At 100 MW{sub th}, SLIMM-1.2 operates for 6.3 FPY without refueling. • SLIMM-1.2 has relatively low power peaking and maximum UN fuel temperature < 1400 K. - Abstract: The Scalable LIquid Metal cooled small Modular (SLIMM-1.0) reactor with uranium nitride fuel enrichment of 17.65% had been developed for generating 10–100 MW{sub th} continuously, without refueling for ∼66 and 5.9 full power years, respectively. Natural circulation of in-vessel liquid sodium (Na) cools the core of this fast energy spectrum reactor during nominal operation and after shutdown, with the aid of a tall chimney and an annular Na/Na heat exchanger (HEX) of concentric helically coiled tubes. The HEX at the top of the downcomer maximizes the static pressure head for natural circulation. In addition to the independent emergency shutdown (RSS) and reactor control (RC), the core negative temperature reactivity feedback safely decreases the reactor thermal power, following modest increases in the temperatures of UN fuel and in-vessel liquid sodium. The decay heat is removed from the core by natural circulation of in-vessel liquid sodium, with aid of the liquid metal heat pipes laid along the reactor vessel wall, and the passive backup cooling system (BCS) using natural circulation of ambient air along the outer surface of the guard vessel wall. This paper investigates modifying the SLIMM-1.0 reactor design to lower the UN fuel enrichment. To arrive at a final reactor design (SLIMM-1.2), the performed neutronics and reactivity depletion analyses examined the effects of various design and material choices on both the cold-clean and the hot-clean excess reactivity, the reactivity shutdown margin, the full power operation life at 100 MW{sub th}, the fissile production and depletion, the

Research Project 'RB research nuclear reactor' (operation and maintenance), Final report

International Nuclear Information System (INIS)

1985-01-01

This final report covers operation and maintenance activities at the RB reactor during period from 1981-1985. First part covers the RB reactor operation, detailed description of reactor components, fuel, heavy water, reactor vessel, cooling system, equipment and instrumentation, auxiliary systems. It contains data concerned with dosimetry and radiation protection, reactor staff, and financial data. Second part deals maintenance, regular control and testing of reactor equipment and instrumentation. Third part is devoted to basic experimental options and utilization of the RB reactor including training

Maintaining excellence: planning a new multi-purpose research reactor for Canada

International Nuclear Information System (INIS)

Whitlock, J.

2011-01-01

This paper outlines the need for a multi-purpose research reactor for Canada. The main objective of this paper is to stimulate a discussion and increase the profile for the need to develop a national strategy to meet the long term research reactor needs.

Small Modular Reactors for Countries with Small to Medium Electric Grids - An Economically Sensible Solution

International Nuclear Information System (INIS)

Young, P.

2012-01-01

There has recently been a renewed interest throughout the world in small nuclear units for generating electricity and for other applications. A report by the World Nuclear Association discussing the advantages of small modular nuclear reactors (SMRs) over traditional nuclear reactor designs, states that ''modern small reactors for power generation are expected to have greater simplicity of design, economy of mass production, and reduced siting costs. Many are also designed for a high level of passive or inherent safety in the event of malfunction.'' Since the inception of nuclear power, the size of reactor units has grown from under 100 MWe to more than 1600 MWe. Today, due partly to the high capital cost of large power reactors and partly to the need to service small electricity grids, there is a move to develop smaller units. These may be built individually or as modules in a larger plant. SMRs are a good fit in markets where anticipated electricity demand is projected to increase incrementally, because SMRs could be built in series as needed. SMRs might be particularly attractive in countries that currently rely on diesel generators for producing electricity. Small reactors could make economic sense because of the high cost of diesel generation compared to the low marginal cost of producing electricity from nuclear energy. (Keeping in mind the initial investment costs and the need to establish a national regulatory program.) Some SMR designs are fabricated in a factory and then delivered to the site. This could be a solution for markets that lack the qualified engineers and skilled craft workers needed to construct large reactors on site. This paper will provide an overview of the types and attributes of SMRs in use or under development worldwide, describe the similarities and important differences between designs, discuss potential applications for SMRs, including baseload electricity generation, electricity generation for remote locations and areas with

GT-MHR as economical highly efficient inherently safe modular gas cooled reactor for electric power generation

International Nuclear Information System (INIS)

Baydakov, A.G.; Kodochigov, N.G.; Kuzavkov, N.G.; Vorontsov, V.E.

2001-01-01

The Gas Turbine Modular Helium Reactor (GT-MHR) project is a joint effort of organisations in the Russian Federation co-ordinated by the Ministry on Atomic Energy (MINATOM), General Atomics in the United States, Framatome in France, and Fuji Electric in Japan. This paper describes the structure, objectives and history of the GT-MHR project, as well as the plant design, including plant layout, components and fuel. It summarises the results of evaluations of the design safety characteristics and economics, for both a plutonium fueled single module and a uranium fueled four module plant. The plans and anticipated structure for an international project to construct the plant are also summarised. (author)

Helium circulator design concepts for the modular high temperature gas-cooled reactor (MHTGR) plant

International Nuclear Information System (INIS)

McDonald, C.F.; Nichols, M.K.; Kaufman, J.S.

1988-01-01

Two helium circulators are featured in the Modular High-Temperature Gas-Cooled Reactor (MHTGR) power plant - (1) the main circulator, which facilitates the transfer of reactor thermal energy to the steam generator, and (2) a small shutdown cooling circulator that enables rapid cooling of the reactor system to be realized. The 3170 kW(e) main circulator has an axial flow compressor, the impeller being very similar to the unit in the Fort St. Vrain (FSV) plant. The 164 kW(e) shutdown cooling circulator, the design of which is controlled by depressurized conditions, has a radial flow compressor. Both machines are vertically oriented, have submerged electric motor drives, and embody rotors that are supported on active magnetic bearings. As outlined in this paper, both machines have been conservatively designed based on established practice. The circulators have features and characteristics that have evolved from actual plant operating experience. With a major goal of high reliability, emphasis has been placed on design simplicity, and both machines are readily accessible for inspection, repair, and replacement, if necessary. In this paper, conceptual design aspects of both machines are discussed, together with the significant technology bases. As appropriate for a plant that will see service well into the 21st century, new and emerging technologies have been factored into the design. Examples of this are the inclusion of active magnetic bearings, and an automated circulator condition monitoring system. (author). 18 refs, 20 figs, 13 tabs

Deep-Burn Modular Helium Reactor Fuel Development Plan

Energy Technology Data Exchange (ETDEWEB)

McEachern, D

2002-12-02

This document contains the workscope, schedule and cost for the technology development tasks needed to satisfy the fuel and fission product transport Design Data Needs (DDNs) for the Gas Turbine-Modular Helium Reactor (GT-MHR), operating in its role of transmuting transuranic (TRU) nuclides in spent fuel discharged from commercial light-water reactors (LWRs). In its application for transmutation, the GT-MHR is referred to as the Deep-Burn MHR (DB-MHR). This Fuel Development Plan (FDP) describes part of the overall program being undertaken by the U.S. Department of Energy (DOE), utilities, and industry to evaluate the use of the GT-MHR to transmute transuranic nuclides from spent nuclear fuel. The Fuel Development Plan (FDP) includes the work on fuel necessary to support the design and licensing of the DB-MHR. The FDP is organized into ten sections. Section 1 provides a summary of the most important features of the plan, including cost and schedule information. Section 2 describes the DB-MHR concept, the features of its fuel and the plan to develop coated particle fuel for transmutation. Section 3 describes the knowledge base for fabrication of coated particles, the experience with irradiation performance of coated particle fuels, the database for fission product transport in HTGR cores, and describes test data and calculations for the performance of coated particle fuel while in a repository. Section 4 presents the fuel performance requirements in terms of as-manufactured quality and performance of the fuel coatings under irradiation and accident conditions. These requirements are provisional because the design of the DB-MHR is in an early stage. However, the requirements are presented in this preliminary form to guide the initial work on the fuel development. Section 4 also presents limits on the irradiation conditions to which the coated particle fuel can be subjected for the core design. These limits are based on past irradiation experience. Section 5 describes

Reactor scale modeling of multi-walled carbon nanotube growth

International Nuclear Information System (INIS)

Lombardo, Jeffrey J.; Chiu, Wilson K.S.

2011-01-01

As the mechanisms of carbon nanotube (CNT) growth becomes known, it becomes important to understand how to implement this knowledge into reactor scale models to optimize CNT growth. In past work, we have reported fundamental mechanisms and competing deposition regimes that dictate single wall carbon nanotube growth. In this study, we will further explore the growth of carbon nanotubes with multiple walls. A tube flow chemical vapor deposition reactor is simulated using the commercial software package COMSOL, and considered the growth of single- and multi-walled carbon nanotubes. It was found that the limiting reaction processes for multi-walled carbon nanotubes change at different temperatures than the single walled carbon nanotubes and it was shown that the reactions directly governing CNT growth are a limiting process over certain parameters. This work shows that the optimum conditions for CNT growth are dependent on temperature, chemical concentration, and the number of nanotube walls. Optimal reactor conditions have been identified as defined by (1) a critical inlet methane concentration that results in hydrogen abstraction limited versus hydrocarbon adsorption limited reaction kinetic regime, and (2) activation energy of reaction for a given reactor temperature and inlet methane concentration. Successful optimization of a CNT growth processes requires taking all of those variables into account.

Contribution to the optimization of the coupling of nuclear reactors to desalination processes

International Nuclear Information System (INIS)

Dardour, S.

2007-04-01

This work deals with modelling, simulation and optimization of the coupling between nuclear reactors (PWR, modular high temperature reactors) and desalination processes (multiple effect distillation, reverse osmosis). The reactors considered in this study are PWR (Pressurized Water Reactor) and GTMHR (Gas Turbine Modular Helium Reactor). The desalination processes retained are MED (Multi Effect Distillation) and SWRO (Sea Water Reverse Osmosis). A software tool: EXCELEES of thermodynamic modelling of coupled systems, based on the Engineering Algebraic Equation Solver has been developed. Models of energy conversion systems and of membrane desalination processes and distillation have been developed. Based on the first and second principles of thermodynamics, these models have allowed to determine the optimal running point of the coupled systems. The thermodynamic analysis has been completed by a first economic evaluation. Based on the use of the DEEP software of the IAEA, this evaluation has confirmed the interest to use these types of reactors for desalination. A modelling tool of thermal processes of desalination in dynamic condition has been developed too. This tool has been applied to the study of the dynamics of an existing plant and has given satisfying results. A first safety checking has been at last carried out. The transients able to jeopardize the integrated system have been identified. Several measures aiming at consolidate the safety have been proposed. (O.M.)

Defining Modules, Modularity and Modularization

DEFF Research Database (Denmark)

Miller, Thomas Dedenroth; Pedersen, Per Erik Elgård

The paper describes the evolution of the concept of modularity in a historical perspective. The main reasons for modularity are: create variety, utilize similarities, and reduce complexity. The paper defines the terms: Module, modularity, and modularization.......The paper describes the evolution of the concept of modularity in a historical perspective. The main reasons for modularity are: create variety, utilize similarities, and reduce complexity. The paper defines the terms: Module, modularity, and modularization....

Scaled Facility Design Approach for Pool-Type Lead-Bismuth Eutectic Cooled Small Modular Reactor Utilizing Natural Circulation

Energy Technology Data Exchange (ETDEWEB)

Park, Sangrok; Shin, Yong-Hoon; Lee, Jueun; Hwang, Il Soon [Seoul National University, Seoul (Korea, Republic of)

2015-10-15

In low carbon era, nuclear energy is the most prominent energy source of electricity. For steady ecofriendly nuclear energy supply, Generation IV reactors which are future nuclear reactor require safety, sustainability, economics and non-proliferation as four criteria. Lead cooled fast reactor (LFR) is one of these reactor type and Generation IV international forum (GIF) adapted three reference LFR systems which are a small and movable systems with long life without refueling, intermediate size and huge electricity generation system for power grid. NUTRECK (Nuclear Transmutation Energy Center of Korea) has been designed reactor called URANUS (Ubiquitous, Rugged, Accident-forgiving, Non-proliferating, and Ultra-lasting Sustainer) which is small modular reactor and using lead-bismuth eutectic coolant. To prove natural circulation capability of URANUS and analyze design based accidents, scaling mock-up experiment facility will be constructed. In this paper, simple specifications of URANUS will be presented. Then based on this feature, scaling law and scaled facility design results are presented. To validate safety feature and thermodynamics characteristic of URANUS, scaled mockup facility of URANUS is designed based on the scaling law. This mockup adapts two area scale factors, core and lower parts of mock-up are scaled for 3D flow experiment. Upper parts are scaled different size to reduce electricity power and LBE tonnage. This hybrid scaling method could distort some thermal-hydraulic parameters, however, key parameters for experiment will be matched for up-scaling. Detailed design of mock-up will be determined through iteration for design optimization.

A multi-modular helical magnetic millirobot navigating in curved tubular environments

Directory of Open Access Journals (Sweden)

Seungmun Jeon

2017-05-01

Full Text Available This paper proposes a novel multi-modular helical magnetic millirobot (MHMM able to navigate in curved tubular environments. The proposed MHMM is basically composed of head, middle, and tail modules serially connected by a universal joint, and can employ more modules if necessary. The head module is a helical body with a magnetic rotary tip attached to the front end. Thus, once located in a curved tube filled with a fluid, the MHMM can effectively generate navigating or unclogging motions simply actuated by an external rotating magnetic field. This paper also examines the dynamic characteristics of the MHMM under various working conditions to manipulate the MHMM in a stable and safe manner. Then, it demonstrates various controlled motions of the MHMM in an in-vitro pulsatile flow environment to show the validity of the proposed structure and method.

Stability of test environments for performance evaluation of materials for the modular high-temperature gas-cooled reactor

International Nuclear Information System (INIS)

Edgemon, G.L.; Wilson, D.F.; Bell, G.E.C.

1993-01-01

Stability of the primary helium-based coolant test gas for use in performance ests of materials for the Modular High-Temperature Gas-Cooled Reactor (MHTGR) was determined. Results of tests of the initial gas chemistry from General Atomics (GA) at elevated temperatures, and the associated results predicted by the SOLGASMIX trademark modelling package are presented. Results indicated that for this gas composition and at flow rates obtainable in the test loop, 466 ± 24C is the highest temperature that can be maintained without significantly altering the specified gas chemistry. Four additional gas chemistries were modelled using SOLGASMIX trademark

Simulation of thermal response of the 250 MWT modular HTGR during hypothetical uncontrolled heatup accidents

International Nuclear Information System (INIS)

Harrington, R.M.; Ball, S.J.

1985-01-01

One of the central design features of the 250 MWT modular HTGR is the ability to withstand uncontrolled heatup accidents without severe consequences. This paper describes calculational studies, conducted to test this design feature. A multi-node thermal-hydraulic model of the 250 MWT modular HTGR reactor core was developed and implemented in the IBM CSMP (Continuous System Modeling Program) simulation language. Survey calculations show that the loss of forced circulation accident with loss of steam generator cooling water and with accidental depressurization is the most severe heatup accident. The peak hot-spot fuel temperature is in the neighborhood of 1600 0 C. Fuel failure and fission product releases for such accidents would be minor. Sensitivity studies show that code input assumptions for thermal properties such as the side reflector conductivity have a significant effect on the peak temperature. A computer model of the reactor vessel cavity concrete wall and its surrounding earth was developed to simulate the extremely unlikely and very slowly-developing heatup accident that would take place if the worst-case loss of forced primary coolant circulation accident were further compounded by the loss of cooling water to the reactor vessel cavity liner cooling system. Results show that the ability of the earth surrounding the cavity to act as a satisfactory long-term heat sink is very sensitive to the assumed rate of decay heat generation and on the effective thermal conductivity of the earth

Theoretical Analysis of the Relationships Between Modularity in Design and Modularity in Production

DEFF Research Database (Denmark)

Kubota, Flávio Issao; Hsuan, Juliana; Cauchick-Miguel, Paulo Augusto

2017-01-01

This paper investigates the relationships between modularity in design (MID) and modularity in production (MIP) in the automotive industry in terms of how automotive companies obtain benefits and/or drawbacks through MID/MIP relationships. A literature analysis was conducted in order to identify...... the possible relationships between MID and MIP as well as the concepts behind these connections. Sixty-one papers were identified to portray relationships between modular product architecture and modular production systems. Results show a representation of MID and MIP relationships by illustrating that many...... propositions are offered for future field research. Finally, relationships between MID and MIP might be connected with modularity’s maturity level in companies. This is a literature review paper; therefore, empirical evidence is needed to further support current findings. Future studies could analyze...

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-07-01

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

Evaluation of the applicability of existing nuclear power plant regulatory requirements in the U.S. to advanced small modular reactors.

Energy Technology Data Exchange (ETDEWEB)

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

2013-05-01

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

Modular representation and analysis of fault trees

Energy Technology Data Exchange (ETDEWEB)

Olmos, J; Wolf, L [Massachusetts Inst. of Tech., Cambridge (USA). Dept. of Nuclear Engineering

1978-08-01

An analytical method to describe fault tree diagrams in terms of their modular compositions is developed. Fault tree structures are characterized by recursively relating the top tree event to all its basic component inputs through a set of equations defining each of the modulus for the fault tree. It is shown that such a modular description is an extremely valuable tool for making a quantitative analysis of fault trees. The modularization methodology has been implemented into the PL-MOD computer code, written in PL/1 language, which is capable of modularizing fault trees containing replicated components and replicated modular gates. PL-MOD in addition can handle mutually exclusive inputs and explicit higher order symmetric (k-out-of-n) gates. The step-by-step modularization of fault trees performed by PL-MOD is demonstrated and it is shown how this procedure is only made possible through an extensive use of the list processing tools available in PL/1. A number of nuclear reactor safety system fault trees were analyzed. PL-MOD performed the modularization and evaluation of the modular occurrence probabilities and Vesely-Fussell importance measures for these systems very efficiently. In particular its execution time for the modularization of a PWR High Pressure Injection System reduced fault tree was 25 times faster than that necessary to generate its equivalent minimal cut-set description using MOCUS, a code considered to be fast by present standards.

Advanced Small Modular Reactor Economics Model Development

Energy Technology Data Exchange (ETDEWEB)

Harrison, Thomas J [ORNL

2014-10-01

The US Department of Energy Office of Nuclear Energy’s Advanced Small Modular Reactor (SMR) research and development activities focus on four key areas: Developing assessment methods for evaluating advanced SMR technologies and characteristics; and Developing and testing of materials, fuels and fabrication techniques; and Resolving key regulatory issues identified by US Nuclear Regulatory Commission and industry; and Developing advanced instrumentation and controls and human-machine interfaces. This report focuses on development of assessment methods to evaluate advanced SMR technologies and characteristics. Specifically, this report describes the expansion and application of the economic modeling effort at Oak Ridge National Laboratory. Analysis of the current modeling methods shows that one of the primary concerns for the modeling effort is the handling of uncertainty in cost estimates. Monte Carlo–based methods are commonly used to handle uncertainty, especially when implemented by a stand-alone script within a program such as Python or MATLAB. However, a script-based model requires each potential user to have access to a compiler and an executable capable of handling the script. Making the model accessible to multiple independent analysts is best accomplished by implementing the model in a common computing tool such as Microsoft Excel. Excel is readily available and accessible to most system analysts, but it is not designed for straightforward implementation of a Monte Carlo–based method. Using a Monte Carlo algorithm requires in-spreadsheet scripting and statistical analyses or the use of add-ons such as Crystal Ball. An alternative method uses propagation of error calculations in the existing Excel-based system to estimate system cost uncertainty. This method has the advantage of using Microsoft Excel as is, but it requires the use of simplifying assumptions. These assumptions do not necessarily bring into question the analytical results. In fact, the

Modular development mat reinforcing and liners containment/annulus building. Final report supplement. Volume I

International Nuclear Information System (INIS)

Wainrib, J.

1978-06-01

The feasibility of developing a modular system approach for the construction of containment mats has been established. The investigative results obtained from a plant used for the modular study has determined that (1) modularization results in a 10-week reduction in the construction schedule critical path over that for the conventional approach, and (2) modularization is cost effective and could result in a savings of approximately $10,700,000. Accordingly, it is recommended that the modular system approach be adopted as standard practice for the construction of containment mats

Modularized construction, structural design and analysis of CANDU 3 plant

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-31

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

Modularized construction, structural design and analysis of CANDU 3 plant

International Nuclear Information System (INIS)

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

1995-01-01

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

Advanced multi-physics simulation capability for very high temperature reactors

International Nuclear Information System (INIS)

Lee, Hyun Chul; Tak, Nam Il; Jo Chang Keun; Noh, Jae Man; Cho, Bong Hyun; Cho, Jin Woung; Hong, Ser Gi

2012-01-01

The purpose of this research is to develop methodologies and computer code for high-fidelity multi-physics analysis of very high temperature gas-cooled reactors(VHTRs). The research project was performed through Korea-US I-NERI program. The main research topic was development of methodologies for high-fidelity 3-D whole core transport calculation, development of DeCART code for VHTR reactor physics analysis, generation of VHTR specific 190-group cross-section library for DeCART code, development of DeCART/CORONA coupled code system for neutronics/thermo-fluid multi-physics analysis, and benchmark analysis against various benchmark problems derived from PMR200 reactor. The methodologies and the code systems will be utilized a key technologies in the Nuclear Hydrogen Development and Demonstration program. Export of code system is expected in the near future and the code systems developed in this project are expected to contribute to development and export of nuclear hydrogen production system

Preliminary Design of Supercritical CO{sub 2} Radial Turbine for Micro Modular Reactor

Energy Technology Data Exchange (ETDEWEB)

Cho, Seong Kuk; Lee, Jekyoung; Kim, Seong Gu; Lee, Jeong Ik [KAIST (Korea, Republic of); Cha, Jae Eun [KAERI, Daejeon (Korea, Republic of)

2016-05-15

The KAIST research team suggested an innovative concept of SMR called KAIST Micro Modular Reactor (MMR). It is aimed for achieving complete modularization of a nuclear power plant including the Power Conversion Unit (PCU) for simple transportation and installation. In consideration of the maintenance, S-CO{sub 2} has superior characteristic to the light water because it is less corrosive. Table I represents summary of the main design results of KAIST MMR. As such efforts the study on S-CO{sub 2} turbomachinery has been actively conducted because the S-CO{sub 2} Brayton cycle operates under extreme conditions such as high rotational speed, high pressure, and dramatic change of thermodynamic properties near the critical point (30.98 .deg. C, 7377kPa). To achieve higher thermal efficiency, the operation near the critical point is prerequisite in the S-CO{sub 2} Brayton cycle. For these reasons, research works on the S-CO{sub 2} turbomachinery naturally have been focused on the compressor which operates near the critical point. Due to the absence of loss models in S-CO{sub 2} field, loss models with air turbine were first utilized to design and predict the off design performance of a S-CO{sub 2} turbine. To check propriety of the code logic the code results compared with experimental data measured in air condition. The predicted values showed good agreement with the experiment data. Lastly, performance maps for S-CO{sub 2} turbine for the MMR were generated with change of mass flow rate and rotational speed.

The Use of Multi-Reactor Cascade Plasma Electrolysis for Linear Alkylbenzene Sulfonate Degradation

Science.gov (United States)

Saksono, Nelson; Ibrahim; Zainah; Budikania, Trisutanti

2018-03-01

Plasma electrolysis is a method that can produce large amounts of hydroxyl radicals to degrade organic waste. The purpose of this study is to improve the effectiveness of Linear alkylbenzene sulfonate (LAS) degradation by using multi-reactor cascade plasma electrolysis. The reactor which operated in circulation system, using 3 reactors series flow and 6 L of LAS with initial concentration of 100 ppm. The results show that the LAS degradation can be improved multi-reactor cascade plasma electrolysis. The greatest LAS degradation is achieved up to 81.91% with energy consumption of 2227.34 kJ/mmol that is obtained during 120 minutes by using 600 Volt, 0.03 M of KOH, and 0.5 cm of the anode depth.

DESAIN TERAS PLTN JENIS PEBBLE BED MODULAR REACTOR (PBMR MENGGUNAKAN PAKET PROGRAM MCNP-5 PADA KONDISI BEGINNING OF LIFE

Directory of Open Access Journals (Sweden)

Ralind Re Marla

2015-03-01

Full Text Available Telah dilakukan desain teras Pembangkit Listrik Tenaga Nuklir (PLTN untuk jenis Pebble Bed Modular Reactor (PBMR dengan daya 70 MWe untuk keperluan proses smelter pada keadaan beginning of life (BOL. Analisis ini bertujuan untuk mengetahui persen pengkayaan, distribusi suhu dan nilai keselamatan dengan koefisien reaktivitas teras yang negatif pada reaktor jenis PBMR apabila daya reaktor 70 MWe. Analisis menggunakan program Monte Carlo N-Particle-5 (MCNP5 dan dari hasil analisis ini diharapkan dapat memenuhi syarat dalam mendukung program percepatan pembangunan kelistrikan batubara 10.000 MWe khususnya untuk proses smelter, yang tersebar merata di wilayah Indonesia. Hasil penelitian menunjukkan bahwa, faktor perlipatan efektif (k-eff Reaktor jenis PBMR daya 70 MWe mengalami kondisi kritis pada pengkayaan 5,626 % dengan nilai faktor perlipatan efektif 1,00031±0,00087 dan nilai koefisien reaktivitas suhu pada -10,0006 pcm/K. Dari hasil analisis daat disimpulkan bahwa reaktor jenis PBMR daya 70 MWe adalah aman.   ABSTRACT The core design of Nuclear Power Plant for Pebble Bed Modular Reactor (PBMR type with 70 MWe capacity power in Beginning of Life (BOL has been performed. The aim of this analysis, to know percent enrichment, temperature distribution and safety value by negative temperature coefficient at type PBMR if reactor power become lower equal to 70 MWe. This analysis was expected become one part of overview project development the power plant with 10.000 MWe of total capacity, spread evenly in territory of Indonesia especially to support of smelter industries. The results showed that, effective multiplication factor (keff with power 70 MWe critical condition at enrichment 5,626 %is 1,00031±0,00087, based on enrichment result, a value of the temperature coefficient reactivity is - 10,0006 pcm/K. Based on the results of these studies, it can beconcluded that the PBMR 70 MWe design is theoritically safe.

Small modular reactors (SMRs) - the way forward for the nuclear industry in Canada?

International Nuclear Information System (INIS)

Sam-Aggrey, H.

2014-01-01

Small Modular Reactors (SMRs) are being touted as safer, more cost effective and more flexible than traditional nuclear power plants (NPPs). Consequently, it has been argued that SMR technology is pivotal to the revitalization of the nuclear industry at the national and global levels. Drawing mainly on previously published literature, this paper explores the suitability of SMRs for various niche market applications in Canada. The paper examines the potential role of SMRs in providing an opportunity for remote mines and communities in northern Canada to reduce their vulnerability and dependence on costly, high-carbon diesel fuel. Other niche market applications of SMRs explored include: SMRs deployment in Saskatchewan for grid augmentation and as replacement options for Saskatchewan's ageing coal plants; the use of SMRs for bitumen extraction in the Oil Sands, and the potential use of SMRs in Canadian-owned foreign based mines. The socio-economic benefits of SMR deployments are also discussed. Building an SMR industry in Canada could complement the country's extensive expertise in uranium mining, reactor technology, plant operation, nuclear research, and environmental and safety standards, thereby enhancing Canada's ability to offer services throughout the entire nuclear life cycle. The paper also outlines some of the technical, economic and social barriers that could impede the successful introduction of SMRs in Canada. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-02-23

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Safety analysis of the experimental multi-purpose high-temperature gas-cooled reactor

International Nuclear Information System (INIS)

Mitake, Susumu; Ezaki, Masahiro; Suzuki, Katsuo; Takaya, Junichi; Shimazu, Akira

1976-02-01

Safety features of the experimental multi-purpose high-temperature gas-cooled reactor being developed in JAERI were studied or the basis of its preliminary conceptual design of the reactor plant. Covered are control of the plant in transients, plant behaviour in accidents, and functions of engineered safeguards, and also dynamics of the uprant and frequencies of the accidents. These studies have shown, (i) the reactor plant can be operated both in plant slave to reactor and reactor slave to plant control, (ii) stable control of

The use of concrete-filled steel structures for modular construction of advanced reactors

International Nuclear Information System (INIS)

Braverman, J.; Morante, R.; Hofmayer, C.; Graves, H.

1997-01-01

Modular construction techniques have been successfully used in a number of industries, both domestically and internationally. Recently, the use of structural modules has been proposed for advanced nuclear power plants. This paper presents the results of a research program which evaluated the use of modular construction for safety-related structures in advanced nuclear power plant designs. The research program included review of current modular construction technology, development of licensing review criteria for modular construction, and initial validation of currently available analytical techniques applied to concrete-filled steel structural modules

Market potential of heat utilization of modular HTR in Japan

International Nuclear Information System (INIS)

Ide, Akira; Tasaka, Kanji.

1993-01-01

HTR is considered to be the most suitable reactor type to use in the field other than power generation. So it is useful to know market potential of this type of reactor in Japan to justify its development. This potential was estimated to be about 400 200MWt modular HTR reactors. This number will be double if the market of hydrogen is developed. (J.P.N.)

Some Issues of the Paradigm of Multi-learning Machine - Modular Neural Networks

DEFF Research Database (Denmark)

Wang, Pan; Feng, Shuai; Fan, Zhun

2009-01-01

This paper addresses some issues on the weighted linear integration of modular neural networks (MNN: a paradigm of hybrid multi-learning machines). First, from the general meaning of variable weights and variable elements synthesis, three basic kinds of integrated models are discussed...... a general form while the corresponding computational algorithms are described briefly. The authors present a new training algorithm of sub-networks named “'Expert in one thing and good at many' (EOGM).” In this algorithm, every sub-network is trained on a primary dataset with some of its near neighbors...... as the accessorial datasets. Simulated results with a kind of dynamic integration methods show the effectiveness of these algorithms, where the performance of the algorithm with EOGM is better than that of the algorithm with a common training method....

Final Report: Self Consolidating Concrete Construction for Modular Units

Energy Technology Data Exchange (ETDEWEB)

Gentry, Russell [Georgia Inst. of Technology, Atlanta, GA (United States); Kahn, Lawrence [Georgia Inst. of Technology, Atlanta, GA (United States); Kurtis, Kimberly [Georgia Inst. of Technology, Atlanta, GA (United States); Petrovic, Bojan [Georgia Inst. of Technology, Atlanta, GA (United States); Loreto, Giovanni [Georgia Inst. of Technology, Atlanta, GA (United States); Van Wyk, Jurie [Westinghouse Electric Company, Cranberry Township, PA (United States); Canterero-Leal, Carlos [Westinghouse Electric Company, Cranberry Township, PA (United States)

2016-07-29

This report outlines the development of a self-consolidating concrete (also termed “self-compacting concrete” or SCC) so that concrete placement can be made into steel plate composite (SC) modular structures without the need for continuous concrete placement. As part of the research, SCC mixtures were developed and validated to ensure sufficient shear capacity across cold-joints, while minimizing shrinkage and temperature increase during curing to enhance concrete bonding with the steel plate construction found in modular units. The self-roughening concrete produced as part of this research was assessed in SC structures at three scales: small-scale shear-friction specimens, mid-scale beams tested in in-plane and out-of-plane bending, and a full-scale validation test using an SC module produced by Westinghouse as part of the Plant Vogtle expansion. The experiments show that the self-roughening concrete can produce a cold-joint surface of 0.25 inches (6 mm) without external vibration during concrete placement. The experiments and subsequent analysis show that the shear friction provisions of ACI 318-14, Section 22.9 can be used to assess the shear capacity of the cold-joints in SC modular construction, and that friction coefficient of 1.35 is appropriate for use with these provisions.

Modular construction: 30 years of experience in the naval sector, useful for nuclear power plants

International Nuclear Information System (INIS)

Lepelletier, P.; Danguy des Deserts, S.

2014-01-01

This article reviews the benefits of modular construction in the sector of submarines that was implemented in the mid eighties in French shipyards. The design of the submarine is cut in different longitudinal parts so that each part can be built and completely equipped independently from the other parts. The different parts are then joined and welded at the end of the process of fabrication to form the submarine. The main advantages of such a construction are shorter construction times, standardized construction and a higher quality standard. DCNS proposes to use modular construction for the design of Flexblue which is an immersed small modular reactor (SMR). More generally, modular construction will add economic competitiveness to SMR reactors that will be necessary to their full development on a worldwide scale

Compact modular BWR (CM-BWR)

International Nuclear Information System (INIS)

Fennern, Larry; Boardman, Charles; Carroll, Douglas G.; Hida, Takahiko

2003-01-01

A preliminary assessment has shown that a small 350 MWe BWR reactor can be placed within a close fitting steel containment vessel that is 7.1 meters inside diameter. This allows the technology and manufacturing capability currently used to fabricate large ABWR reactor vessels to be used to provide a factory fabricated containment vessel for a 350 MWe BWR. When a close fitted steel containment is combined with a passive closed loop isolation condenser system and a natural circulating reactor system that contains a large water inventory, primary system leaks cannot uncover the core. This eliminates many of the safety systems needed in response to a LOCA that are common to large, conventional plant designs including. Emergency Core Flooding, Automatic Depressurization System, Active Residual Heat Removal, Safety Related Auxiliary Cooling, Safety Related Diesel Generators, Hydrogen Re-Combiners, Ex-vessel Core Retention and Cooling. By fabricating the containment in a factory and eliminating most of the conventional safety systems, the construction schedule is shortened and the capital cost reduced to levels that would not otherwise be possible for a relatively small modular BWR. This makes the CM-BWR a candidate for applications where smaller incremental power additions are desired relative to a large ALWR or where the local infrastructure is not able to accommodate a conventional ALWR plant rated at 1350 MWe or more. This paper presents a preliminary design description of a Compact Modular BWR (CM-BWR) whose design features dramatically reduce the size and cost of the reactor building and associated safety systems. (author)

Modular Universal Scalable Ion-trap Quantum Computer

Science.gov (United States)

2016-06-02

SECURITY CLASSIFICATION OF: The main goal of the original MUSIQC proposal was to construct and demonstrate a modular and universally- expandable ion...Distribution Unlimited UU UU UU UU 02-06-2016 1-Aug-2010 31-Jan-2016 Final Report: Modular Universal Scalable Ion-trap Quantum Computer The views...P.O. Box 12211 Research Triangle Park, NC 27709-2211 Ion trap quantum computation, scalable modular architectures REPORT DOCUMENTATION PAGE 11

Study of core flow distribution for small modular natural circulation lead or lead-alloy cooled fast reactors

International Nuclear Information System (INIS)

Chen, Zhao; Zhao, Pengcheng; Zhou, Guangming; Chen, Hongli

2014-01-01

Highlights: • A core flow distribution calculation code for natural circulation LFRs was developed. • The comparison study between the channel method and the CFD method was conducted. • The core flow distribution analysis and optimization design for a 10MW natural circulation LFR was conducted. - Abstract: Small modular natural circulation lead or lead-alloy cooled fast reactor (LFR) is a potential candidate for LFR development. It has many attractive advantages such as reduced capital costs and inherent safety. The core flow distribution calculation is an important issue for nuclear reactor design, which will provide important input parameters to thermal-hydraulic analysis and safety analysis. The core flow distribution calculation of a natural circulation LFR is different from that of a forced circulation reactor. In a forced circulation reactor, the core flow distribution can be controlled and adjusted by the pump power and the flow distributor, while in a natural circulation reactor, the core flow distribution is automatically adjusted according to the relationship between the local power and the local resistance feature. In this paper, a non-uniform heated parallel channel flow distribution calculation code was developed and the comparison study between the channel method and the CFD method was carried out to assess the exactness of the developed code. The core flow distribution analysis and optimization design for a 10MW natural circulation LFR was conducted using the developed code. A core flow distribution optimization design scheme for a 10MW natural circulation LFR was proposed according to the optimization analysis results

Improving Battery Reactor Core Design Using Optimization Method

International Nuclear Information System (INIS)

Son, Hyung M.; Suh, Kune Y.

2011-01-01

The Battery Omnibus Reactor Integral System (BORIS) is a small modular fast reactor being designed at Seoul National University to satisfy various energy demands, to maintain inherent safety by liquid-metal coolant lead for natural circulation heat transport, and to improve power conversion efficiency with the Modular Optimal Balance Integral System (MOBIS) using the supercritical carbon dioxide as working fluid. This study is focused on developing the Neutronics Optimized Reactor Analysis (NORA) method that can quickly generate conceptual design of a battery reactor core by means of first principle calculations, which is part of the optimization process for reactor assembly design of BORIS

Modular Multi-level converter based HVDC System for Grid Connection of Offshore Wind Power Plant

DEFF Research Database (Denmark)

Gnanarathna, U.N.; Chaudhary, Sanjay Kumar; Gole, A.M.

2010-01-01

This paper explores the application of modular multi-level converters (MMC) as a means for harnessing the power from off-shore wind power plants. The MMC consists of a large number of simple voltage sourced converter (VSC) submodules that can be easily assembled into a converter for high......-voltage and high power. The paper shows that the MMC converter has a fast response and low harmonic content in comparison with a two-level VSC option. The paper discusses the modeling approach used, including a solution to the modeling challenge imposed by the very large number of switching devices in the MMC....

Topological string theory, modularity and non-perturbative physics

Energy Technology Data Exchange (ETDEWEB)

Rauch, Marco

2011-09-15

In this thesis the holomorphic anomaly of correlators in topological string theory, matrix models and supersymmetric gauge theories is investigated. In the first part it is shown how the techniques of direct integration known from topological string theory can be used to solve the closed amplitudes of Hermitian multi-cut matrix models with polynomial potentials. In the case of the cubic matrix model, explicit expressions for the ring of non-holomorphic modular forms that are needed to express all closed matrix model amplitudes are given. This allows to integrate the holomorphic anomaly equation up to holomorphic modular terms that are fixed by the gap condition up to genus four. There is an one-dimensional submanifold of the moduli space in which the spectral curve becomes the Seiberg-Witten curve and the ring reduces to the non-holomorphic modular ring of the group {gamma}(2). On that submanifold, the gap conditions completely fix the holomorphic ambiguity and the model can be solved explicitly to very high genus. Using these results it is possible to make precision tests of the connection between the large order behavior of the 1/N expansion and non-perturbative effects due to instantons. Finally, it is argued that a full understanding of the large genus asymptotics in the multi-cut case requires a new class of non-perturbative sectors in the matrix model. In the second part a holomorphic anomaly equation for the modified elliptic genus of two M5-branes wrapping a rigid divisor inside a Calabi-Yau manifold is derived using wall-crossing formulae and the theory of mock modular forms. The anomaly originates from restoring modularity of an indefinite theta-function capturing the wall-crossing of BPS invariants associated to D4- D2-D0 brane systems. The compatibility of this equation with anomaly equations previously observed in the context of N=4 topological Yang-Mills theory on P{sup 2} and E-strings obtained from wrapping M5-branes on a del Pezzo surface which in

Topological string theory, modularity and non-perturbative physics

International Nuclear Information System (INIS)

Rauch, Marco

2011-09-01

In this thesis the holomorphic anomaly of correlators in topological string theory, matrix models and supersymmetric gauge theories is investigated. In the first part it is shown how the techniques of direct integration known from topological string theory can be used to solve the closed amplitudes of Hermitian multi-cut matrix models with polynomial potentials. In the case of the cubic matrix model, explicit expressions for the ring of non-holomorphic modular forms that are needed to express all closed matrix model amplitudes are given. This allows to integrate the holomorphic anomaly equation up to holomorphic modular terms that are fixed by the gap condition up to genus four. There is an one-dimensional submanifold of the moduli space in which the spectral curve becomes the Seiberg-Witten curve and the ring reduces to the non-holomorphic modular ring of the group Γ(2). On that submanifold, the gap conditions completely fix the holomorphic ambiguity and the model can be solved explicitly to very high genus. Using these results it is possible to make precision tests of the connection between the large order behavior of the 1/N expansion and non-perturbative effects due to instantons. Finally, it is argued that a full understanding of the large genus asymptotics in the multi-cut case requires a new class of non-perturbative sectors in the matrix model. In the second part a holomorphic anomaly equation for the modified elliptic genus of two M5-branes wrapping a rigid divisor inside a Calabi-Yau manifold is derived using wall-crossing formulae and the theory of mock modular forms. The anomaly originates from restoring modularity of an indefinite theta-function capturing the wall-crossing of BPS invariants associated to D4- D2-D0 brane systems. The compatibility of this equation with anomaly equations previously observed in the context of N=4 topological Yang-Mills theory on P 2 and E-strings obtained from wrapping M5-branes on a del Pezzo surface which in turn is

Optimization of TRU burnup in modular helium reactor

International Nuclear Information System (INIS)

Yonghee, Kim; Venneri, F.

2007-01-01

An optimization study of a single-pass TRU (transuranic) deep-burn (DB) has been performed for a block-type MHR (Modular Helium Reactor) proposed by General Atomics. Assuming a future equilibrium scenario of advanced LWRs, a high-burnup TRU vector is considered: 50 GWD/MTU and 5-year cooling. For 3-D equilibrium cores, the performance analysis is done by using a continuous energy Monte Carlo depletion code MCCARD. The core optimization is performed from the viewpoints of the core configuration, fuel management, TRISO fuel specification, and neutron spectrum. With regard to core configuration, two annular cores are investigated in terms of the neutron economy. A conventional radial shuffling scheme of fuel blocks is compared with an axial block shuffling strategy in terms of the fuel burnup and core power distributions. The impact of the kernel size of TRISO fuel is evaluated and a diluted kernel, instead of a conventional concentrated kernel, is introduced to maximize the TRU burnup by reducing the self-shielding effects of TRISO fuels. A higher graphite density is evaluated in terms of the fuel burnup. In addition, it is shown that the core power distribution can be effectively controlled by zoning of the packing fraction of TRISO fuels. We also have shown that a long-cycle DB-MHR core can be designed by using a small batch size for fuel reloading, at the expense of a marginal decrease of the TRU discharge burnup. Depending on the fuel management scheme, fuel specifications, and core parameters, the TRU burnup in an optimized DB-MHR core is over 60% in a single-pass irradiation campaign. (authors)

Multi-objective optimization of the reactor coolant system

International Nuclear Information System (INIS)

Chen Lei; Yan Changqi; Wang Jianjun

2014-01-01

Background: Weight and size are important criteria in evaluating the performance of a nuclear power plant. It is of great theoretical value and engineering significance to reduce the weight and volume of the components for a nuclear power plant by the optimization methodology. Purpose: In order to provide a new method for the optimization of nuclear power plant multi-objective, the concept of the non-dominated solution was introduced. Methods: Based on the parameters of Qinshan I nuclear power plant, the mathematical models of the reactor core, the reactor vessel, the main pipe, the pressurizer and the steam generator were built and verified. The sensitivity analyses were carried out to study the influences of the design variables on the objectives. A modified non-dominated sorting genetic algorithm was proposed and employed to optimize the weight and the volume of the reactor coolant system. Results: The results show that the component mathematical models are reliable, the modified non-dominated sorting generic algorithm is effective, and the reactor inlet temperature is the most important variable which influences the distribution of the non-dominated solutions. Conclusion: The optimization results could provide a reference to the design of such reactor coolant system. (authors)

A Multi-Physics simulation of the Reactor Core using CUPID/MASTER

International Nuclear Information System (INIS)

Lee, Jae Ryong; Cho, Hyoung Kyu; Yoon, Han Young; Cho, Jin Young; Jeong, Jae Jun

2011-01-01

KAERI has been developing a component-scale thermal hydraulics code, CUPID. The aim of the code is for multi-dimensional, multi-physics and multi-scale thermal hydraulics analysis. In our previous papers, the CUPID code has proved to be able to reproduce multidimensional thermal hydraulic analysis by validated with various conceptual problems and experimental data. For the numerical closure, it adopts a three dimensional, transient, two-phase and three-field model, and includes physical models and correlations of the interfacial mass, momentum, and energy transfer. For the multi-scale analysis, the CUPID is on progress to merge into system-scale thermal hydraulic code, MARS. In the present paper, a multi-physics simulation was performed by coupling the CUPID with three dimensional neutron kinetics code, MASTER. The MASTER is merged into the CUPID as a dynamic link library (DLL). The APR1400 reactor core during control rod drop/ejection accident was simulated as an example by adopting a porous media approach to employ fuel assembly. The following sections present the numerical modeling for the reactor core, coupling of the kinetics code, and the simulation results

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

Spatial atomic layer deposition on flexible substrates using a modular rotating cylinder reactor

International Nuclear Information System (INIS)

Sharma, Kashish; Hall, Robert A.; George, Steven M.

2015-01-01

Spatial atomic layer deposition (ALD) is a new version of ALD based on the separation of reactant gases in space instead of time. In this paper, the authors present results for spatial ALD on flexible substrates using a modular rotating cylinder reactor. The design for this reactor is based on two concentric cylinders. The outer cylinder remains fixed and contains a series of slits. These slits can accept a wide range of modules that attach from the outside. The modules can easily move between the various slit positions and perform precursor dosing, purging, or pumping. The inner cylinder rotates with the flexible substrate and passes underneath the various spatially separated slits in the outer cylinder. Trimethyl aluminum and ozone were used to grow Al 2 O 3 ALD films at 40 °C on metallized polyethylene terephthalate (PET) substrates to characterize this spatial ALD reactor. Spectroscopic ellipsometry measurements revealed a constant Al 2 O 3 ALD growth rate of 1.03 Å/cycle with rotation speeds from 40 to 100 RPM with the outer cylinder configured for one Al 2 O 3 ALD cycle per rotation. The Al 2 O 3 ALD growth rate then decreased at higher rotation rates for reactant residence times < 5 ms. The Al 2 O 3 ALD films were also uniform to within <1% across the central portion of metallized PET substrate. Fixed deposition time experiments revealed that Al 2 O 3 ALD films could be deposited at 2.08 Å/s at higher rotation speeds of 175 RPM. Even faster deposition rates are possible by adding more modules for additional Al 2 O 3 ALD cycles for every one rotation of the inner cylinder

[Modular enteral nutrition in pediatrics].

Science.gov (United States)

Murillo Sanchís, S; Prenafeta Ferré, M T; Sempere Luque, M D

1991-01-01

Modular Enteral Nutrition may be a substitute for Parenteral Nutrition in children with different pathologies. Study of 4 children with different pathologies selected from a group of 40 admitted to the Maternal-Childrens Hospital "Valle de Hebrón" in Barcelona, who received modular enteral nutrition. They were monitored on a daily basis by the Dietician Service. Modular enteral nutrition consists of modules of proteins, peptides, lipids, glucids and mineral salts-vitamins. 1.--Craneo-encephalic traumatisms with loss of consciousness, Feeding with a combination of parenteral nutrition and modular enteral nutrition for 7 days. In view of the tolerance and good results of the modular enteral nutrition, the parenteral nutrition was suspended and modular enteral nutrition alone used up to a total of 43 days. 2.--55% burns with 36 days of hyperproteic modular enteral nutrition together with normal feeding. A more rapid recovery was achieved with an increase in total proteins and albumin. 3.--Persistent diarrhoea with 31 days of modular enteral nutrition, 5 days on parenteral nutrition alone and 8 days on combined parenteral nutrition and modular enteral nutrition. In view of the tolerance and good results of the modular enteral nutrition, the parenteral nutrition was suspended. 4.--Mucoviscidosis with a total of 19 days on modular enteral nutrition, 12 of which were exclusively on modular enteral nutrition and 7 as a night supplement to normal feeding. We administered proteic intakes of up to 20% of the total calorific intake and in concentrations of up to 1.2 calories/ml of the final preparation, always with a good tolerance. Modular enteral nutrition can and should be used as a substitute for parenteral nutrition in children with different pathologies, thus preventing the complications inherent in parenteral nutrition.

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

International Nuclear Information System (INIS)

Susilowati, E.

2015-01-01

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

Modular robotic applications in nuclear power plant maintenance

International Nuclear Information System (INIS)

Glass, S.W.; Ranson, C.C.; Reinholtz, C.F.; Calkins, J.M.

1996-01-01

General-purpose factory automation robots have experienced limited use in nuclear maintenance and hazardous-environment work spaces due to demanding requirements on size, weight, mobility and adaptability. Robotic systems in nuclear power plants are frequently custom designed to meet specific space and performance requirements. Examples of these custom configurations include Framatome Technologies COBRA trademark Steam Generator Manipulator and URSULA trademark Reactor Vessel Inspection Manipulator. The use of custom robots in nuclear plants has been limited because of the lead time and expense associated with custom design. Developments in modular robotics and advanced robot control software coupled with more powerful low-cost computers, however, are helping to reduce the cost and schedule for deploying custom robots. A modular robotic system allows custom robot configurations to be implemented using standard (modular) joints and adaptable controllers. This paper discusses Framatome Technologies (FTI) current and planned developments in the area of modular robot system design

Technical Needs for Enhancing Risk Monitors with Equipment Condition Assessment for Advanced Small Modular Reactors

Energy Technology Data Exchange (ETDEWEB)

Coble, Jamie B.; Coles, Garill A.; Ramuhalli, Pradeep; Meyer, Ryan M.; Berglin, Eric J.; Wootan, David W.; Mitchell, Mark R.

2013-04-04

Advanced small modular reactors (aSMRs) can provide the United States with a safe, sustainable, and carbon-neutral energy source. The controllable day-to-day costs of aSMRs are expected to be dominated by operation and maintenance costs. Health and condition assessment coupled with online risk monitors can potentially enhance affordability of aSMRs through optimized operational planning and maintenance scheduling. Currently deployed risk monitors are an extension of probabilistic risk assessment (PRA). For complex engineered systems like nuclear power plants, PRA systematically combines event likelihoods and the probability of failure (POF) of key components, so that when combined with the magnitude of possible adverse consequences to determine risk. Traditional PRA uses population-based POF information to estimate the average plant risk over time. Currently, most nuclear power plants have a PRA that reflects the as-operated, as-modified plant; this model is updated periodically, typically once a year. Risk monitors expand on living PRA by incorporating changes in the day-by-day plant operation and configuration (e.g., changes in equipment availability, operating regime, environmental conditions). However, population-based POF (or population- and time-based POF) is still used to populate fault trees. Health monitoring techniques can be used to establish condition indicators and monitoring capabilities that indicate the component-specific POF at a desired point in time (or over a desired period), which can then be incorporated in the risk monitor to provide a more accurate estimate of the plant risk in different configurations. This is particularly important for active systems, structures, and components (SSCs) proposed for use in aSMR designs. These SSCs may differ significantly from those used in the operating fleet of light-water reactors (or even in LWR-based SMR designs). Additionally, the operating characteristics of aSMRs can present significantly different

SMITHERS: An object-oriented modular mapping methodology for MCNP-based neutronic–thermal hydraulic multiphysics

International Nuclear Information System (INIS)

Richard, Joshua; Galloway, Jack; Fensin, Michael; Trellue, Holly

2015-01-01

Highlights: • A modular mapping methodogy for neutronic-thermal hydraulic nuclear reactor multiphysics, SMITHERS, has been developed. • Written in Python, SMITHERS takes a novel object-oriented approach for facilitating data transitions between solvers. This approach enables near-instant compatibility with existing MCNP/MONTEBURNS input decks. • It also allows for coupling with thermal-hydraulic solvers of various levels of fidelity. • Two BWR and PWR test problems are presented for verifying correct functionality of the SMITHERS code routines. - Abstract: A novel object-oriented modular mapping methodology for externally coupled neutronics–thermal hydraulics multiphysics simulations was developed. The Simulator using MCNP with Integrated Thermal-Hydraulics for Exploratory Reactor Studies (SMITHERS) code performs on-the-fly mapping of material-wise power distribution tallies implemented by MCNP-based neutron transport/depletion solvers for use in estimating coolant temperature and density distributions with a separate thermal-hydraulic solver. The key development of SMITHERS is that it reconstructs the hierarchical geometry structure of the material-wise power generation tallies from the depletion solver automatically, with only a modicum of additional information required from the user. Additionally, it performs the basis mapping from the combinatorial geometry of the depletion solver to the required geometry of the thermal-hydraulic solver in a generalizable manner, such that it can transparently accommodate varying levels of thermal-hydraulic solver geometric fidelity, from the nodal geometry of multi-channel analysis solvers to the pin-cell level of discretization for sub-channel analysis solvers. The mapping methodology was specifically developed to be flexible enough such that it could successfully integrate preexisting depletion solver case files with different thermal-hydraulic solvers. This approach allows the user to tailor the selection of a

A development strategy for the business plan of Mitsubishi Small-sized High Temperature Gas-cooled Modular Reactor (MHR-50/100is)

International Nuclear Information System (INIS)

Minatsuki, Isao; Otani, Tomomi; Shimizu, Katsusuke; Mizokami, Yorikata; Oyama, Sunao; Tsukamoto, Hiroki

2014-01-01

A business plan and a new concept of Mitsubishi Small-sized High Temperature Gas-cooled Modular Reactor (MHR-50/100is) has been investigated toward a commercialization in near future by Mitsubishi Heavy Industries cooperated with Japan Atomic Energy Agency (JAEA) in Japan. The potential market of small sized reactor is expected to increase from the points of view of smaller investment, industrial use of the nuclear heat and IPP (Independent Power Producer). Especially minimization of construction unit cost including R and D and plant construction period are important issues in order to realize a business plan for them. The study includes four pertinent subject areas of (1) a market analysis, (2) a conceptual design, (3) improvement of safety design and (4) plant dynamics. In summary, the MHR-50/100 is designed to target a short construction period, competitive cost, and an inherent safety feature while applying only the verified technology of the High Temperature Engineering Test Reactor (HTTR) of JAEA or conventional technologies

A development strategy for the business plan of Mitsubishi Small-sized High Temperature Gas-cooled Modular Reactor (MHR-50/100is)

Energy Technology Data Exchange (ETDEWEB)

Minatsuki, Isao, E-mail: isao_minatsuki@mhi.co.jp [Mitsubishi Heavy Industries, Ltd., 16-5, Konan 2-Chome, Minato-ku, Tokyo (Japan); Otani, Tomomi; Shimizu, Katsusuke [Mitsubishi Heavy Industries, Ltd., 16-5, Konan 2-Chome, Minato-ku, Tokyo (Japan); Mizokami, Yorikata; Oyama, Sunao; Tsukamoto, Hiroki [Mitsubishi Heavy Industries, Ltd., 1-1 Wadasaki-cho 1-Chome, Hyogo-ku, Kobe (Japan)

2014-05-01

A business plan and a new concept of Mitsubishi Small-sized High Temperature Gas-cooled Modular Reactor (MHR-50/100is) has been investigated toward a commercialization in near future by Mitsubishi Heavy Industries cooperated with Japan Atomic Energy Agency (JAEA) in Japan. The potential market of small sized reactor is expected to increase from the points of view of smaller investment, industrial use of the nuclear heat and IPP (Independent Power Producer). Especially minimization of construction unit cost including R and D and plant construction period are important issues in order to realize a business plan for them. The study includes four pertinent subject areas of (1) a market analysis, (2) a conceptual design, (3) improvement of safety design and (4) plant dynamics. In summary, the MHR-50/100 is designed to target a short construction period, competitive cost, and an inherent safety feature while applying only the verified technology of the High Temperature Engineering Test Reactor (HTTR) of JAEA or conventional technologies.

Utility/user requirements for the modular high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Boyer, V.S.; Kendall, J.M.; Gotschall, H.L.

1989-01-01

This paper describes the approach used by Gas-Cooled Reactor Associates (GCRA) in developing Utility/User Requirements for the Modular High Temperature Gas-cooled Reactor (MHTGR). As representatives of the Utility/User industry, it is GCRA's goal that the MHTGR concept be established as an attractive nuclear option offering competitive economics and limited ownership risks. Commercially deployed MHTGR systems should then compete favorably in a mixed-fuel economy with options using fossil, other nuclear and other non-fossil sources. To achieve this goal, the design of the MHTGR plant must address the problems experienced by the U.S. industrial infrastructure during deployment of the first generation of nuclear plants. Indeed, it is GCRA's intent to utilize the characteristics of MHTGR technology for the development of a nuclear alternative that poses regulatory, financial and operational demands on the Owner/Operator that are, in aggregate, comparable to those encountered with non-nuclear options. The dominant risks faced by U.S. Utilities with current nuclear plants derive from their operational complexity and the degree of regulatory involvement in virtually all aspects of utility operations. The MHTGR approach of using ceramic fuel coatings to contain fission products provides the technical basis for simplification of the plant and stabilization of licensing requirements and thus the opportunity for reducing the risks of nuclear plant ownership. The paper describes the rationale for the selection of key requirements for public safety, plant size and performance, operations and maintenance, investment protection, economics and siting in the context of a risk management philosophy. It also describes the ongoing participation of the Utility/User in interpreting requirements, conducting program and design reviews and establishing priorities from the Owner/Operator perspective. (author). 7 refs, 1 fig

Development of a methodology for simulation of gas cooled reactors with purpose of transmutation

International Nuclear Information System (INIS)

Silva, Clarysson Alberto da

2009-01-01

This work proposes a methodology of MHR (Modular Helium Reactor) simulation using the WIMSD-5B (Winfrith Improved Multi/group Scheme) nuclear code which is validated by MCNPX 2.6.0 (Monte Carlo N-Particle transport eXtend) nuclear code. The goal is verify the capability of WIMSD-5B to simulate a reactor type GT-MHR (Gas Turbine Modular Helium Reactor), considering all the fuel recharges possibilities. Also is evaluated the possibility of WIMSD-5B to represent adequately the fuel evolution during the fuel recharge. Initially was verified the WIMSD-5B capability to simulate the recharge specificities of this model by analysis of neutronic parameters and isotopic composition during the burnup. After the model was simulated using both WIMSD-5B and MCNPX 2.6.0 codes and the results of k eff , neutronic flux and isotopic composition were compared. The results show that the deterministic WIMSD-5B code can be applied to a qualitative evaluation, representing adequately the core behavior during the fuel recharges being possible in a short period of time to inquire about the burned core that, once optimized, can be quantitatively evaluated by a code type MCNPX 2.6.0. (author)

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

A Framework to Expand and Advance Probabilistic Risk Assessment to Support Small Modular Reactors

Energy Technology Data Exchange (ETDEWEB)

Curtis Smith; David Schwieder; Robert Nourgaliev; Cherie Phelan; Diego Mandelli; Kellie Kvarfordt; Robert Youngblood

2012-09-01

During the early development of nuclear power plants, researchers and engineers focused on many aspects of plant operation, two of which were getting the newly-found technology to work and minimizing the likelihood of perceived accidents through redundancy and diversity. As time, and our experience, has progressed, the realization of plant operational risk/reliability has entered into the design, operation, and regulation of these plants. But, to date, we have only dabbled at the surface of risk and reliability technologies. For the next generation of small modular reactors (SMRs), it is imperative that these technologies evolve into an accepted, encompassing, validated, and integral part of the plant in order to reduce costs and to demonstrate safe operation. Further, while it is presumed that safety margins are substantial for proposed SMR designs, the depiction and demonstration of these margins needs to be better understood in order to optimize the licensing process.

High-temperature reactors: a recent past, a near future

International Nuclear Information System (INIS)

Ballot, B.

2007-01-01

While high-temperature reactors did experience major developments in the past, in Europe in particular, significant R and D efforts are required, if a major innovation deployment is to be made possible, of modular reactors having the capability of being coupled, in reliable, economic fashion, to an industrial process. The aim: the construction, before the next decade is out more swiftly than is feasible for other fourth-generation systems of an industrial prototype, coupled to such a process. The Areva Group takes up this approach, with its ANTARES project. For the purposes of characterizing the thermal properties of heterogeneous, multi-scale materials, as a function of temperature, experimental and numerical instruments have been developed at the Microstructure and Behavior Laboratory (Laboratoire microstructure et comportement), at CEA Le Ripault Center. They have been applied to the thermal characterization of the various layers in a high temperature reactor (HTR) fuel ball. (authors)

A multi-crucible core-catcher concept: Design considerations and basic results

International Nuclear Information System (INIS)

Szabo, I.

1995-01-01

A multi-crucible core-catcher concept to be implemented in new light water reactor containments has recently been proposed. This paper deals with conceptual design considerations and the various ways this type of core-catcher could be designed to meet requirements for reactor application. A systematic functional analysis of the multi-crucible core-catcher concept and the results of the preliminary design calculation are presented. Finally, the adequacy of the multi-crucible core-catcher concept for reactor application is discussed. (orig.)

Linear and Non-linear Multi-Input Multi-Output Model Predictive Control of Continuous Stirred Tank Reactor

Directory of Open Access Journals (Sweden)

Muayad Al-Qaisy

2015-02-01

Full Text Available In this article, multi-input multi-output (MIMO linear model predictive controller (LMPC based on state space model and nonlinear model predictive controller based on neural network (NNMPC are applied on a continuous stirred tank reactor (CSTR. The idea is to have a good control system that will be able to give optimal performance, reject high load disturbance, and track set point change. In order to study the performance of the two model predictive controllers, MIMO Proportional-Integral-Derivative controller (PID strategy is used as benchmark. The LMPC, NNMPC, and PID strategies are used for controlling the residual concentration (CA and reactor temperature (T. NNMPC control shows a superior performance over the LMPC and PID controllers by presenting a smaller overshoot and shorter settling time.

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-08-15

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

Demonstration of a Small Modular BioPower System Using Poultry Litter; FINAL

International Nuclear Information System (INIS)

John P. Reardon; Art Lilley; Jim Wimberly; Kingsbury Browne; Kelly Beard; Jack Avens

2002-01-01

The purpose of this project was to assess poultry grower residue, or litter (manure plus absorbent biomass), as a fuel source for Community Power Corporation's small modular biopower system (SMB). A second objective was to assess the poultry industry to identify potential ''on-site'' applications of the SMB system using poultry litter residue as a fuel source, and to adapt CPC's existing SMB to generate electricty and heat from the poultry litter biomass fuel. Bench-scale testing and pilot testing were used to gain design information for the SMB retrofit. System design approach for the Phase II application of the SMB was the goal of Phase I testing. Cost estimates for an onsite poultry litter SMB were prepared. Finally, a market estimate was prepared for implementation of the on-farm SMB using poultry litter

Study on disposal method of graphite blocks and storage of spent fuel for modular gas-cooled reactor. Joint research

Energy Technology Data Exchange (ETDEWEB)

Sumita, Junya; Sawa, Kazuhiro; Kunitomi, Kazuhiko [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Tsuchie, Yasuo; Urakami, Masao [Japan Atomic Power Co., Tokyo (Japan)

2003-02-01

This report describes the result of study on disposal method of graphite blocks in future block-type reactor. Present study was carried out within a framework of joint research, ''Research of Modular High Temperature Gas-cooled Reactors (No. 3)'', between Japan Atomic Energy Research Institute (JAERI) and the Japan Atomic Power Company (JAPCO), in 2000. In this study, activities in fuel and reflector graphite blocks were evaluated and were compared with the disposal limits defined as low-level of radioactive waste. As a result, it was found that the activity for only C-14 was higher than disposal limits for the low-level of radioactive waste and that the amount of air in the graphite is important to evaluate precisely of C-14 activity. In addition, spent fuels can be stored in air-cooled condition at least after two years cooling in the storage pool. (author)

Nuclear Burning Wave Modular Fast Reactor Concept

International Nuclear Information System (INIS)

Kodochigov, N.G.; Sukharev, Yu.P.

2014-01-01

The necessity to provide nuclear power industry, comparable in a scope with power industry based on a traditional fuel, inspired studies of an open-cycle fast reactor aimed at: - solution of the problem of fuel provision by implementing the highest breeding characteristics of new fissile materials of raw isotopes in a fast reactor and applying accumulated fissile isotopes in the same reactor, independently on a spent fuel reprocessing rate in the external fuel cycle; - application of natural or depleted uranium for makeup fuel, which, with no spent fuel reprocessing, forms the most favorable non-proliferation conditions; - application of inherent properties of the core and reactor for safety provision. The present report, based on previously published papers, gives the theoretical backgrounds of the concept of the reactor with a nuclear burning wave, in which an enriched-fuel core (driver) is replaced by a blanket, and basic conditions for nuclear burning wave initiating and keeping are shown. (author)

Combining a gas turbine modular helium reactor and an accelerator and for near total destruction of weapons grade plutonium

Energy Technology Data Exchange (ETDEWEB)

Baxter, A.M.; Lane, R.K.; Sherman, R. [General Atomics, San Diego, CA (United States)

1995-10-01

Fissioning surplus weapons-grade plutonium (WG-Pu) in a reactor is an effective means of rendering this stockpile non-weapons useable. In addition the enormous energy content of the plutonium is released by the fission process and can be captured to produce valuable electric power. While no fission option has been identified that can accomplish the destruction of more than about 70% of the WG-Pu without repeated reprocessing and recycling, which presents additional opportunities for diversion, the gas turbine modular helium-cooled reactor (GT-MHR), using an annular graphite core and graphite inner and outer reflectors combines the maximum plutonium destruction and highest electrical production efficiency and economics in an inherently safe system. Accelerator driven sub-critical assemblies have also been proposed for WG-Pu destruction. These systems offer almost complete WG-Pu destruction, but achieve this goal by using circulating aqueous or molten salt solutions of the fuel, with potential safety implications. By combining the GT-MHR with an accelerator-driven sub-critical MHR assembly, the best features of both systems can be merged to achieve the near total destruction of WG-Pu in an inherently safe, diversion-proof system in which the discharged fuel elements are suitable for long term high level waste storage without the need for further processing. More than 90% total plutonium destruction, and more than 99.9% Pu-239 destruction, could be achieved. The modular concept minimizes the size of each unit so that both the GT-MHR and the accelerator would be straightforward extensions of current technology.

Pellet bed reactor for multi-modal space power

International Nuclear Information System (INIS)

Buden, D.; Williams, K.; Mast, P.; Mims, J.

1987-01-01

A review of forthcoming space power needs for both civil and military missions indicates that power requirements will be in the tens of megawatts. The electrical power requirements are envisioned to be twofold: long-duration lower power levels will be needed for station keeping, communications, and/or surveillance; short-duration higher power levels will be required for pulsed power devices. These power characteristics led to the proposal of a multi-modal space power reactor using a pellet bed design. Characteristics desired for such a multimegawatt reactor power source are standby, alert, and pulsed power modes; high-thermal output heat source (approximately 1000 MWt peak power); long lifetime station keeping power (10 to 30 years); high temperature output (1500 K to 1800 K); rapid-burst power transition; high reliability (above 95 percent); and stringent safety standards compliance. The proposed pellet bed reactor is designed to satisfy these characteristics

Development of refined MCNPX-PARET multi-channel model for transient analysis in research reactors

Energy Technology Data Exchange (ETDEWEB)

Kalcheva, S.; Koonen, E. [SCK-CEN, BR2 Reactor Dept., Boeretang 200, 2400 Mol (Belgium); Olson, A. P. [RERTR Program, Nuclear Engineering Div., Argonne National Laboratory, Cass Avenue, Argonne, IL 60439 (United States)

2012-07-01

Reactivity insertion transients are often analyzed (RELAP, PARET) using a two-channel model, representing the hot assembly with specified power distribution and an average assembly representing the remainder of the core. For the analysis of protected by the reactor safety system transients and zero reactivity feedback coefficients this approximation proves to give adequate results. However, a more refined multi-channel model representing the various assemblies, coupled through the reactivity feedback effects to the whole reactor core is needed for the analysis of unprotected transients with excluded over power and period trips. In the present paper a detailed multi-channel PARET model has been developed which describes the reactor core in different clusters representing typical BR2 fuel assemblies. The distribution of power and reactivity feedback in each cluster of the reactor core is obtained from a best-estimate MCNPX calculation using the whole core geometry model of the BR2 reactor. The sensitivity of the reactor response to power, temperature and energy distributions is studied for protected and unprotected reactivity insertion transients, with zero and non-zero reactivity feedback coefficients. The detailed multi-channel model is compared vs. simplified fewer-channel models. The sensitivities of transient characteristics derived from the different models are tested on a few reactivity insertion transients with reactivity feedback from coolant temperature and density change. (authors)

Electric Energy Consumption of Multi Purpose Reactor GA. Siwabessy During Reactor Operation

International Nuclear Information System (INIS)

Koes Indrakoesoema

2012-01-01

Electrical power supply of Reactor Center Multi Purpose obtained from PT PLN to 3030 kVA power contracts. Distribution to existing loads in PRSG divided into 3 (three) lines, each of which is supplied through a transformer BHT01, BHT02 and BHT03, each transformer have capacity of 1600 kVA. During reactor operation, only 2 lines that serve loads, each line serve 2 primary pump motor and 2 secondary pump motor. Electrical power for 24 hours for measurement BHT01, the average is 288 kW, for BHT02 is 641 kW and BHT03 is 466 kW. The energy absorbed by each transformer for 24 hours of measurement, for BHT01 is 6.44 MWh, BHT02 absorb 14.8 MWh and BHT03 absorb 10.9 MWh. (author)

Introduction of an agent-based multi-scale modular architecture for dynamic knowledge representation of acute inflammation.

Science.gov (United States)

An, Gary

2008-05-27

One of the greatest challenges facing biomedical research is the integration and sharing of vast amounts of information, not only for individual researchers, but also for the community at large. Agent Based Modeling (ABM) can provide a means of addressing this challenge via a unifying translational architecture for dynamic knowledge representation. This paper presents a series of linked ABMs representing multiple levels of biological organization. They are intended to translate the knowledge derived from in vitro models of acute inflammation to clinically relevant phenomenon such as multiple organ failure. ABM development followed a sequence starting with relatively direct translation from in-vitro derived rules into a cell-as-agent level ABM, leading on to concatenated ABMs into multi-tissue models, eventually resulting in topologically linked aggregate multi-tissue ABMs modeling organ-organ crosstalk. As an underlying design principle organs were considered to be functionally composed of an epithelial surface, which determined organ integrity, and an endothelial/blood interface, representing the reaction surface for the initiation and propagation of inflammation. The development of the epithelial ABM derived from an in-vitro model of gut epithelial permeability is described. Next, the epithelial ABM was concatenated with the endothelial/inflammatory cell ABM to produce an organ model of the gut. This model was validated against in-vivo models of the inflammatory response of the gut to ischemia. Finally, the gut ABM was linked to a similarly constructed pulmonary ABM to simulate the gut-pulmonary axis in the pathogenesis of multiple organ failure. The behavior of this model was validated against in-vivo and clinical observations on the cross-talk between these two organ systems. A series of ABMs are presented extending from the level of intracellular mechanism to clinically observed behavior in the intensive care setting. The ABMs all utilize cell-level agents

Numerical investigation of flow and heat transfer in a novel configuration multi-tubular fixed bed reactor for propylene to acrolein process

Science.gov (United States)

Jiang, Bin; Hao, Li; Zhang, Luhong; Sun, Yongli; Xiao, Xiaoming

2015-01-01

In the present contribution, a numerical study of fluid flow and heat transfer performance in a pilot-scale multi-tubular fixed bed reactor for propylene to acrolein oxidation reaction is presented using computational fluid dynamics (CFD) method. Firstly, a two-dimensional CFD model is developed to simulate flow behaviors, catalytic oxidation reaction, heat and mass transfer adopting porous medium model on tube side to achieve the temperature distribution and investigate the effect of operation parameters on hot spot temperature. Secondly, based on the conclusions of tube-side, a novel configuration multi-tubular fixed-bed reactor comprising 790 tubes design with disk-and-doughnut baffles is proposed by comparing with segmental baffles reactor and their performance of fluid flow and heat transfer is analyzed to ensure the uniformity condition using molten salt as heat carrier medium on shell-side by three-dimensional CFD method. The results reveal that comprehensive performance of the reactor with disk-and-doughnut baffles is better than that of with segmental baffles. Finally, the effects of operating conditions to control the hot spots are investigated. The results show that the flow velocity range about 0.65 m/s is applicable and the co-current cooling system flow direction is better than counter-current flow to control the hottest temperature.

Evaluation and optimization of General Atomics' GT-MHR reactor cavity cooling system using an axiomatic design approach

International Nuclear Information System (INIS)

Thielman, Jeff; Ge, Ping; Wu, Qiao; Parme, Laurence

2005-01-01

The development of the Generation IV (Gen-IV) nuclear reactors has presented social, technical, and economical challenges to nuclear engineering design and research. To develop a robust, reliable nuclear reactor system with minimal environmental impact and cost, modularity has been gradually accepted as a key concept in designing high-quality nuclear reactor systems. While the establishment and reliability of a nuclear power plant is largely facilitated by the installment of standardized base units, the realization of modularity at the sub-system/sub-unit level in a base unit is still highly heuristic, and lacks consistent, quantifiable measures. In this work, an axiomatic design approach is developed to evaluate and optimize the reactor cavity cooling system (RCCS) of General Atomics' Gas Turbine-Modular Helium Reactor (GT-MHR) nuclear reactor, for the purpose of constructing a quantitative tool that is applicable to Gen-IV systems. According to Suh's axiomatic design theory, modularity is consistently represented by functional independence through the design process. Both qualitative and quantitative measures are developed here to evaluate the modularity of the current RCCS design. Optimization techniques are also used to improve the modularity at both conceptual and parametric level. The preliminary results of this study have demonstrated that the axiomatic design approach has great potential in enhancing modular design, and generating more robust, safer, and less expensive nuclear reactor sub-units

Study on modular technology based on the reference nuclear power plant

International Nuclear Information System (INIS)

Lu Qinwu; Zhao Shuyu; Li Yi; Han Xiaoping; Chai Weidong; Zhang Shuxia

2014-01-01

As an important technology for shortening the construction period and reducing the cost of nuclear power project, modular technology has become one of the development orientations of the third generation nuclear power plant. Taking CPR1000 nuclear power reactor as reference plant and drawing on advanced foreign technology of modularization, the overall concept design has been completed with the help of the self-developed 3D modular design system. On this basis, a typical detailed design work has been carried out so as to verify the feasibility of technology. Besides, an analysis on implementation of modular technology has been made from two aspects (procurement and construction), and the feasibility of research results in actual project has been evaluated. It provides references for the promotion of self-reliant modular technology in nuclear power project in China. (authors)

Modular high-temperature reactor launched (and wallchart)

International Nuclear Information System (INIS)

Steinwarz, W.

1987-01-01

In view of the need for a technically unsophisticated, safe and economic reactor system, the KWU group has integrated the experience gained from German light-water reactor engineering and from successful operation of the German AVR experimental high-temperature reactor into the development of the High-Temperature Reactor (HTR)-module. The main components are illustrated and explained and technical data for the HTR-module is given. Safety is also considered. This includes graphs of core heat-up temperature for pebble-bed HTR and a graph of the temperature load of the fuel elements. The operation, control and applications are considered. The latter includes use in combined heat and power generation and community heating. Feasibility studies have shown that the HTR-module is cheaper, comparatively, than coal-fired power stations. (U.K.)

Managing the reactivity excess of the gas turbine-modular helium reactor by burnable poison and control rods

International Nuclear Information System (INIS)

Talamo, Alberto

2006-01-01

The gas turbine-modular helium reactor coupled to the deep burn in-core fuel management strategy offers the extraordinary capability to incinerate over 50% of the initial inventory of fissile material. This extraordinary feature, coming from an advanced and well tested fuel element design, which takes advantage of the TRISO particles technology, is maintained while the reactor is loaded with the most different types of fuels. In the present work, we assumed the reactor operating at the equilibrium of the fuel composition, obtained by a 6 years irradiation of light water reactor waste, and we investigated the effects of the introduction of the burnable poison and the control rods; we equipped the core with all the three types of control rods: operational, startup and shutdown ones. We employed as burnable poison natural erbium, due to the 167 Er increasing neutron capture microscopic cross-section in the energy range where the neutron spectrum exhibits the thermal peak; in addition, we utilized boron carbide, with 90% enrichment in 1 B, as the absorption material of the control rods. Concerning the burnable poison studies, we focused on the k eff value, the 167 Er mass during burnup, the influence of modifying the radius of the BISO particles kernel and the fuel and moderator coefficients of temperature. Concerning the control rods studies, we investigated the reactivity worth, the changes in the neutron flux profile due to a partial insertion, the influence of modifying the radius of the BISO particles kernel and the β eff , at the beginning of the operation

Managing the reactivity excess of the gas turbine-modular helium reactor by burnable poison and control rods

Energy Technology Data Exchange (ETDEWEB)

Talamo, Alberto [Department of Nuclear and Reactor Physics, Royal Institute of Technology, Roslagstullsbacken 21, S-10691, Stockholm (Sweden)]. E-mail: alby@neutron.kth.se

2006-01-15

The gas turbine-modular helium reactor coupled to the deep burn in-core fuel management strategy offers the extraordinary capability to incinerate over 50% of the initial inventory of fissile material. This extraordinary feature, coming from an advanced and well tested fuel element design, which takes advantage of the TRISO particles technology, is maintained while the reactor is loaded with the most different types of fuels. In the present work, we assumed the reactor operating at the equilibrium of the fuel composition, obtained by a 6 years irradiation of light water reactor waste, and we investigated the effects of the introduction of the burnable poison and the control rods; we equipped the core with all the three types of control rods: operational, startup and shutdown ones. We employed as burnable poison natural erbium, due to the {sup 167}Er increasing neutron capture microscopic cross-section in the energy range where the neutron spectrum exhibits the thermal peak; in addition, we utilized boron carbide, with 90% enrichment in {sup 1}B, as the absorption material of the control rods. Concerning the burnable poison studies, we focused on the k {sub eff} value, the {sup 167}Er mass during burnup, the influence of modifying the radius of the BISO particles kernel and the fuel and moderator coefficients of temperature. Concerning the control rods studies, we investigated the reactivity worth, the changes in the neutron flux profile due to a partial insertion, the influence of modifying the radius of the BISO particles kernel and the {beta} {sub eff}, at the beginning of the operation.

Preliminary safety evaluation of the Gas Turbine-Modular Helium Reactor (GT-MHR)

International Nuclear Information System (INIS)

Dunn, T.D.; Lommers, L.J.; Tangirala, V.E.

1994-04-01

A qualitative comparison between the safety characteristics of the Gas Turbine-Modular Helium Reactor (GT-MHR) and those of the steam cycle shows that the two designs achieve equivalent levels of overall safety performance. This comparison is obtained by applying the scaling laws to detailed steam-cycle computations as well as the conclusions obtained from preliminary GT-MHR model simulations. The gas turbine design is predicted to be superior for some event categories, while the steam cycle design is better for others. From a safety perspective, the GT-MHR has a modest advantage for pressurized conduction cooldown events. Recent computational simulations of 102 column, 550 MW(t) GT-MHR during a depressurized conduction cooldown show that peak fuel temperatures are within the limits. The GT-MHR has a significantly lower risk due to water ingress events under operating conditions. Two additional scenarios, namely loss of load event and turbine deblading event that are specific to the GT-MHR design are discussed. Preliminary evaluation of the GT-MHR's safety characteristics indicate that the GT-MHR can be expected to satisfy or exceed its safety requirements

Application and development analysis of nuclear power plant modular construction

International Nuclear Information System (INIS)

Fang Xiaopeng

2015-01-01

Modular Construction is currently one of the major development trends for the nuclear power plant construction technology worldwide. In the first-of-a-kind AP1000 Nuclear Power Project practiced in China, the large-scale structural modules and mechanical modules have been successfully fabricated, assembled and installed. However, in the construction practice of the project, some quality issues are identified with the assembly and installation process of large-scale structural modules in addition to the issue of incomplete supply of mechanical modules, which has failed to fully demonstrate the features and merits of modular construction. This paper collects and consolidates the issues of modular construction of AP1000 first of a kind reactor, providing root cause analysis in the aspects of process design, quality control, site construction logic, interface management in the process of module fabrication, assembly and installation; modular construction feasibility assessment index is proved based on the quantification and qualitative analysis of the impact element. Based on the modular construction feasibility, NPP modular construction improvement suggestions are provided in the aspect of modular assembly optimization definition, tolerance control during the fitting process and the construction logic adjustment. (author)

WARRIOR II, a high performance modular electric robot system

International Nuclear Information System (INIS)

Downton, G.C.

1996-01-01

Initially designed for in-reactor welding by the Central Electricity Generating Board, WARRIOR has been developed using the concept of modular technology to become a light-weight, high performance robotic system. Research work on existing machines for in-reactor inspection and repair and heavy duty hydraulic manipulators was progressed in order to develop WARRIOR II, a versatile in-reactor welding system usable at any nuclear power station light enough to be deployed by existing remote handling equipment. WARRIOR II can be significantly reconfigured quickly to pursue different ends. (UK)

Traffic-aware energy saving scheme with modularization supporting in TWDM-PON

Science.gov (United States)

Xiong, Yu; Sun, Peng; Liu, Chuanbo; Guan, Jianjun

2017-01-01

Time and wavelength division multiplexed passive optical network (TWDM-PON) is considered to be a primary solution for next-generation passive optical network stage 2 (NG-PON2). Due to the feature of multi-wavelength transmission of TWDM-PON, some of the transmitters/receivers at the optical line terminal (OLT) could be shut down to reduce the energy consumption. Therefore, a novel scheme called traffic-aware energy saving scheme with modularization supporting is proposed. Through establishing the modular energy consumption model of OLT, the wavelength transmitters/receivers at OLT could be switched on or shut down adaptively depending on sensing the status of network traffic load, thus the energy consumption of OLT will be effectively reduced. Furthermore, exploring the technology of optical network unit (ONU) modularization, each module of ONU could be switched to sleep or active mode independently in order to reduce the energy consumption of ONU. Simultaneously, the polling sequence of ONU could be changed dynamically via sensing the packet arrival time. In order to guarantee the delay performance of network traffic, the sub-cycle division strategy is designed to transmit the real-time traffic preferentially. Finally, simulation results verify that the proposed scheme is able to reduce the energy consumption of the network while maintaining the traffic delay performance.

Preliminary LOCA analysis of the westinghouse small modular reactor using the WCOBRA/TRAC-TF2 thermal-hydraulics code

Energy Technology Data Exchange (ETDEWEB)

Liao, J.; Kucukboyaci, V. N.; Nguyen, L.; Frepoli, C. [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

2012-07-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (> 225 MWe) integral pressurized water reactor (iPWR) with all primary components, including the steam generator and the pressurizer located inside the reactor vessel. The reactor core is based on a partial-height 17x17 fuel assembly design used in the AP1000{sup R} reactor core. The Westinghouse SMR utilizes passive safety systems and proven components from the AP1000 plant design with a compact containment that houses the integral reactor vessel and the passive safety systems. A preliminary loss of coolant accident (LOCA) analysis of the Westinghouse SMR has been performed using the WCOBRA/TRAC-TF2 code, simulating a transient caused by a double ended guillotine (DEG) break in the direct vessel injection (DVI) line. WCOBRA/TRAC-TF2 is a new generation Westinghouse LOCA thermal-hydraulics code evolving from the US NRC licensed WCOBRA/TRAC code. It is designed to simulate PWR LOCA events from the smallest break size to the largest break size (DEG cold leg). A significant number of fluid dynamics models and heat transfer models were developed or improved in WCOBRA/TRAC-TF2. A large number of separate effects and integral effects tests were performed for a rigorous code assessment and validation. WCOBRA/TRAC-TF2 was introduced into the Westinghouse SMR design phase to assist a quick and robust passive cooling system design and to identify thermal-hydraulic phenomena for the development of the SMR Phenomena Identification Ranking Table (PIRT). The LOCA analysis of the Westinghouse SMR demonstrates that the DEG DVI break LOCA is mitigated by the injection and venting from the Westinghouse SMR passive safety systems without core heat up, achieving long term core cooling. (authors)

Summary report of the final technical meeting on 'International Reactor Dosimetry File: IRDF-2002'

International Nuclear Information System (INIS)

Griffin, Patrick J.; Paviotti-Corcuera, R.

2003-10-01

Presentations, recommendations and conclusions of the Final Technical Meeting on 'International Reactor Dosimetry File: IRDF-2002' are summarized in this report. The main aims of this meeting were to discuss scientific and technical matters related to reactor dosimetry and to assign responsibilities for the preparation of the final version of the IRDF- 2002 library and the associated TECDOC. Tasks were assigned and deadlines were agreed. Participants emphasized that accurate and complete nuclear data for reactor dosimetry are essential to improve the assessment accuracies for reactor pressure vessel service lifetimes in nuclear power plants, as well as for other neutron metrology applications such as boron neutron capture therapy, therapeutic use of medical isotopes, nuclear physics measurements, and reactor safety applications. (author)

Playware Soccer – flexibility through modularity and layered multi-modal feedback

DEFF Research Database (Denmark)

Lund, Henrik Hautop

2010-01-01

We developed the Playware Soccer game and tested this with more than 1,000 users during the FIFA World Cup 2010 in South Africa in townships, orphanages for HIV/AIDS children, markets, FIFA fan parks, etc. The playware game is set up to motivate players to engage in training of technical soccer...... skills by receiving motivating, immediate feedback on the soccer playing on a modular interactive wall composed of modular interactive tiles that respond with coloured light, sound and scores on the players performance. The flexibility of the system was designed for with the modular interactive tiles......, and simple construction with these building blocks should give a high degree of flexibility for the designer and the user to create various set-ups and interaction possibilities in an easy manner....

Development of CFD software for the simulation of thermal hydraulics in advanced nuclear reactors. Final report

International Nuclear Information System (INIS)

Bachar, Abdelaziz; Haslinger, Wolfgang; Scheuerer, Georg; Theodoridis, Georgios

2015-01-01

The objectives of the project were: Improvement of the simulation accuracy for nuclear reactor thermo-hydraulics by coupling system codes with three-dimensional CFD software; Extension of CFD software to predict thermo-hydraulics in advanced reactor concepts; Validation of the CFD software by simulation different UPTF TRAM-C test cases and development of best practice guidelines. The CFD module was based on the ANSYS CFD software and the system code ATHLET of GRS. All three objectives were met: The coupled ATHLET-ANSYS CFD software is in use at GRS and TU Muenchen. Besides the test cases described in the report, it has been used for other applications, for instance the TALL-3D experiment of KTH Stockholm. The CFD software was extended with material properties for liquid metals, and validated using existing data. Several new concepts were tested when applying the CFD software to the UPTF test cases: Simulations with Conjugate Heat Transfer (CHT) were performed for the first time. This led to better agreement between predictions and data and reduced uncertainties when applying temperature boundary conditions. The meshes for the CHT simulation were also used for a coupled fluid-structure-thermal analysis which was another novelty. The results of the multi-physics analysis showed plausible results for the mechanical and thermal stresses. The workflow developed as part of the current project can be directly used for industrial nuclear reactor simulations. Finally, simulations for two-phase flows with and without interfacial mass transfer were performed. These showed good agreement with data. However, a persisting problem for the simulation of multi-phase flows are the long simulation times which make use for industrial applications difficult.

Thermal-Hydraulic Design of the Modular Once Through Helical Steam Generator

International Nuclear Information System (INIS)

Mazufri, C.M

2000-01-01

The steam generator system of the CAREM reactor consists of twelve individual modules located in the annular place between the pressure vessel and barrel walls. Each steam generator module consists of a tube system, an upper header, an external shroud, a collector and a lower seal.The tube system is an arrangement of several multi-start cylindrical coils.In the present work the computation of the necessary heat transfer area to fulfill the heat removal requirements from the primary circuit, and the pressure drop in the primary and secondary side of the helical design of a modular steam generator is presented. Additionally, a first order estimation of the restriction to be used in the secondary side to assure the thermal-hydraulic stability is also made.It is concluded that an array of 6 concentric cylindrical coils fulfills the necessary design requirements

Modular remote radiation monitor

International Nuclear Information System (INIS)

Lacerda, Fabio; Farias, Marcos S.; Aghina, Mauricio A.C.; Oliveira, Mauro V.