Molten salt reactor neutronics and fuel cycle modeling and simulation with SCALE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Betzler, Benjamin R.; Powers, Jeffrey J.; Worrall, Andrew

Current interest in advanced nuclear energy and molten salt reactor (MSR) concepts has enhanced interest in building the tools necessary to analyze these systems. A Python script known as ChemTriton has been developed to simulate equilibrium MSR fuel cycle performance by modeling the changing isotopic composition of an irradiated fuel salt using SCALE for neutron transport and depletion calculations. Some capabilities in ChemTriton that have improved, include a generic geometry capable of modeling multi-zone and multi-fluid systems, enhanced time-dependent feed and separations, and a critical concentration search. Although more generally applicable, the capabilities developed to date are illustrated in thismore » paper in three applied problems: (1) simulating the startup of a thorium-based MSR fuel cycle (a likely scenario requires the first of these MSRs to be started without available 233U); (2) determining the effect of the removal of different fission products on MSR operations; and (3) obtaining the equilibrium concentration of a mixed-oxide light-water reactor fuel in a two-stage fuel cycle with a sodium fast reactor. Moreover, the third problem is chosen to demonstrate versatility in an application to analyze the fuel cycle of a non-MSR system. During the first application, the initial fuel salt compositions fueled with different sources of fissile material are made feasible after (1) removing the associated nonfissile actinides after much of the initial fissile isotopes have burned and (2) optimizing the thorium concentration to maintain a critical configuration without significantly reducing breeding capability. In the second application, noble metal, volatile gas, and rare earth element fission products are shown to have a strong negative effect on criticality in a uranium-fueled thermal-spectrum MSR; their removal significantly increases core lifetime (by 30%) and fuel utilization. In the third application, the fuel of a mixed-oxide light

Molten salt reactor neutronics and fuel cycle modeling and simulation with SCALE

DOE PAGES

Betzler, Benjamin R.; Powers, Jeffrey J.; Worrall, Andrew

2017-03-01

Current interest in advanced nuclear energy and molten salt reactor (MSR) concepts has enhanced interest in building the tools necessary to analyze these systems. A Python script known as ChemTriton has been developed to simulate equilibrium MSR fuel cycle performance by modeling the changing isotopic composition of an irradiated fuel salt using SCALE for neutron transport and depletion calculations. Some capabilities in ChemTriton that have improved, include a generic geometry capable of modeling multi-zone and multi-fluid systems, enhanced time-dependent feed and separations, and a critical concentration search. Although more generally applicable, the capabilities developed to date are illustrated in thismore » paper in three applied problems: (1) simulating the startup of a thorium-based MSR fuel cycle (a likely scenario requires the first of these MSRs to be started without available 233U); (2) determining the effect of the removal of different fission products on MSR operations; and (3) obtaining the equilibrium concentration of a mixed-oxide light-water reactor fuel in a two-stage fuel cycle with a sodium fast reactor. Moreover, the third problem is chosen to demonstrate versatility in an application to analyze the fuel cycle of a non-MSR system. During the first application, the initial fuel salt compositions fueled with different sources of fissile material are made feasible after (1) removing the associated nonfissile actinides after much of the initial fissile isotopes have burned and (2) optimizing the thorium concentration to maintain a critical configuration without significantly reducing breeding capability. In the second application, noble metal, volatile gas, and rare earth element fission products are shown to have a strong negative effect on criticality in a uranium-fueled thermal-spectrum MSR; their removal significantly increases core lifetime (by 30%) and fuel utilization. In the third application, the fuel of a mixed-oxide light

A combined gas cooled nuclear reactor and fuel cell cycle

NASA Astrophysics Data System (ADS)

Palmer, David J.

Rising oil costs, global warming, national security concerns, economic concerns and escalating energy demands are forcing the engineering communities to explore methods to address these concerns. It is the intention of this thesis to offer a proposal for a novel design of a combined cycle, an advanced nuclear helium reactor/solid oxide fuel cell (SOFC) plant that will help to mitigate some of the above concerns. Moreover, the adoption of this proposal may help to reinvigorate the Nuclear Power industry while providing a practical method to foster the development of a hydrogen economy. Specifically, this thesis concentrates on the importance of the U.S. Nuclear Navy adopting this novel design for its nuclear electric vessels of the future with discussion on efficiency and thermodynamic performance characteristics related to the combined cycle. Thus, the goals and objectives are to develop an innovative combined cycle that provides a solution to the stated concerns and show that it provides superior performance. In order to show performance, it is necessary to develop a rigorous thermodynamic model and computer program to analyze the SOFC in relation with the overall cycle. A large increase in efficiency over the conventional pressurized water reactor cycle is realized. Both sides of the cycle achieve higher efficiencies at partial loads which is extremely important as most naval vessels operate at partial loads as well as the fact that traditional gas turbines operating alone have poor performance at reduced speeds. Furthermore, each side of the cycle provides important benefits to the other side. The high temperature exhaust from the overall exothermic reaction of the fuel cell provides heat for the reheater allowing for an overall increase in power on the nuclear side of the cycle. Likewise, the high temperature helium exiting the nuclear reactor provides a controllable method to stabilize the fuel cell at an optimal temperature band even during transients helping

Sustainable Thorium Nuclear Fuel Cycles: A Comparison of Intermediate and Fast Neutron Spectrum Systems

DOE PAGES

Brown, Nicholas R.; Powers, Jeffrey J.; Feng, B.; ...

2015-05-21

This paper presents analyses of possible reactor representations of a nuclear fuel cycle with continuous recycling of thorium and produced uranium (mostly U-233) with thorium-only feed. The analysis was performed in the context of a U.S. Department of Energy effort to develop a compendium of informative nuclear fuel cycle performance data. The objective of this paper is to determine whether intermediate spectrum systems, having a majority of fission events occurring with incident neutron energies between 1 eV and 10 5 eV, perform as well as fast spectrum systems in this fuel cycle. The intermediate spectrum options analyzed include tight latticemore » heavy or light water-cooled reactors, continuously refueled molten salt reactors, and a sodium-cooled reactor with hydride fuel. All options were modeled in reactor physics codes to calculate their lattice physics, spectrum characteristics, and fuel compositions over time. Based on these results, detailed metrics were calculated to compare the fuel cycle performance. These metrics include waste management and resource utilization, and are binned to accommodate uncertainties. The performance of the intermediate systems for this selfsustaining thorium fuel cycle was similar to a representative fast spectrum system. However, the number of fission neutrons emitted per neutron absorbed limits performance in intermediate spectrum systems.« less

Synfuels from fusion: using the tandem mirror reactor and a thermochemical cycle to produce hydrogen

DOE Office of Scientific and Technical Information (OSTI.GOV)

Werner, R.W.

1982-11-01

This study is concerned with the following area: (1) the tandem mirror reactor and its physics; (2) energy balance; (3) the lithium oxide canister blanket system; (4) high-temperature blanket; (5) energy transport system-reactor to process; (6) thermochemical hydrogen processes; (7) interfacing the GA cycle; (8) matching power and temperature demands; (9) preliminary cost estimates; (10) synfuels beyond hydrogen; and (11) thermodynamics of the H/sub 2/SO/sub 4/-H/sub 2/O system. (MOW)

The basic features of a closed fuel cycle without fast reactors

NASA Astrophysics Data System (ADS)

Bobrov, E. A.; Alekseev, P. N.; Teplov, P. S.

2017-01-01

In this paper the basic features of a closed fuel cycle with thermal reactors are considered. The three variants of multiple Pu and U recycling in VVER reactors was investigated. The comparison of MOX and REMIX fuel approaches for closed fuel cycle with thermal reactors is presented. All variants make possible to recycle several times the total amount of Pu and U obtained from spent fuel. The reported study was funded by RFBR according to the research project № 16-38-00021

Preliminary Design Study of Medium Sized Gas Cooled Fast Reactor with Natural Uranium as Fuel Cycle Input

NASA Astrophysics Data System (ADS)

Meriyanti, Su'ud, Zaki; Rijal, K.; Zuhair, Ferhat, A.; Sekimoto, H.

2010-06-01

In this study a fesibility design study of medium sized (1000 MWt) gas cooled fast reactors which can utilize natural uranium as fuel cycle input has been conducted. Gas Cooled Fast Reactor (GFR) is among six types of Generation IV Nuclear Power Plants. GFR with its hard neuron spectrum is superior for closed fuel cycle, and its ability to be operated in high temperature (850° C) makes various options of utilizations become possible. To obtain the capability of consuming natural uranium as fuel cycle input, modified CANDLE burn-up scheme[1-6] is adopted this GFR system by dividing the core into 10 parts of equal volume axially. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. As an optimization results, a design of 1000 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input is discussed. The average discharge burn-up is about 280 GWd/ton HM. Enough margin for criticallity was obtained for this reactor.

Dynamic analysis of gas-core reactor system

NASA Technical Reports Server (NTRS)

Turner, K. H., Jr.

1973-01-01

A heat transfer analysis was incorporated into a previously developed model CODYN to obtain a model of open-cycle gaseous core reactor dynamics which can predict the heat flux at the cavity wall. The resulting model was used to study the sensitivity of the model to the value of the reactivity coefficients and to determine the system response for twenty specified perturbations. In addition, the model was used to study the effectiveness of several control systems in controlling the reactor. It was concluded that control drums located in the moderator region capable of inserting reactivity quickly provided the best control.

Apparatus and process to eliminate diffusional limitations in a membrane biological reactor by pressure cycling

DOEpatents

Efthymiou, George S.; Shuler, Michael L.

1989-08-29

An improved multilayer continuous biological membrane reactor and a process to eliminate diffusional limitations in membrane reactors in achieved by causing a convective flux of nutrient to move into and out of an immobilized biocatalyst cell layer. In a pressure cycled mode, by increasing and decreasing the pressure in the respective layers, the differential pressure between the gaseous layer and the nutrient layer is alternately changed from positive to negative. The intermittent change in pressure differential accelerates the transfer of nutrient from the nutrient layers to the biocatalyst cell layer, the transfer of product from the cell layer to the nutrient layer and the transfer of byproduct gas from the cell layer to the gaseous layer. Such intermittent cycling substantially eliminates mass transfer gradients in diffusion inhibited systems and greatly increases product yield and throughput in both inhibited and noninhibited systems.

Supercritical Brayton Cycle Nuclear Power System Concepts

NASA Astrophysics Data System (ADS)

Wright, Steven A.

2007-01-01

Both the NASA and DOE have programs that are investigating advanced power conversion cycles for planetary surface power on the moon or Mars, and for next generation nuclear power plants on earth. The gas Brayton cycle offers many practical solutions for space nuclear power systems and was selected as the nuclear power system of choice for the NASA Prometheus project. An alternative Brayton cycle that offers high efficiency at a lower reactor coolant outlet temperature is the supercritical Brayton cycle (SCBC). The supercritical cycle is a true Brayton cycle because it uses a single phase fluid with a compressor inlet temperature that is just above the critical point of the fluid. This paper describes the use of a supercritical Brayton cycle that achieves a cycle efficiency of 26.6% with a peak coolant temperature of 750 K and for a compressor inlet temperature of 390 K. The working fluid uses a clear odorless, nontoxic refrigerant C318 perflurocarbon (C4F8) that always operates in the gas phase. This coolant was selected because it has a critical temperature and pressure of 388.38 K and 2.777 MPa. The relatively high critical temperature allows for efficient thermal radiation that keeps the radiator mass small. The SCBC achieves high efficiency because the loop design takes advantage of the non-ideal nature of the coolant equation of state just above the critical point. The lower coolant temperature means that metal fuels, uranium oxide fuels, and uranium zirconium hydride fuels with stainless steel, ferretic steel, or superalloy cladding can be used with little mass penalty or reduction in cycle efficiency. The reactor can use liquid-metal coolants and no high temperature heat exchangers need to be developed. Indirect gas cooling or perhaps even direct gas cooling can be used if the C4F8 coolant is found to be sufficiently radiation tolerant. Other fluids can also be used in the supercritical Brayton cycle including Propane (C3H8, Tcritical = 369 K) and Hexane (C6

Fuel Cycle System Analysis Handbook

DOE Office of Scientific and Technical Information (OSTI.GOV)

Steven J. Piet; Brent W. Dixon; Dirk Gombert

2009-06-01

This Handbook aims to improve understanding and communication regarding nuclear fuel cycle options. It is intended to assist DOE, Campaign Managers, and other presenters prepare presentations and reports. When looking for information, check here. The Handbook generally includes few details of how calculations were performed, which can be found by consulting references provided to the reader. The Handbook emphasizes results in the form of graphics and diagrams, with only enough text to explain the graphic, to ensure that the messages associated with the graphic is clear, and to explain key assumptions and methods that cause the graphed results. Some ofmore » the material is new and is not found in previous reports, for example: (1) Section 3 has system-level mass flow diagrams for 0-tier (once-through), 1-tier (UOX to CR=0.50 fast reactor), and 2-tier (UOX to MOX-Pu to CR=0.50 fast reactor) scenarios - at both static and dynamic equilibrium. (2) To help inform fast reactor transuranic (TRU) conversion ratio and uranium supply behavior, section 5 provides the sustainable fast reactor growth rate as a function of TRU conversion ratio. (3) To help clarify the difference in recycling Pu, NpPu, NpPuAm, and all-TRU, section 5 provides mass fraction, gamma, and neutron emission for those four cases for MOX, heterogeneous LWR IMF (assemblies mixing IMF and UOX pins), and a CR=0.50 fast reactor. There are data for the first 10 LWR recycle passes and equilibrium. (4) Section 6 provides information on the cycle length, planned and unplanned outages, and TRU enrichment as a function of fast reactor TRU conversion ratio, as well as the dilution of TRU feedstock by uranium in making fast reactor fuel. (The recovered uranium is considered to be more pure than recovered TRU.) The latter parameter impacts the required TRU impurity limits specified by the Fuels Campaign. (5) Section 7 provides flows for an 800-tonne UOX separation plant. (6) To complement 'tornado' economic

Evaluation of isotopic composition of fast reactor core in closed nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Tikhomirov, Georgy; Ternovykh, Mikhail; Saldikov, Ivan; Fomichenko, Peter; Gerasimov, Alexander

2017-09-01

The strategy of the development of nuclear power in Russia provides for use of fast power reactors in closed nuclear fuel cycle. The PRORYV (i.e. «Breakthrough» in Russian) project is currently under development. Within the framework of this project, fast reactors BN-1200 and BREST-OD-300 should be built to, inter alia, demonstrate possibility of the closed nuclear fuel cycle technologies with plutonium as a main source of energy. Russia has a large inventory of plutonium which was accumulated in the result of reprocessing of spent fuel of thermal power reactors and conversion of nuclear weapons. This kind of plutonium will be used for development of initial fuel assemblies for fast reactors. The closed nuclear fuel cycle concept of the PRORYV assumes self-supplied mode of operation with fuel regeneration by neutron capture reaction in non-enriched uranium, which is used as a raw material. Operating modes of reactors and its characteristics should be chosen so as to provide the self-sufficient mode by using of fissile isotopes while refueling by depleted uranium and to support this state during the entire period of reactor operation. Thus, the actual issue is modeling fuel handling processes. To solve these problems, the code REPRORYV (Recycle for PRORYV) has been developed. It simulates nuclide streams in non-reactor stages of the closed fuel cycle. At the same time various verified codes can be used to evaluate in-core characteristics of a reactor. By using this approach various options for nuclide streams and assess the impact of different plutonium content in the fuel, fuel processing conditions, losses during fuel processing, as well as the impact of initial uncertainties on neutron-physical characteristics of reactor are considered in this study.

Coupling a Supercritical Carbon Dioxide Brayton Cycle to a Helium-Cooled Reactor.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Middleton, Bobby; Pasch, James Jay; Kruizenga, Alan Michael

2016-01-01

This report outlines the thermodynamics of a supercritical carbon dioxide (sCO 2) recompression closed Brayton cycle (RCBC) coupled to a Helium-cooled nuclear reactor. The baseline reactor design for the study is the AREVA High Temperature Gas-Cooled Reactor (HTGR). Using the AREVA HTGR nominal operating parameters, an initial thermodynamic study was performed using Sandia's deterministic RCBC analysis program. Utilizing the output of the RCBC thermodynamic analysis, preliminary values of reactor power and of Helium flow rate through the reactor were calculated in Sandia's HelCO 2 code. Some research regarding materials requirements was then conducted to determine aspects of corrosion related tomore » both Helium and to sCO 2 , as well as some mechanical considerations for pressures and temperatures that will be seen by the piping and other components. This analysis resulted in a list of materials-related research items that need to be conducted in the future. A short assessment of dry heat rejection advantages of sCO 2> Brayton cycles was also included. This assessment lists some items that should be investigated in the future to better understand how sCO 2 Brayton cycles and nuclear can maximally contribute to optimizing the water efficiency of carbon free power generation« less

Core design of a direct-cycle, supercritical-water-cooled fast breeder reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jevremovic, T.; Oka, Yoshiaki; Koshizuka, Seiichi

1994-10-01

The conceptual design of a direct-cycle fast breeder reactor (FBR) core cooled by supercritical water is carried out as a step toward a low-cost FBR plant. The supercritical water does not exhibit change of phase. The turbines are directly driven by the core outlet coolant. In comparison with a boiling water reactor (BWR), the recirculation systems, steam separators, and dryers are eliminated. The reactor system is much simpler than the conventional steam-cooled FBRs, which adopted Loeffler boilers and complicated coolant loops for generating steam and separating it from water. Negative complete and partial coolant void reactivity are provided without muchmore » deterioration in the breeding performances by inserting thin zirconium-hydride layers between the seeds and blankets in a radially heterogeneous core. The net electric power is 1245 MW (electric). The estimated compound system doubling time is 25 yr. The discharge burnup is 77.7 GWd/t, and the refueling period is 15 months with a 73% load factor. The thermal efficiency is high (41.5%), an improvement of 24% relative to a BWR's. The pressure vessel is not thick at 30.3 cm.« less

75 FR 51025 - Blue Ribbon Commission on America's Nuclear Future, Reactor and Fuel Cycle Technology Subcommittee

Federal Register 2010, 2011, 2012, 2013, 2014

2010-08-18

... DEPARTMENT OF ENERGY Blue Ribbon Commission on America's Nuclear Future, Reactor and Fuel Cycle... meeting. SUMMARY: This notice announces an open meeting of the Reactor and Fuel Cycle Technology (RFCT... back end of the nuclear fuel cycle. The Commission will provide advice and make recommendations on...

The CANDU Reactor System: An Appropriate Technology.

PubMed

Robertson, J A

1978-02-10

CANDU power reactors are characterized by the combination of heavy water as moderator and pressure tubes to contain the fuel and coolant. Their excellent neutron economy provides the simplicity and low costs of once-through natural-uranium fueling. Future benefits include the prospect of a near-breeder thorium fuel cycle to provide security of fuel supply without the need to develop a new reactor such as the fast breeder. These and other features make the CANDU system an appropriate technology for countries, like Canada, of intermediate economic and industrial capacity.

Long lifetime fast spectrum reactor for lunar surface power system

NASA Astrophysics Data System (ADS)

Kambe, Mitsuru

1993-01-01

In the framework of innovative reactor research activities, a conceptual design study of fast spectrum reactor and primary system for 800 kWe lunar surface power system to be combined with potassium Rankine cycle power conversion has been conducted to meet the power requirements of the lunar base activities in the next century. The reactor subsystem is characterized by RAPID (Refueling by All Pins Integrated Design) concept to enhance inherent safety and to enable quick and simplifed refueling in every 10 years. RAPID concept affords power plant design lifetime of up to 30 years. Integrity of the reactor structure and replacement of failed primary circuits are also discussed. Substantial reduction in per-kWh cost on considering launch, emplacement, and final disposition can be expected by a long system lifetime.

Hybrid sulfur cycle operation for high-temperature gas-cooled reactors

DOEpatents

Gorensek, Maximilian B

2015-02-17

A hybrid sulfur (HyS) cycle process for the production of hydrogen is provided. The process uses a proton exchange membrane (PEM) SO.sub.2-depolarized electrolyzer (SDE) for the low-temperature, electrochemical reaction step and a bayonet reactor for the high-temperature decomposition step The process can be operated at lower temperature and pressure ranges while still providing an overall energy efficient cycle process.

A Burst Mode, Ultrahigh Temperature UF4 Vapor Core Reactor Rankine Cycle Space Power System Concept

NASA Technical Reports Server (NTRS)

Dugan, E. T.; Kahook, S. D.; Diaz, N. J.

1996-01-01

Static and dynamic neutronic analyses have been performed on an innovative burst mode (100's of MW output for a few thousand seconds) Ulvahigh Temperature Vapor Core Reactor (UTVR) space nuclear power system. The NVTR employs multiple, neutronically-coupled fissioning cores and operates on a direct, closed Rankine cycle using a disk Magnetohydrodynamic (MHD) generater for energy conversion. The UTVR includes two types of fissioning core regions: (1) the central Ultrahigh Temperature Vapor Core (UTVC) which contains a vapor mixture of highly enriched UF4 fuel and a metal fluoride working fluid and (2) the UF4 boiler column cores located in the BeO moderator/reflector region. The gaseous nature of the fuel the fact that the fuel is circulating, the multiple coupled fissioning cores, and the use of a two phase fissioning fuel lead to unique static and dynamic neutronic characteristics. Static neutronic analysis was conducted using two-dimensional S sub n, transport theory calculations and three-dimensional Monte Carlo transport theory calculations. Circulating-fuel, coupled-core point reactor kinetics equations were used for analyzing the dynamic behavior of the UTVR. In addition to including reactivity feedback phenomena associated with the individual fissioning cores, the effects of core-to-core neutronic and mass flow coupling between the UTVC and the surrounding boiler cores were also included in the dynamic model The dynamic analysis of the UTVR reveals the existence of some very effectlve inherent reactivity feedback effects that are capable of quickly stabilizing this system, within a few seconds, even when large positive reactivity insertions are imposed. If the UTVC vapor fuel density feedback is suppressed, the UTVR is still inherently stable because of the boiler core liquid-fuel volume feedback; in contrast, suppression of the vapor fuel density feedback in 'conventional" gas core cavity reactors causes them to become inherently unstable. Due to the

The combined hybrid system: A symbiotic thermal reactor/fast reactor system for power generation and radioactive waste toxicity reduction

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hollaway, W.R.

1991-08-01

If there is to be a next generation of nuclear power in the United States, then the four fundamental obstacles confronting nuclear power technology must be overcome: safety, cost, waste management, and proliferation resistance. The Combined Hybrid System (CHS) is proposed as a possible solution to the problems preventing a vigorous resurgence of nuclear power. The CHS combines Thermal Reactors (for operability, safety, and cost) and Integral Fast Reactors (for waste treatment and actinide burning) in a symbiotic large scale system. The CHS addresses the safety and cost issues through the use of advanced reactor designs, the waste management issuemore » through the use of actinide burning, and the proliferation resistance issue through the use of an integral fuel cycle with co-located components. There are nine major components in the Combined Hybrid System linked by nineteen nuclear material mass flow streams. A computer code, CHASM, is used to analyze the mass flow rates CHS, and the reactor support ratio (the ratio of thermal/fast reactors), IFR of the system. The primary advantages of the CHS are its essentially actinide-free high-level radioactive waste, plus improved reactor safety, uranium utilization, and widening of the option base. The primary disadvantages of the CHS are the large capacity of IFRs required (approximately one MW{sub e} IFR capacity for every three MW{sub e} Thermal Reactor) and the novel radioactive waste streams produced by the CHS. The capability of the IFR to burn pure transuranic fuel, a primary assumption of this study, has yet to be proven. The Combined Hybrid System represents an attractive option for future nuclear power development; that disposal of the essentially actinide-free radioactive waste produced by the CHS provides an excellent alternative to the disposal of intact actinide-bearing Light Water Reactor spent fuel (reducing the toxicity based lifetime of the waste from roughly 360,000 years to about 510 years).« less

Challenges to deployment of twenty-first century nuclear reactor systems.

PubMed

Ion, Sue

2017-02-01

The science and engineering of materials have always been fundamental to the success of nuclear power to date. They are also the key to the successful deployment and operation of a new generation of nuclear reactor systems and their associated fuel cycles. This article reflects on some of the historical issues, the challenges still prevalent today and the requirement for significant ongoing materials R&D and discusses the potential role of small modular reactors.

The benefits of a fast reactor closed fuel cycle in the UK

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gregg, R.; Hesketh, K.

2013-07-01

The work has shown that starting a fast reactor closed fuel cycle in the UK, requires virtually all of Britain's existing and future PWR spent fuel to be reprocessed, in order to obtain the plutonium needed. The existing UK Pu stockpile is sufficient to initially support only a modest SFR 'closed' fleet assuming spent fuel can be reprocessed shortly after discharge (i.e. after two years cooling). For a substantial fast reactor fleet, most Pu will have to originate from reprocessing future spent PWR fuel. Therefore, the maximum fast reactor fleet size will be limited by the preceding PWR fleet size,more » so scenarios involving fast reactors still require significant quantities of uranium ore indirectly. However, once a fast reactor fuel cycle has been established, the very substantial quantities of uranium tails in the UK would ensure there is sufficient material for several centuries. Both the short and long term impacts on a repository have been considered in this work. Over the short term, the decay heat emanating from the HLW and spent fuel will limit the density of waste within a repository. For scenarios involving fast reactors, the only significant heat bearing actinide content will be present in the final cores, resulting in a 50% overall reduction in decay energy deposited within the repository when compared with an equivalent open fuel cycle. Over the longer term, radiological dose becomes more important. Total radiotoxicity (normalised by electricity generated) is lower for scenarios with Pu recycle after 2000 years. Scenarios involving fast reactors have the lowest radiotoxicity since the quantities of certain actinides (Np, Pu and Am) eventually stabilise. However, total radiotoxicity as a measure of radiological risk does not account for differences in radionuclide mobility once in repository. Radiological dose is dominated by a small number of fission products so is therefore not affected significantly by reactor type or recycling strategy (since

Challenges to deployment of twenty-first century nuclear reactor systems

PubMed Central

2017-01-01

The science and engineering of materials have always been fundamental to the success of nuclear power to date. They are also the key to the successful deployment and operation of a new generation of nuclear reactor systems and their associated fuel cycles. This article reflects on some of the historical issues, the challenges still prevalent today and the requirement for significant ongoing materials R&D and discusses the potential role of small modular reactors. PMID:28293142

Advanced Multi-Effect Distillation System for Desalination Using Waste Heat fromGas Brayton Cycles

DOE Office of Scientific and Technical Information (OSTI.GOV)

Haihua Zhao; Per F. Peterson

2012-10-01

Generation IV high temperature reactor systems use closed gas Brayton Cycles to realize high thermal efficiency in the range of 40% to 60%. The waste heat is removed through coolers by water at substantially greater average temperature than in conventional Rankine steam cycles. This paper introduces an innovative Advanced Multi-Effect Distillation (AMED) design that can enable the production of substantial quantities of low-cost desalinated water using waste heat from closed gas Brayton cycles. A reference AMED design configuration, optimization models, and simplified economics analysis are presented. By using an AMED distillation system the waste heat from closed gas Brayton cyclesmore » can be fully utilized to desalinate brackish water and seawater without affecting the cycle thermal efficiency. Analysis shows that cogeneration of electricity and desalinated water can increase net revenues for several Brayton cycles while generating large quantities of potable water. The AMED combining with closed gas Brayton cycles could significantly improve the sustainability and economics of Generation IV high temperature reactors.« less

Scale-4 Analysis of Pressurized Water Reactor Critical Configurations: Volume 2-Sequoyah Unit 2 Cycle 3

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bowman, S.M.

1995-01-01

The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit for the negative reactivity of the depleted (or spent) fuel isotopics is desired, it is necessary to benchmark computational methods against spent fuel critical configurations. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using critical configurations from commercial pressurized-water reactors. The analysis methodology selected for all the calculations reported herein is based on the codes and data provided in the SCALE-4 code system. The isotopic densities for the spent fuel assemblies inmore » the critical configurations were calculated using the SAS2H analytical sequence of the SCALE-4 system. The sources of data and the procedures for deriving SAS2H input parameters are described in detail. The SNIKR code module was used to extract the necessary isotopic densities from the SAS2H results and provide the data in the format required by the SCALE criticality analysis modules. The CSASN analytical sequence in SCALE-4 was used to perform resonance processing of the cross sections. The KENO V.a module of SCALE-4 was used to calculate the effective multiplication factor (k{sub eff}) of each case. The SCALE-4 27-group burnup library containing ENDF/B-IV (actinides) and ENDF/B-V (fission products) data was used for all the calculations. This volume of the report documents the SCALE system analysis of three reactor critical configurations for the Sequoyah Unit 2 Cycle 3. This unit and cycle were chosen because of the relevance in spent fuel benchmark applications: (1) the unit had a significantly long downtime of 2.7 years during the middle of cycle (MOC) 3, and (2) the core consisted entirely of burned fuel at the MOC restart. The first benchmark critical calculation was the MOC restart at hot, full-power (HFP) critical

The scheme for evaluation of isotopic composition of fast reactor core in closed nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Saldikov, I. S.; Ternovykh, M. Yu; Fomichenko, P. A.; Gerasimov, A. S.

2017-01-01

The PRORYV (i.e. «Breakthrough» in Russian) project is currently under development. Within the framework of this project, fast reactors BN-1200 and BREST-OD-300 should be built to, inter alia, demonstrate possibility of the closed nuclear fuel cycle technologies with plutonium as a main source of power. Russia has a large inventory of plutonium which was accumulated in the result of reprocessing of spent fuel of thermal power reactors and conversion of nuclear weapons. This kind of plutonium will be used for development of initial fuel assemblies for fast reactors. To solve the closed nuclear fuel modeling tasks REPRORYV code was developed. It simulates the mass flow for nuclides in the closed fuel cycle. This paper presents the results of modeling of a closed nuclear fuel cycle, nuclide flows considering the influence of the uncertainty on the outcome of neutron-physical characteristics of the reactor.

DIRECT-CYCLE, BOILING-WATER NUCLEAR REACTOR

DOEpatents

Harrer, J.M.; Fromm, L.W. Jr.; Kolba, V.M.

1962-08-14

A direct-cycle boiling-water nuclear reactor is described that employs a closed vessel and a plurality of fuel assemblies, each comprising an outer tube closed at its lower end, an inner tube, fuel rods in the space between the tubes and within the inner tube. A body of water lying within the pressure vessel and outside the fuel assemblies is converted to saturated steam, which enters each fuel assembly at the top and is converted to superheated steam in the fuel assembly while it is passing therethrough first downward through the space between the inner and outer tubes of the fuel assembly and then upward through the inner tube. (AEC)

Dynamic Modeling and Control of Nuclear Reactors Coupled to Closed-Loop Brayton Cycle Systems using SIMULINK{sup TM}

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wright, Steven A.; Sanchez, Travis

2005-02-06

The operation of space reactors for both in-space and planetary operations will require unprecedented levels of autonomy and control. Development of these autonomous control systems will require dynamic system models, effective control methodologies, and autonomous control logic. This paper briefly describes the results of reactor, power-conversion, and control models that are implemented in SIMULINK{sup TM} (Simulink, 2004). SIMULINK{sup TM} is a development environment packaged with MatLab{sup TM} (MatLab, 2004) that allows the creation of dynamic state flow models. Simulation modules for liquid metal, gas cooled reactors, and electrically heated systems have been developed, as have modules for dynamic power-conversion componentsmore » such as, ducting, heat exchangers, turbines, compressors, permanent magnet alternators, and load resistors. Various control modules for the reactor and the power-conversion shaft speed have also been developed and simulated. The modules are compiled into libraries and can be easily connected in different ways to explore the operational space of a number of potential reactor, power-conversion system configurations, and control approaches. The modularity and variability of these SIMULINK{sup TM} models provides a way to simulate a variety of complete power generation systems. To date, both Liquid Metal Reactors (LMR), Gas Cooled Reactors (GCR), and electric heaters that are coupled to gas-dynamics systems and thermoelectric systems have been simulated and are used to understand the behavior of these systems. Current efforts are focused on improving the fidelity of the existing SIMULINK{sup TM} modules, extending them to include isotopic heaters, heat pipes, Stirling engines, and on developing state flow logic to provide intelligent autonomy. The simulation code is called RPC-SIM (Reactor Power and Control-Simulator)« less

Hybrid fusion reactor for production of nuclear fuel with minimum radioactive contamination of the fuel cycle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Velikhov, E. P.; Kovalchuk, M. V.; Azizov, E. A., E-mail: Azizov-EA@nrcki.ru

2015-12-15

The paper presents the results of the system research on the coordinated development of nuclear and fusion power engineering in the current century. Considering the increasing problems of resource procurement, including limited natural uranium resources, it seems reasonable to use fusion reactors as high-power neutron sources for production of nuclear fuel in a blanket. It is shown that the share of fusion sources in this structural configuration of the energy system can be relatively small. A fundamentally important aspect of this solution to the problem of closure of the fuel cycle is that recycling of highly active spent fuel canmore » be abandoned. Radioactivity released during the recycling of the spent fuel from the hybrid reactor blanket is at least two orders of magnitude lower than during the production of the same number of fissile isotopes after the recycling of the spent fuel from a fast reactor.« less

Hybrid fusion-fission reactor with a thorium blanket: Its potential in the fuel cycle of nuclear reactors

NASA Astrophysics Data System (ADS)

Shmelev, A. N.; Kulikov, G. G.; Kurnaev, V. A.; Salahutdinov, G. H.; Kulikov, E. G.; Apse, V. A.

2015-12-01

Discussions are currently going on as to whether it is suitable to employ thorium in the nuclear fuel cycle. This work demonstrates that the 231Pa-232U-233U-Th composition to be produced in the thorium blanket of a hybrid thermonuclear reactor (HTR) as a fuel for light-water reactors opens up the possibility of achieving high, up to 30% of heavy metals (HM), or even ultrahigh fuel burnup. This is because the above fuel composition is able to stabilize its neutron-multiplying properties in the process of high fuel burnup. In addition, it allows the nuclear fuel cycle (NFC) to be better protected against unauthorized proliferation of fissile materials owing to an unprecedentedly large fraction of 232U (several percent!) in the uranium bred from the Th blanket, which will substantially hamper the use of fissile materials in a closed NFC for purposes other than power production.

Hybrid fusion–fission reactor with a thorium blanket: Its potential in the fuel cycle of nuclear reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shmelev, A. N., E-mail: shmelan@mail.ru; Kulikov, G. G., E-mail: ggkulikov@mephi.ru; Kurnaev, V. A., E-mail: kurnaev@yandex.ru

2015-12-15

Discussions are currently going on as to whether it is suitable to employ thorium in the nuclear fuel cycle. This work demonstrates that the {sup 231}Pa–{sup 232}U–{sup 233}U–Th composition to be produced in the thorium blanket of a hybrid thermonuclear reactor (HTR) as a fuel for light-water reactors opens up the possibility of achieving high, up to 30% of heavy metals (HM), or even ultrahigh fuel burnup. This is because the above fuel composition is able to stabilize its neutron-multiplying properties in the process of high fuel burnup. In addition, it allows the nuclear fuel cycle (NFC) to be bettermore » protected against unauthorized proliferation of fissile materials owing to an unprecedentedly large fraction of {sup 232}U (several percent!) in the uranium bred from the Th blanket, which will substantially hamper the use of fissile materials in a closed NFC for purposes other than power production.« less

System Analysis for Decay Heat Removal in Lead-Bismuth Cooled Natural Circulated Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Takaaki Sakai; Yasuhiro Enuma; Takashi Iwasaki

2002-07-01

Decay heat removal analyses for lead-bismuth cooled natural circulation reactors are described in this paper. A combined multi-dimensional plant dynamics code (MSG-COPD) has been developed to conduct the system analysis for the natural circulation reactors. For the preliminary study, transient analysis has been performed for a 100 MWe lead-bismuth-cooled reactor designed by Argonne National Laboratory (ANL). In addition, decay heat removal characteristics of a 400 MWe lead-bismuth-cooled natural circulation reactor designed by Japan Nuclear Cycle Development Institute (JNC) has been evaluated by using MSG-COPD. PRACS (Primary Reactor Auxiliary Cooling System) is prepared for the JNC's concept to get sufficient heatmore » removal capacity. During 2000 sec after the transient, the outlet temperature shows increasing tendency up to the maximum temperature of 430 Centigrade, because the buoyancy force in a primary circulation path is temporary reduced. However, the natural circulation is recovered by the PRACS system and the out let temperature decreases successfully. (authors)« less

System Analysis for Decay Heat Removal in Lead-Bismuth-Cooled Natural-Circulation Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sakai, Takaaki; Enuma, Yasuhiro; Iwasaki, Takashi

2004-03-15

Decay heat removal analyses for lead-bismuth-cooled natural-circulation reactors are described in this paper. A combined multidimensional plant dynamics code (MSG-COPD) has been developed to conduct the system analysis for the natural-circulation reactors. For the preliminary study, transient analysis has been performed for a 300-MW(thermal) lead-bismuth-cooled reactor designed by Argonne National Laboratory. In addition, decay heat removal characteristics of a 400-MW(electric) lead-bismuth-cooled natural-circulation reactor designed by the Japan Nuclear Cycle Development Institute (JNC) has been evaluated by using MSG-COPD. The primary reactor auxiliary cooling system (PRACS) is prepared for the JNC concept to get sufficient heat removal capacity. During 2000 smore » after the transient, the outlet temperature shows increasing tendency up to the maximum temperature of 430 deg. C because the buoyancy force in a primary circulation path is temporarily reduced. However, the natural circulation is recovered by the PRACS system, and the outlet temperature decreases successfully.« less

The effectiveness of using the combined-cycle technology in a nuclear power plant unit equipped with an SVBR-100 reactor

NASA Astrophysics Data System (ADS)

Kasilov, V. F.; Dudolin, A. A.; Gospodchenkov, I. V.

2015-05-01

The design of a modular SVBR-100 reactor with a lead-bismuth alloy liquid-metal coolant is described. The basic thermal circuit of a power unit built around the SVBR-100 reactor is presented together with the results of its calculation. The gross electrical efficiency of the turbine unit driven by saturated steam at a pressure of 6.7 MPa is estimated at η{el/gr} = 35.5%. Ways for improving the efficiency of this power unit and increasing its power output by applying gas-turbine and combined-cycle technologies are considered. With implementing a combined-cycle power-generating system comprising two GE-6101FA gas-turbine units with a total capacity of 140 MW, it becomes possible to obtain the efficiency of the combined-cycle plant equipped with the SVBR-100 reactor η{el/gr} = 45.39% and its electrical power output equal to 328 MW. The heat-recovery boiler used as part of this power installation generates superheated steam with a temperature of 560°C, due to which there is no need to use a moisture separator/steam reheater in the turbine unit thermal circuit.

HORIZONTAL BOILING REACTOR SYSTEM

DOEpatents

Treshow, M.

1958-11-18

Reactors of the boiling water type are described wherein water serves both as the moderator and coolant. The reactor system consists essentially of a horizontal pressure vessel divided into two compartments by a weir, a thermal neutronic reactor core having vertical coolant passages and designed to use water as a moderator-coolant posltioned in one compartment, means for removing live steam from the other compartment and means for conveying feed-water and water from the steam compartment to the reactor compartment. The system further includes auxiliary apparatus to utilize the steam for driving a turbine and returning the condensate to the feed-water inlet of the reactor. The entire system is designed so that the reactor is self-regulating and has self-limiting power and self-limiting pressure features.

The study of capability natural uranium as fuel cycle input for long life gas cooled fast reactors with helium as coolant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ariani, Menik, E-mail: menikariani@gmail.com; Satya, Octavianus Cakra; Monado, Fiber

The objective of the present research is to assess the feasibility design of small long-life Gas Cooled Fast Reactor with helium as coolant. GCFR included in the Generation-IV reactor systems are being developed to provide sustainable energy resources that meet future energy demand in a reliable, safe, and proliferation-resistant manner. This reactor can be operated without enrichment and reprocessing forever, once it starts. To obtain the capability of consuming natural uranium as fuel cycle input modified CANDLE burn-up scheme was adopted in this system with different core design. This study has compared the core with three designs of core reactorsmore » with the same thermal power 600 MWth. The fuel composition each design was arranged by divided core into several parts of equal volume axially i.e. 6, 8 and 10 parts related to material burn-up history. The fresh natural uranium is initially put in region 1, after one cycle of 10 years of burn-up it is shifted to region 2 and the region 1 is filled by fresh natural uranium fuel. This concept is basically applied to all regions, i.e. shifted the core of the region (i) into region (i+1) region after the end of 10 years burn-up cycle. The calculation results shows that for the burn-up strategy on “Region-8” and “Region-10” core designs, after the reactors start-up the operation furthermore they only needs natural uranium supply to the next life operation until one period of refueling (10 years).« less

Reactor Operations Monitoring System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hart, M.M.

1989-01-01

The Reactor Operations Monitoring System (ROMS) is a VME based, parallel processor data acquisition and safety action system designed by the Equipment Engineering Section and Reactor Engineering Department of the Savannah River Site. The ROMS will be analyzing over 8 million signal samples per minute. Sixty-eight microprocessors are used in the ROMS in order to achieve a real-time data analysis. The ROMS is composed of multiple computer subsystems. Four redundant computer subsystems monitor 600 temperatures with 2400 thermocouples. Two computer subsystems share the monitoring of 600 reactor coolant flows. Additional computer subsystems are dedicated to monitoring 400 signals from assortedmore » process sensors. Data from these computer subsystems are transferred to two redundant process display computer subsystems which present process information to reactor operators and to reactor control computers. The ROMS is also designed to carry out safety functions based on its analysis of process data. The safety functions include initiating a reactor scram (shutdown), the injection of neutron poison, and the loadshed of selected equipment. A complete development Reactor Operations Monitoring System has been built. It is located in the Program Development Center at the Savannah River Site and is currently being used by the Reactor Engineering Department in software development. The Equipment Engineering Section is designing and fabricating the process interface hardware. Upon proof of hardware and design concept, orders will be placed for the final five systems located in the three reactor areas, the reactor training simulator, and the hardware maintenance center.« less

REACTOR CONTROL SYSTEM

DOEpatents

MacNeill, J.H.; Estabrook, J.Y.

1960-05-10

A reactor control system including a continuous tape passing through a first coolant passageway, over idler rollers, back through another parallel passageway, and over motor-driven rollers is described. Discrete portions of fuel or poison are carried on two opposed active sections of the tape. Driving the tape in forward or reverse directions causes both active sections to be simultaneously inserted or withdrawn uniformly, tending to maintain a more uniform flux within the reactor. The system is particularly useful in mobile reactors, where reduced inertial resistance to control rod movement is important.

Fuel cycle for a fusion neutron source

NASA Astrophysics Data System (ADS)

Ananyev, S. S.; Spitsyn, A. V.; Kuteev, B. V.

2015-12-01

The concept of a tokamak-based stationary fusion neutron source (FNS) for scientific research (neutron diffraction, etc.), tests of structural materials for future fusion reactors, nuclear waste transmutation, fission reactor fuel production, and control of subcritical nuclear systems (fusion-fission hybrid reactor) is being developed in Russia. The fuel cycle system is one of the most important systems of FNS that provides circulation and reprocessing of the deuterium-tritium fuel mixture in all fusion reactor systems: the vacuum chamber, neutral injection system, cryogenic pumps, tritium purification system, separation system, storage system, and tritium-breeding blanket. The existing technologies need to be significantly upgraded since the engineering solutions adopted in the ITER project can be only partially used in the FNS (considering the capacity factor higher than 0.3, tritium flow up to 200 m3Pa/s, and temperature of reactor elements up to 650°C). The deuterium-tritium fuel cycle of the stationary FNS is considered. The TC-FNS computer code developed for estimating the tritium distribution in the systems of FNS is described. The code calculates tritium flows and inventory in tokamak systems (vacuum chamber, cryogenic pumps, neutral injection system, fuel mixture purification system, isotope separation system, tritium storage system) and takes into account tritium loss in the fuel cycle due to thermonuclear burnup and β decay. For the two facility versions considered, FNS-ST and DEMO-FNS, the amount of fuel mixture needed for uninterrupted operation of all fuel cycle systems is 0.9 and 1.4 kg, consequently, and the tritium consumption is 0.3 and 1.8 kg per year, including 35 and 55 g/yr, respectively, due to tritium decay.

Morphological evolution of copper nanoparticles: Microemulsion reactor system versus batch reactor system

NASA Astrophysics Data System (ADS)

Xia, Ming; Tang, Zengmin; Kim, Woo-Sik; Yu, Taekyung; Park, Bum Jun

2017-07-01

In the synthesis of nanoparticles, the reaction rate is important to determine the morphology of nanoparticles. We investigated morphology evolution of Cu nanoparticles in this two different reactors, microemulsion reactor and batch reactor. In comparison with the batch reactor system, the enhanced mass and heat transfers in the emulsion system likely led to the relatively short nucleation time and the highly homogeneous environment in the reaction mixture, resulting in suppressing one or two dimensional growth of the nanoparticles. We believe that this work can offer a good model system to quantitatively understand the crystal growth mechanism that depends strongly on the local monomer concentration, the efficiency of heat transfer, and the relative contribution of the counter ions (Br- and Cl-) as capping agents.

A Novel Fuel/Reactor Cycle to Implement the 300 Years Nuclear Waste Policy Approach - 12377

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carelli, M.D.; Franceschini, F.; Lahoda, E.J.

2012-07-01

A thorium-based fuel cycle system can effectively burn the currently accumulated commercial used nuclear fuel and move to a sustainable equilibrium where the actinide levels in the high level waste are low enough to yield a radiotoxicity after 300 years lower than that of the equivalent uranium ore. The second step of the Westinghouse approach to solving the waste 'problem' has been completed. The thorium fuel cycle has indeed the potential of burning the legacy TRU and achieve the waste objective proposed. Initial evaluations have been started for the third step, development and selection of appropriate reactors. Indications are thatmore » the probability of show-stoppers is rather remote. It is, therefore, believed that development of the thorium cycle and associated technologies will provide a permanent solution to the waste management. Westinghouse is open to the widest collaboration to make this a reality. (authors)« less

Assessment for advanced fuel cycle options in CANDU

DOE Office of Scientific and Technical Information (OSTI.GOV)

Morreale, A.C.; Luxat, J.C.; Friedlander, Y.

2013-07-01

The possible options for advanced fuel cycles in CANDU reactors including actinide burning options and thorium cycles were explored and are feasible options to increase the efficiency of uranium utilization and help close the fuel cycle. The actinide burning TRUMOX approach uses a mixed oxide fuel of reprocessed transuranic actinides from PWR spent fuel blended with natural uranium in the CANDU-900 reactor. This system reduced actinide content by 35% and decreased natural uranium consumption by 24% over a PWR once through cycle. The thorium cycles evaluated used two CANDU-900 units, a generator and a burner unit along with a drivermore » fuel feedstock. The driver fuels included plutonium reprocessed from PWR, from CANDU and low enriched uranium (LEU). All three cycles were effective options and reduced natural uranium consumption over a PWR once through cycle. The LEU driven system saw the largest reduction with a 94% savings while the plutonium driven cycles achieved 75% savings for PWR and 87% for CANDU. The high neutron economy, online fuelling and flexible compact fuel make the CANDU system an ideal reactor platform for many advanced fuel cycles.« less

Tritium Mitigation/Control for Advanced Reactor System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sun, Xiaodong; Christensen, Richard; Saving, John P.

A tritium removal facility, which is similar to the design used for tritium recovery in fusion reactors, is proposed in this study for fluoride-salt-cooled high-temperature reactors (FHRs) to result in a two-loop FHR design with the elimination of an intermediate loop. Using this approach, an economic benefit can potentially be obtained by removing the intermediate loop, while the safety concern of tritium release can be mitigated. In addition, an intermediate heat exchanger (IHX) that can yield a similar tritium permeation rate to the production rate of 1.9 Ci/day in a 1,000 MWe PWR needs to be designed to prevent themore » residual tritium that is not captured in the tritium removal system from escaping into the power cycle and ultimately the environment. The main focus of this study is to aid the mitigation of tritium permeation issue from the FHR primary side to significantly reduce the concentration of tritium in the secondary side and the process heat application side (if applicable). The goal of the research is to propose a baseline FHR system without the intermediate loop. The specific objectives to accomplish the goals are: To estimate tritium permeation behavior in FHRs; To design a tritium removal system for FHRs; To meet the same tritium permeation level in FHRs as the tritium production rate of 1.9 Ci/day in 1,000 MWe PWRs; To demonstrate economic benefits of the proposed FHR system via comparing with the three-loop FHR system. The objectives were accomplished by designing tritium removal facilities, developing a tritium analysis code, and conducting an economic analysis. In the fusion reactor community, tritium extraction has been widely investigated and researched. Borrowing the experiences from the fusion reactor community, a tritium control and mitigation system was proposed. Based on mass transport theories, a tritium analysis code was developed, and the tritium behaviors were analyzed using the developed code. Tritium removal facilities were

Hybrid systems for transuranic waste transmutation in nuclear power reactors: state of the art and future prospects

NASA Astrophysics Data System (ADS)

Yurov, D. V.; Prikhod'ko, V. V.

2014-11-01

The features of subcritical hybrid systems (HSs) are discussed in the context of burning up transuranic wastes from the U-Pu nuclear fuel cycle. The advantages of HSs over conventional atomic reactors are considered, and fuel cycle closure alternatives using HSs and fast neutron reactors are comparatively evaluated. The advantages and disadvantages of two HS types with neutron sources (NSs) of widely different natures -- nuclear spallation in a heavy target by protons and nuclear fusion in magnetically confined plasma -- are discussed in detail. The strengths and weaknesses of HSs are examined, and demand for them for closing the U-Pu nuclear fuel cycle is assessed.

75 FR 36648 - Blue Ribbon Commission on America's Nuclear Future, Reactor and Fuel Cycle Technologies Subcommittee

Federal Register 2010, 2011, 2012, 2013, 2014

2010-06-28

... DEPARTMENT OF ENERGY Blue Ribbon Commission on America's Nuclear Future, Reactor and Fuel Cycle Technologies Subcommittee AGENCY: Office of Nuclear Energy, DOE. ACTION: Notice of open meeting correction. On June 21, 2010, the Department of Energy published a notice announcing an open meeting of the Reactor...

Helium heater design for the helium direct cycle component test facility. [for gas-cooled nuclear reactor power plant

NASA Technical Reports Server (NTRS)

Larson, V. R.; Gunn, S. V.; Lee, J. C.

1975-01-01

The paper describes a helium heater to be used to conduct non-nuclear demonstration tests of the complete power conversion loop for a direct-cycle gas-cooled nuclear reactor power plant. Requirements for the heater include: heating the helium to a 1500 F temperature, operating at a 1000 psia helium pressure, providing a thermal response capability and helium volume similar to that of the nuclear reactor, and a total heater system helium pressure drop of not more than 15 psi. The unique compact heater system design proposed consists of 18 heater modules; air preheaters, compressors, and compressor drive systems; an integral control system; piping; and auxiliary equipment. The heater modules incorporate the dual-concentric-tube 'Variflux' heat exchanger design which provides a controlled heat flux along the entire length of the tube element. The heater design as proposed will meet all system requirements. The heater uses pressurized combustion (50 psia) to provide intensive heat transfer, and to minimize furnace volume and heat storage mass.

NEUTRONIC REACTOR SYSTEM

DOEpatents

Goett, J.J.

1961-01-24

A system is described which includes a neutronic reactor containing a dispersion of fissionable material in a liquid moderator as fuel and a conveyor to which a portion of the dispersion may be passed and wherein the self heat of the slurry evaporates the moderator. Means are provided for condensing the liquid moderator and returning it to the reactor and for conveying the dried fissionable material away from the reactor.

NEUTRONIC REACTOR SYSTEM

DOEpatents

Treshow, M.

1959-02-10

A reactor system incorporating a reactor of the heterogeneous boiling water type is described. The reactor is comprised essentially of a core submerged adwater in the lower half of a pressure vessel and two distribution rings connected to a source of water are disposed within the pressure vessel above the reactor core, the lower distribution ring being submerged adjacent to the uppcr end of the reactor core and the other distribution ring being located adjacent to the top of the pressure vessel. A feed-water control valve, responsive to the steam demand of the load, is provided in the feedwater line to the distribution rings and regulates the amount of feed water flowing to each distribution ring, the proportion of water flowing to the submerged distribution ring being proportional to the steam demand of the load. This invention provides an automatic means exterior to the reactor to control the reactivity of the reactor over relatively long periods of time without relying upon movement of control rods or of other moving parts within the reactor structure.

Impact of thorium based molten salt reactor on the closure of the nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Jaradat, Safwan Qasim Mohammad

Molten salt reactor (MSR) is one of six reactors selected by the Generation IV International Forum (GIF). The liquid fluoride thorium reactor (LFTR) is a MSR concept based on thorium fuel cycle. LFTR uses liquid fluoride salts as a nuclear fuel. It uses 232Th and 233U as the fertile and fissile materials, respectively. Fluoride salt of these nuclides is dissolved in a mixed carrier salt of lithium and beryllium (FLiBe). The objective of this research was to complete feasibility studies of a small commercial thermal LFTR. The focus was on neutronic calculations in order to prescribe core design parameter such as core size, fuel block pitch (p), fuel channel radius, fuel path, reflector thickness, fuel salt composition, and power. In order to achieve this objective, the applicability of Monte Carlo N-Particle Transport Code (MCNP) to MSR modeling was verified. Then, a prescription for conceptual small thermal reactor LFTR and relevant calculations were performed using MCNP to determine the main neutronic parameters of the core reactor. The MCNP code was used to study the reactor physics characteristics for the FUJI-U3 reactor. The results were then compared with the results obtained from the original FUJI-U3 using the reactor physics code SRAC95 and the burnup analysis code ORIPHY2. The results were comparable with each other. Based on the results, MCNP was found to be a reliable code to model a small thermal LFTR and study all the related reactor physics characteristics. The results of this study were promising and successful in demonstrating a prefatory small commercial LFTR design. The outcome of using a small core reactor with a diameter/height of 280/260 cm that would operate for more than five years at a power level of 150 MWth was studied. The fuel system 7LiF - BeF2 - ThF4 - UF4 with a (233U/ 232Th) = 2.01 % was the candidate fuel for this reactor core.

Optimization of power-cycle arrangements for Supercritical Water cooled Reactors (SCWRs)

NASA Astrophysics Data System (ADS)

Lizon-A-Lugrin, Laure

The world energy demand is continuously rising due to the increase of both the world population and the standard of life quality. Further, to assure both a healthy world economy as well as adequate social standards, in a relatively short term, new energy-conversion technologies are mandatory. Within this framework, a Generation IV International Forum (GIF) was established by the participation of 10 countries to collaborate for developing nuclear power reactors that will replace the present technology by 2030. The main goals of these nuclear-power reactors are: economic competitiveness, sustainability, safety, reliability and resistance to proliferation. As a member of the GIF, Canada has decided to orient its efforts towards the design of a CANDU-type Super Critical Water-cooled Reactor (SCWR). Such a system must run at a coolant outlet temperature of about 625°C and at a pressure of 25 MPa. It is obvious that at such conditions the overall efficiency of this kind of Nuclear Power Plant (NPP) will compete with actual supercritical water-power boilers. In addition, from a heat-transfer viewpoint, the use of a supercritical fluid allows the limitation imposed by Critical Heat Flux (CHF) conditions, which characterize actual technologies, to be removed. Furthermore, it will be also possible to use direct thermodynamic cycles where the supercritical fluid expands right away in a turbine without the necessity of using intermediate steam generators and/or separators. This work presents several thermodynamic cycles that could be appropriate to run SCWR power plants. Improving both thermal efficiency and mechanical power constitutes a multi-objective optimization problem and requires specific tools. To this aim, an efficient and robust evolutionary algorithm, based on genetic algorithm, is used and coupled to an appropriate power plant thermodynamic simulation model. The results provide numerous combinations to achieve a thermal efficiency higher than 50% with a

Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

2009-01-01

A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, OH. This is a closed-cycle system that incorporates an electrically heated reactor core module, turbo alternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

2010-01-01

A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, Ohio. This is a closed-cycle system that incorporates an electrically heated reactor core module, turboalternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

Thermochemical reactor systems and methods

DOEpatents

Lipinski, Wojciech; Davidson, Jane Holloway; Chase, Thomas Richard

2016-11-29

Thermochemical reactor systems that may be used to produce a fuel, and methods of using the thermochemical reactor systems, utilizing a reactive cylindrical element, an optional energy transfer cylindrical element, an inlet gas management system, and an outlet gas management system.

Immobilization of Fast Reactor First Cycle Raffinate

DOE Office of Scientific and Technical Information (OSTI.GOV)

Langley, K. F.; Partridge, B. A.; Wise, M.

This paper describes the results of work to bring forward the timing for the immobilization of first cycle raffinate from reprocessing fuel from the Dounreay Prototype Fast Reactor (PFR). First cycle raffinate is the liquor which contains > 99% of the fission products separated from spent fuel during reprocessing. Approximately 203 m3 of raffinate from the reprocessing of PFR fuel is held in four tanks at the UKAEA's site at Dounreay, Scotland. Two methods of immobilization of this high level waste (HLW) have been considered: vitrification and cementation. Vitrification is the standard industry practice for the immobilization of first cyclemore » raffinate, and many papers have been presented on this technique elsewhere. However, cementation is potentially feasible for immobilizing first cycle raffinate because the heat output is an order of magnitude lower than typical HLW from commercial reprocessing operations such as that at the Sellafield site in Cumbria, England. In fact, it falls within the upper end of the UK definition of intermediate level waste (ILW). Although the decision on which immobilization technique will be employed has yet to be made, initial development work has been undertaken to identify a suitable cementation formulation using inactive simulant of the raffinate. An approach has been made to the waste disposal company Nirex to consider the disposability of the cemented product material. The paper concentrates on the process development work that is being undertaken on cementation to inform the decision making process for selection of the immobilization method.« less

Reactor water cleanup system

DOEpatents

Gluntz, Douglas M.; Taft, William E.

1994-01-01

A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling.

Reactor water cleanup system

DOEpatents

Gluntz, D.M.; Taft, W.E.

1994-12-20

A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling. 1 figure.

NUCLEAR REACTOR CONTROL SYSTEM

DOEpatents

Epler, E.P.; Hanauer, S.H.; Oakes, L.C.

1959-11-01

A control system is described for a nuclear reactor using enriched uranium fuel of the type of the swimming pool and other heterogeneous nuclear reactors. Circuits are included for automatically removing and inserting the control rods during the course of normal operation. Appropriate safety circuits close down the nuclear reactor in the event of emergency.

A coupled nuclear reactor thermal energy storage system for enhanced load following operation

NASA Astrophysics Data System (ADS)

Alameri, Saeed A.

Nuclear power plants usually provide base-load electric power and operate most economically at a constant power level. In an energy grid with a high fraction of renewable energy sources, future nuclear reactors may be subject to significantly variable power demands. These variable power demands can negatively impact the effective capacity factor of the reactor and result in severe economic penalties. Coupling the reactor to a large Thermal Energy Storage (TES) block will allow the reactor to better respond to variable power demands. In the system described in this thesis, a Prismatic-core Advanced High Temperature Reactor (PAHTR) operates at constant power with heat provided to a TES block that supplies power as needed to a secondary energy conversion system. The PAHTR is designed to have a power rating of 300 MW th, with 19.75 wt% enriched Tri-Structural-Isotropic UO 2 fuel and a five year operating cycle. The passive molten salt TES system will operate in the latent heat region with an energy storage capacity of 150 MWd. Multiple smaller TES blocks are used instead of one large block to enhance the efficiency and maintenance complexity of the system. A transient model of the coupled reactor/TES system is developed to study the behavior of the system in response to varying load demands. The model uses six-delayed group point kinetics and decay heat models coupled to thermal-hydraulic and heat transfer models of the reactor and TES system. Based on the transient results, the preferred TES design consists of 1000 blocks, each containing 11000 LiCl phase change material tubes. A safety assessment of major reactor events demonstrates the inherent safety of the coupled system. The loss of forced circulation study determined the minimum required air convection heat removal rate from the reactor core and the lowest possible reduced primary flow rate that can maintain the reactor in a safe condition. The loss of ultimate heat sink study demonstrated the ability of the TES

Thermodynamic Analysis of the Use a Chemical Heat Pump to Link a Supercritical Water-Cooled Nuclear Reactor and a Thermochemical Water-Splitting Cycle for Hydrogen Production

NASA Astrophysics Data System (ADS)

Granovskii, Mikhail; Dincer, Ibrahim; Rosen, Marc A.; Pioro, Igor

of the combined system comprising a SCW nuclear power generation plant and a chemical heat pump, which provides high-temperature heat to a thermochemical water splitting cycle for hydrogen production. It is concluded that the proposed chemical heat pump permits the utilization efficiency of nuclear energy to be improved by at least 2% without jeopardizing nuclear reactor safety. Based on this analysis, further research appears to be merited on the proposed advanced design of a nuclear power generation plant combined with a chemical heat pump, and implementation in appropriate applications seems worthwhile.

The benefits of an advanced fast reactor fuel cycle for plutonium management

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hannum, W.H.; McFarlane, H.F.; Wade, D.C.

1996-12-31

The United States has no program to investigate advanced nuclear fuel cycles for the large-scale consumption of plutonium from military and civilian sources. The official U.S. position has been to focus on means to bury spent nuclear fuel from civilian reactors and to achieve the spent fuel standard for excess separated plutonium, which is considered by policy makers to be an urgent international priority. Recently, the National Research Council published a long awaited report on its study of potential separation and transmutation technologies (STATS), which concluded that in the nuclear energy phase-out scenario that they evaluated, transmutation of plutonium andmore » long-lived radioisotopes would not be worth the cost. However, at the American Nuclear Society Annual Meeting in June, 1996, the STATS panelists endorsed further study of partitioning to achieve superior waste forms for burial, and suggested that any further consideration of transmutation should be in the context of energy production, not of waste management. 2048 The U.S. Department of Energy (DOE) has an active program for the short-term disposition of excess fissile material and a `focus area` for safe, secure stabilization, storage and disposition of plutonium, but has no current programs for fast reactor development. Nevertheless, sufficient data exist to identify the potential advantages of an advanced fast reactor metallic fuel cycle for the long-term management of plutonium. Advantages are discussed.« less

Improved vortex reactor system

DOEpatents

Diebold, James P.; Scahill, John W.

1995-01-01

An improved vortex reactor system for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor.

Development of a thermal scheme for a cogeneration combined-cycle unit with an SVBR-100 reactor

NASA Astrophysics Data System (ADS)

Kasilov, V. F.; Dudolin, A. A.; Krasheninnikov, S. M.

2017-02-01

At present, the prospects for development of district heating that can increase the effectiveness of nuclear power stations (NPS), cut down their payback period, and improve protection of the environment against harmful emissions are being examined in the nuclear power industry of Russia. It is noted that the efficiency of nuclear cogeneration power stations (NCPS) is drastically affected by the expenses for heat networks and heat losses during transportation of a heat carrier through them, since NPSs are usually located far away from urban area boundaries as required for radiation safety of the population. The prospects for using cogeneration power units with small or medium power reactors at NPSs, including combined-cycle units and their performance indices, are described. The developed thermal scheme of a cogeneration combined-cycle unit (CCU) with an SBVR-100 nuclear reactor (NCCU) is presented. This NCCU should use a GE 6FA gasturbine unit (GTU) and a steam-turbine unit (STU) with a two-stage district heating plant. Saturated steam from the nuclear reactor is superheated in a heat-recovery steam generator (HRSG) to 560-580°C so that a separator-superheater can be excluded from the thermal cycle of the turbine unit. In addition, supplemental fuel firing in HRSG is examined. NCCU effectiveness indices are given as a function of the ambient air temperature. Results of calculations of the thermal cycle performance under condensing operating conditions indicate that the gross electric efficiency η el NCCU gr of = 48% and N el NCCU gr = 345 MW can be achieved. This efficiency is at maximum for NCCU with an SVBR-100 reactor. The conclusion is made that the cost of NCCU installed kW should be estimated, and the issue associated with NCCUs siting with reference to urban area boundaries must be solved.

Full reactor coolant system chemical decontamination qualification programs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Miller, P.E.

1995-03-01

Corrosion and wear products are found throughout the reactor coolant system (RCS), or primary loop, of a PWR power plant. These products circulate with the primary coolant through the reactor where they may become activated. An oxide layer including these activated products forms on the surfaces of the RCS (including the fuel elements). The amount of radioactivity deposited on the different surface varies and depends primarily on the corrosion rate of the materials concerned, the amount of cobalt in the coolant and the chemistry of the coolant. The oxide layer, commonly called crud, on the surfaces of nuclear plant systemsmore » leads to personnel radiation exposure. The level of the radiation fields from the crud increases with time from initial plant startup and typically levels off after 4 to 6 cycles of plant operation. Thereafter, significant personnel radiation exposure may be incurred whenever major maintenance is performed. Personnel exposure is highest during refueling outages when routine maintenance on major plant components, such as steam generators and reactor coolant pumps, is performed. Administrative controls are established at nuclear plants to minimize the exposure incurred by an individual and the plant workers as a whole.« less

Cycle Trades for Nuclear Thermal Rocket Propulsion Systems

NASA Technical Reports Server (NTRS)

White, C.; Guidos, M.; Greene, W.

2003-01-01

Nuclear fission has been used as a reliable source for utility power in the United States for decades. Even in the 1940's, long before the United States had a viable space program, the theoretical benefits of nuclear power as applied to space travel were being explored. These benefits include long-life operation and high performance, particularly in the form of vehicle power density, enabling longer-lasting space missions. The configurations for nuclear rocket systems and chemical rocket systems are similar except that a nuclear rocket utilizes a fission reactor as its heat source. This thermal energy can be utilized directly to heat propellants that are then accelerated through a nozzle to generate thrust or it can be used as part of an electricity generation system. The former approach is Nuclear Thermal Propulsion (NTP) and the latter is Nuclear Electric Propulsion (NEP), which is then used to power thruster technologies such as ion thrusters. This paper will explore a number of indirect-NTP engine cycle configurations using assumed performance constraints and requirements, discuss the advantages and disadvantages of each cycle configuration, and present preliminary performance and size results. This paper is intended to lay the groundwork for future efforts in the development of a practical NTP system or a combined NTP/NEP hybrid system.

Improved vortex reactor system

DOEpatents

Diebold, J.P.; Scahill, J.W.

1995-05-09

An improved vortex reactor system is described for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor. 12 figs.

Westinghouse Small Modular Reactor nuclear steam supply system design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Memmott, M. J.; Harkness, A. W.; Van Wyk, J.

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 andmore » 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

Self-actuating reactor shutdown system

DOEpatents

Barrus, Donald M.; Brummond, Willian A; Peterson, Leslie F.

1988-01-01

A control system for the automatic or self-actuated shutdown or "scram" of a nuclear reactor. The system is capable of initiating scram insertion by a signal from the plant protection system or by independent action directly sensing reactor conditions of low-flow or over-power. Self-actuation due to a loss of reactor coolant flow results from a decrease of pressure differential between the upper and lower ends of an absorber element. When the force due to this differential falls below the weight of the element, the element will fall by gravitational force to scram the reactor. Self-actuation due to high neutron flux is accomplished via a valve controlled by an electromagnet and a thermionic diode. In a reactor over-power, the diode will be heated to a change of state causing the electromagnet to be shorted thereby actuating the valve which provides the changed flow and pressure conditions required for scramming the absorber element.

Impact of thermal spectrum small modular reactors on performance of once-through nuclear fuel cycles with low-enriched uranium

DOE PAGES

Brown, Nicholas R.; Worrall, Andrew; Todosow, Michael

2016-11-18

Small modular reactors (SMRs) offer potential benefits, such as enhanced operational flexibility. However, it is vital to understand the holistic impact of SMRs on nuclear fuel cycle performance. The focus of this paper is the fuel cycle impacts of light water SMRs in a once-through fuel cycle with low-enriched uranium fuel. A key objective of this paper is to describe preliminary example reactor core physics and fuel cycle analyses conducted in support of the U.S. Department of Energy, Office of Nuclear Energy, Fuel Cycle Options Campaign. The hypothetical light water SMR example case considered in these preliminary scoping studies ismore » a cartridge type one-batch core with slightly less than 5.0% enrichment. Challenges associated with SMRs include increased neutron leakage, fewer assemblies in the core (and therefore fewer degrees of freedom in the core design), complex enrichment and burnable absorber loadings, full power operation with inserted control rods, the potential for frequent load-following operation, and shortened core height. Each of these will impact the achievable discharge burnup in the reactor and the fuel cycle performance. This paper summarizes a list of the factors relevant to SMR fuel, core, and operation that will impact fuel cycle performance. The high-level issues identified and preliminary scoping calculations in this paper are intended to inform on potential fuel cycle impacts of one-batch thermal spectrum SMRs. In particular, this paper highlights the impact of increased neutron leakage and reduced number of batches on the achievable burnup of the reactor. Fuel cycle performance metrics for a hypothetical example SMR are compared with those for a conventional three-batch light water reactor in the following areas: nuclear waste management, environmental impact, and resource utilization. The metrics performance for such an SMR is degraded for the mass of spent nuclear fuel and high-level waste disposed of, mass of depleted uranium

Innovative open air brayton combined cycle systems for the next generation nuclear power plants

NASA Astrophysics Data System (ADS)

Zohuri, Bahman

The purpose of this research was to model and analyze a nuclear heated multi-turbine power conversion system operating with atmospheric air as the working fluid. The air is heated by a molten salt, or liquid metal, to gas heat exchanger reaching a peak temperature of 660 0C. The effects of adding a recuperator or a bottoming steam cycle have been addressed. The calculated results are intended to identify paths for future work on the next generation nuclear power plant (GEN-IV). This document describes the proposed system in sufficient detail to communicate a good understanding of the overall system, its components, and intended uses. The architecture is described at the conceptual level, and does not replace a detailed design document. The main part of the study focused on a Brayton --- Rankine Combined Cycle system and a Recuperated Brayton Cycle since they offer the highest overall efficiencies. Open Air Brayton power cycles also require low cooling water flows relative to other power cycles. Although the Recuperated Brayton Cycle achieves an overall efficiency slightly less that the Brayton --- Rankine Combined Cycle, it is completely free of a circulating water system and can be used in a desert climate. Detailed results of modeling a combined cycle Brayton-Rankine power conversion system are presented. The Rankine bottoming cycle appears to offer a slight efficiency advantage over the recuperated Brayton cycle. Both offer very significant advantages over current generation Light Water Reactor steam cycles. The combined cycle was optimized as a unit and lower pressure Rankine systems seem to be more efficient. The combined cycle requires a lot less circulating water than current power plants. The open-air Brayton systems appear to be worth investigating, if the higher temperatures predicted for the Next Generation Nuclear Plant do materialize.

Shutdown system for a nuclear reactor

DOEpatents

Groh, E.F.; Olson, A.P.; Wade, D.C.; Robinson, B.W.

1984-06-05

An ultimate shutdown system is provided for termination of neutronic activity in a nuclear reactor. The shutdown system includes bead chains comprising spherical containers suspended on a flexible cable. The containers are comprised of mating hemispherical shells which provide a ruggedized enclosure for reactor poison material. The bead chains, normally suspended above the reactor core on storage spools, are released for downward travel upon command from an external reactor monitor. The chains are capable of horizontal movement, so as to flow around obstructions in the reactor during their downward motion. 8 figs.

Shutdown system for a nuclear reactor

DOEpatents

Groh, Edward F.; Olson, Arne P.; Wade, David C.; Robinson, Bryan W.

1984-01-01

An ultimate shutdown system is provided for termination of neutronic activity in a nuclear reactor. The shutdown system includes bead chains comprising spherical containers suspended on a flexible cable. The containers are comprised of mating hemispherical shells which provide a ruggedized enclosure for reactor poison material. The bead chains, normally suspended above the reactor core on storage spools, are released for downward travel upon command from an external reactor monitor. The chains are capable of horizontal movement, so as to flow around obstructions in the reactor during their downward motion.

Small space reactor power systems for unmanned solar system exploration missions

NASA Technical Reports Server (NTRS)

Bloomfield, Harvey S.

1987-01-01

A preliminary feasibility study of the application of small nuclear reactor space power systems to the Mariner Mark II Cassini spacecraft/mission was conducted. The purpose of the study was to identify and assess the technology and performance issues associated with the reactor power system/spacecraft/mission integration. The Cassini mission was selected because study of the Saturn system was identified as a high priority outer planet exploration objective. Reactor power systems applied to this mission were evaluated for two different uses. First, a very small 1 kWe reactor power system was used as an RTG replacement for the nominal spacecraft mission science payload power requirements while still retaining the spacecraft's usual bipropellant chemical propulsion system. The second use of reactor power involved the additional replacement of the chemical propulsion system with a small reactor power system and an electric propulsion system. The study also provides an examination of potential applications for the additional power available for scientific data collection. The reactor power system characteristics utilized in the study were based on a parametric mass model that was developed specifically for these low power applications. The model was generated following a neutronic safety and operational feasibility assessment of six small reactor concepts solicited from U.S. industry. This assessment provided the validation of reactor safety for all mission phases and generatad the reactor mass and dimensional data needed for the system mass model.

Variants of Regenerated Fissile Materials Usage in Thermal Reactors as the First Stage of Fuel Cycle Closing

NASA Astrophysics Data System (ADS)

Andrianova, E. A.; Tsibul'skiy, V. F.

2017-12-01

At present, 240 000 t of spent nuclear fuel (SF) has been accumulated in the world. Its long-term storage should meet safety conditions and requires noticeable finances, which grow every year. Obviously, this situation cannot exist for a long time; in the end, it will require a final decision. At present, several variants of solution of the problem of SF management are considered. Since most of the operating reactors and those under construction are thermal reactors, it is reasonable to assume that the structure of the nuclear power industry in the near and medium-term future will be unchanged, and it will be necessary to utilize plutonium in thermal reactors. In this study, different strategies of SF management are compared: open fuel cycle with long-term SF storage, closed fuel cycle with MOX fuel usage in thermal reactors and subsequent long-term storage of SF from MOX fuel, and closed fuel cycle in thermal reactors with heterogeneous fuel arrangement. The concept of heterogeneous fuel arrangement is considered in detail. While in the case of traditional fuel it is necessary to reprocess the whole amount of spent fuel, in the case of heterogeneous arrangement, it is possible to separate plutonium and 238U in different fuel rods. In this case, it is possible to achieve nearly complete burning of fissile isotopes of plutonium in fuel rods loaded with plutonium. These fuel rods with burned plutonium can be buried after cooling without reprocessing. They would contain just several percent of initially loaded plutonium, mainly even isotopes. Fuel rods with 238U alone should be reprocessed in the usual way.

In Situ and ex Situ Catalytic Pyrolysis of Pine in a Bench-Scale Fluidized Bed Reactor System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Iisa, Kristiina; French, Richard J.; Orton, Kellene A.

In situ and ex situ catalytic pyrolysis were compared in a system with two 2-in. bubbling fluidized bed reactors. Pine was pyrolyzed in the system with a catalyst, HZSM-5 with a silica-to-alumina ratio of 30, placed either in the first (pyrolysis) reactor or the second (upgrading) reactor. Both the pyrolysis and upgrading temperatures were 500 degrees C, and the weight hourly space velocity was 1.1 h -1. Five catalytic cycles were completed in each experiment. The catalytic cycles were continued until oxygenates in the vapors became dominant. The catalyst was then oxidized, after which a new catalytic cycle was begun.more » The in situ configuration gave slightly higher oil yield but also higher oxygen content than the ex situ configuration, which indicates that the catalyst deactivated faster in the in situ configuration than the ex situ configuration. Analysis of the spent catalysts confirmed higher accumulation of metals in the in situ experiment. In all experiments, the organic oil mass yields varied between 14 and 17% and the carbon efficiencies between 20 and 25%. The organic oxygen concentrations in the oils were 16-18%, which represented a 45% reduction compared to corresponding noncatalytic pyrolysis oils prepared in the same fluidized bed reactor system. GC/MS analysis showed the oils to contain one- to four-ring aromatic hydrocarbons and a variety of oxygenates (phenols, furans, benzofurans, methoxyphenols, naphthalenols, indenols). Lastly, high fractions of oxygen were rejected as water, CO, and CO 2, which indicates the importance of dehydration, decarbonylation, and decarboxylation reactions. Light gases were the major sources of carbon losses, followed by char and coke.« less

In Situ and ex Situ Catalytic Pyrolysis of Pine in a Bench-Scale Fluidized Bed Reactor System

DOE PAGES

Iisa, Kristiina; French, Richard J.; Orton, Kellene A.; ...

2016-02-03

In situ and ex situ catalytic pyrolysis were compared in a system with two 2-in. bubbling fluidized bed reactors. Pine was pyrolyzed in the system with a catalyst, HZSM-5 with a silica-to-alumina ratio of 30, placed either in the first (pyrolysis) reactor or the second (upgrading) reactor. Both the pyrolysis and upgrading temperatures were 500 degrees C, and the weight hourly space velocity was 1.1 h -1. Five catalytic cycles were completed in each experiment. The catalytic cycles were continued until oxygenates in the vapors became dominant. The catalyst was then oxidized, after which a new catalytic cycle was begun.more » The in situ configuration gave slightly higher oil yield but also higher oxygen content than the ex situ configuration, which indicates that the catalyst deactivated faster in the in situ configuration than the ex situ configuration. Analysis of the spent catalysts confirmed higher accumulation of metals in the in situ experiment. In all experiments, the organic oil mass yields varied between 14 and 17% and the carbon efficiencies between 20 and 25%. The organic oxygen concentrations in the oils were 16-18%, which represented a 45% reduction compared to corresponding noncatalytic pyrolysis oils prepared in the same fluidized bed reactor system. GC/MS analysis showed the oils to contain one- to four-ring aromatic hydrocarbons and a variety of oxygenates (phenols, furans, benzofurans, methoxyphenols, naphthalenols, indenols). Lastly, high fractions of oxygen were rejected as water, CO, and CO 2, which indicates the importance of dehydration, decarbonylation, and decarboxylation reactions. Light gases were the major sources of carbon losses, followed by char and coke.« less

Solvent refined coal reactor quench system

DOEpatents

Thorogood, Robert M.

1983-01-01

There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream.

Solvent refined coal reactor quench system

DOEpatents

Thorogood, R.M.

1983-11-08

There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream. 1 fig.

Thermionic switched self-actuating reactor shutdown system

DOEpatents

Barrus, Donald M.; Shires, Charles D.; Brummond, William A.

1989-01-01

A self-actuating reactor shutdown system incorporating a thermionic switched electromagnetic latch arrangement which is responsive to reactor neutron flux changes and to reactor coolant temperature changes. The system is self-actuating in that the sensing thermionic device acts directly to release (scram) the control rod (absorber) without reference or signal from the main reactor plant protective and control systems. To be responsive to both temperature and neutron flux effects, two detectors are used, one responsive to reactor coolant temperatures, and the other responsive to reactor neutron flux increase. The detectors are incorporated into a thermionic diode connected electrically with an electromagnetic mechanism which under normal reactor operating conditions holds the the control rod in its ready position (exterior of the reactor core). Upon reaching either a specified temperature or neutron flux, the thermionic diode functions to short-circuit the electromagnetic mechanism causing same to lose its holding power and release the control rod, which drops into the reactor core region under gravitational force.

Reactor vessel support system. [LMFBR

DOEpatents

Golden, M.P.; Holley, J.C.

1980-05-09

A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

Assessment of the Neutronic and Fuel Cycle Performance of the Transatomic Power Molten Salt Reactor Design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Robertson, Sean; Dewan, Leslie; Massie, Mark

This report presents results from a collaboration between Transatomic Power Corporation (TAP) and Oak Ridge National Laboratory (ORNL) to provide neutronic and fuel cycle analysis of the TAP core design through the Department of Energy Gateway for Accelerated Innovation in Nuclear (GAIN) Nuclear Energy Voucher program. The TAP concept is a molten salt reactor using configurable zirconium hydride moderator rod assemblies to shift the neutron spectrum in the core from mostly epithermal at beginning of life to thermal at end of life. Additional developments in the ChemTriton modeling and simulation tool provide the critical moderator-to-fuel ratio searches and time-dependent parametersmore » necessary to simulate the continuously changing physics in this complex system. The implementation of continuous-energy Monte Carlo transport and depletion tools in ChemTriton provide for full-core three-dimensional modeling and simulation. Results from simulations with these tools show agreement with TAP-calculated performance metrics for core lifetime, discharge burnup, and salt volume fraction, verifying the viability of reducing actinide waste production with this concept. Additional analyses of mass feed rates and enrichments, isotopic removals, tritium generation, core power distribution, core vessel helium generation, moderator rod heat deposition, and reactivity coeffcients provide additional information to make informed design decisions. This work demonstrates capabilities of ORNL modeling and simulation tools for neutronic and fuel cycle analysis of molten salt reactor concepts.« less

Advanced Fuel Cycles for Fusion Reactors: Passive Safety and Zero-Waste Options

NASA Astrophysics Data System (ADS)

Zucchetti, Massimo; Sugiyama, Linda E.

2006-05-01

Nuclear fusion is seen as a much ''cleaner'' energy source than fission. Most of the studies and experiments on nuclear fusion are currently devoted to the Deuterium-Tritium (DT) fuel cycle, since it is the easiest way to reach ignition. The recent stress on safety by the world's community has stimulated the research on other fuel cycles than the DT one, based on 'advanced' reactions, such as the Deuterium-Helium-3 (DHe) one. These reactions pose problems, such as the availability of 3He and the attainment of the higher plasma parameters that are required for burning. However, they have many advantages, like for instance the very low neutron activation, while it is unnecessary to breed and fuel tritium. The extrapolation of Ignitor technologies towards a larger and more powerful experiment using advanced fuel cycles (Candor) has been studied. Results show that Candor does reach the passive safety and zero-waste option. A fusion power reactor based on the DHe cycle could be the ultimate response to the environmental requirements for future nuclear power plants.

Mass tracking and material accounting in the integral fast reactor (IFR)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Orechwa, Y.; Adams, C.H.; White, A.M.

1991-01-01

This paper reports on the Integral Fast Reactor (IFR) which is a generic advanced liquid metal cooled reactor concept being developed at Argonne National Laboratory. There are a number of technical features of the IFR which contribute to its potential as a next-generation reactor. These are associated with large safety margins with regard to off-normal events involving the heat transport system, and the use of metallic fuel which makes possible the utilization of innovative fuel cycle processes. The latter feature permits fuel cycle closure with compact, low-cost reprocessing facilities, collocated with the reactor plant. These primary features are being demonstratedmore » in the facilities at ANL-West, utilizing Experimental Breeder Reactor II and the associated Fuel Cycle Facility (FCF) as an IFR prototype. The demonstration of this IFR prototype includes the design and implementation of the Mass-tracking System (MTG). In this system, data from the operations of the FCF, including weights and batch-process parameters, are collected and maintained by the MTG running on distributed workstations.« less

10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Acceptance criteria for reactor coolant system venting... criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high point vents for the reactor coolant system, for the reactor vessel head, and for other systems required...

10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 1 2011-01-01 2011-01-01 false Acceptance criteria for reactor coolant system venting... criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high point vents for the reactor coolant system, for the reactor vessel head, and for other systems required...

The feasibility study of small long-life gas cooled fast reactor with mixed natural Uranium/Thorium as fuel cycle input

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ariani, Menik; Su'ud, Zaki; Waris, Abdul

2012-06-06

A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE burn-up scheme has been performed. In this study, design GCFR with Helium coolant which can be continuously operated by supplying mixed Natural Uranium/Thorium without fuel enrichment plant or fuel reprocessing plant. The active reactor cores are divided into two region, Thorium fuel region and Uranium fuel region. Each fuel core regions are subdivided into ten parts (region-1 until region-10) with the same volume in the axial direction. The fresh Natural Uranium and Thorium is initially put in region-1, after one cycle of 10 years of burn-up it ismore » shifted to region-2 and the each region-1 is filled by fresh natural Uranium/Thorium fuel. This concept is basically applied to all regions in both cores area, i.e. shifted the core of i{sup th} region into i+1 region after the end of 10 years burn-up cycle. For the next cycles, we will add only Natural Uranium and Thorium on each region-1. The calculation results show the reactivity reached by mixed Natural Uranium/Thorium with volume ratio is 4.7:1. This reactor can results power thermal 550 MWth. After reactor start-up the operation, furthermore reactor only needs Natural Uranium/Thorium supply for continue operation along 100 years.« less

Modeling and Simulations for the High Flux Isotope Reactor Cycle 400

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ilas, Germina; Chandler, David; Ade, Brian J

2015-03-01

A concerted effort over the past few years has been focused on enhancing the core model for the High Flux Isotope Reactor (HFIR), as part of a comprehensive study for HFIR conversion from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel. At this time, the core model used to perform analyses in support of HFIR operation is an MCNP model for the beginning of Cycle 400, which was documented in detail in a 2005 technical report. A HFIR core depletion model that is based on current state-of-the-art methods and nuclear data was needed to serve as reference for the designmore » of an LEU fuel for HFIR. The recent enhancements in modeling and simulations for HFIR that are discussed in the present report include: (1) revision of the 2005 MCNP model for the beginning of Cycle 400 to improve the modeling data and assumptions as necessary based on appropriate primary reference sources HFIR drawings and reports; (2) improvement of the fuel region model, including an explicit representation for the involute fuel plate geometry that is characteristic to HFIR fuel; and (3) revision of the Monte Carlo-based depletion model for HFIR in use since 2009 but never documented in detail, with the development of a new depletion model for the HFIR explicit fuel plate representation. The new HFIR models for Cycle 400 are used to determine various metrics of relevance to reactor performance and safety assessments. The calculated metrics are compared, where possible, with measurement data from preconstruction critical experiments at HFIR, data included in the current HFIR safety analysis report, and/or data from previous calculations performed with different methods or codes. The results of the analyses show that the models presented in this report provide a robust and reliable basis for HFIR analyses.« less

TREAT Reactor Control and Protection System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lipinski, W.C.; Brookshier, W.K.; Burrows, D.R.

1985-01-01

The main control algorithm of the Transient Reactor Test Facility (TREAT) Automatic Reactor Control System (ARCS) resides in Read Only Memory (ROM) and only experiment specific parameters are input via keyboard entry. Prior to executing an experiment, the software and hardware of the control computer is tested by a closed loop real-time simulation. Two computers with parallel processing are used for the reactor simulation and another computer is used for simulation of the control rod system. A monitor computer, used as a redundant diverse reactor protection channel, uses more conservative setpoints and reduces challenges to the Reactor Trip System (RTS).more » The RTS consists of triplicated hardwired channels with one out of three logic. The RTS is automatically tested by a digital Dedicated Microprocessor Tester (DMT) prior to the execution of an experiment. 6 refs., 5 figs., 1 tab.« less

Reactor/Brayton power systems for nuclear electric spacecraft

NASA Technical Reports Server (NTRS)

Layton, J. P.

1980-01-01

Studies are currently underway to assess the technological feasibility of a nuclear-reactor-powered spacecraft propelled by electric thrusters. This vehicle would be capable of performing detailed exploration of the outer planets of the solar system during the remainder of this century. The purpose of this study was to provide comparative information on a closed cycle gas turbine power conversion system. The results have shown that the performance is very competitive and that a 400 kWe space power system is dimensionally compatible with a single Space Shuttle launch. Performance parameters of system mass and radiator area were determined for systems from 100 to 1000 kWe. A 400 kWe reference system received primary attention. The components of this system were defined and a conceptual layout was developed with encouraging results. The preliminary mass determination for the complete power system was very close to the desired goal of 20 kg/kWe. Use of more advanced technology (higher turbine inlet temperature) will substantially improve system performance characteristics.

Plasma reactor waste management systems

NASA Technical Reports Server (NTRS)

Ness, Robert O., Jr.; Rindt, John R.; Ness, Sumitra R.

1992-01-01

The University of North Dakota is developing a plasma reactor system for use in closed-loop processing that includes biological, materials, manufacturing, and waste processing. Direct-current, high-frequency, or microwave discharges will be used to produce plasmas for the treatment of materials. The plasma reactors offer several advantages over other systems, including low operating temperatures, low operating pressures, mechanical simplicity, and relatively safe operation. Human fecal material, sunflowers, oats, soybeans, and plastic were oxidized in a batch plasma reactor. Over 98 percent of the organic material was converted to gaseous products. The solids were then analyzed and a large amount of water and acid-soluble materials were detected. These materials could possibly be used as nutrients for biological systems.

Cooling system for a nuclear reactor

DOEpatents

Amtmann, Hans H.

1982-01-01

A cooling system for a gas-cooled nuclear reactor is disclosed which includes at least one primary cooling loop adapted to pass coolant gas from the reactor core and an associated steam generator through a duct system having a main circulator therein, and at least one auxiliary cooling loop having communication with the reactor core and adapted to selectively pass coolant gas through an auxiliary heat exchanger and circulator. The main and auxiliary circulators are installed in a common vertical cavity in the reactor vessel, and a common return duct communicates with the reactor core and intersects the common cavity at a junction at which is located a flow diverter valve operative to effect coolant flow through either the primary or auxiliary cooling loops.

Improved Nuclear Reactor and Shield Mass Model for Space Applications

NASA Technical Reports Server (NTRS)

Robb, Kevin

2004-01-01

New technologies are being developed to explore the distant reaches of the solar system. Beyond Mars, solar energy is inadequate to power advanced scientific instruments. One technology that can meet the energy requirements is the space nuclear reactor. The nuclear reactor is used as a heat source for which a heat-to-electricity conversion system is needed. Examples of such conversion systems are the Brayton, Rankine, and Stirling cycles. Since launch cost is proportional to the amount of mass to lift, mass is always a concern in designing spacecraft. Estimations of system masses are an important part in determining the feasibility of a design. I worked under Michael Barrett in the Thermal Energy Conversion Branch of the Power & Electric Propulsion Division. An in-house Closed Cycle Engine Program (CCEP) is used for the design and performance analysis of closed-Brayton-cycle energy conversion systems for space applications. This program also calculates the system mass including the heat source. CCEP uses the subroutine RSMASS, which has been updated to RSMASS-D, to estimate the mass of the reactor. RSMASS was developed in 1986 at Sandia National Laboratories to quickly estimate the mass of multi-megawatt nuclear reactors for space applications. In response to an emphasis for lower power reactors, RSMASS-D was developed in 1997 and is based off of the SP-100 liquid metal cooled reactor. The subroutine calculates the mass of reactor components such as the safety systems, instrumentation and control, radiation shield, structure, reflector, and core. The major improvements in RSMASS-D are that it uses higher fidelity calculations, is easier to use, and automatically optimizes the systems mass. RSMASS-D is accurate within 15% of actual data while RSMASS is only accurate within 50%. My goal this summer was to learn FORTRAN 77 programming language and update the CCEP program with the RSMASS-D model.

Thermal tests of a multi-tubular reactor for hydrogen production by using mixed ferrites thermochemical cycle

NASA Astrophysics Data System (ADS)

Gonzalez-Pardo, Aurelio; Denk, Thorsten; Vidal, Alfonso

2017-06-01

The SolH2 project is an INNPACTO initiative of the Spanish Ministry of Economy and Competitiveness, with the main goal to demonstrate the technological feasibility of solar thermochemical water splitting cycles as one of the most promising options to produce H2 from renewable sources in an emission-free way. A multi-tubular solar reactor was designed and build to evaluate a ferrite thermochemical cycle. At the end of this project, the ownership of this plant was transferred to CIEMAT. This paper reviews some additional tests with this pilot plant performed in the Plataforma Solar de Almería with the main goal to assess the thermal behavior of the reactor, evaluating the evolution of the temperatures inside the cavity and the relation between supplied power and reached temperatures. Previous experience with alumina tubes showed that they are very sensitive to temperature and flux gradients, what leads to elaborate an aiming strategy for the heliostat field to achieve a uniform distribution of the radiation inside the cavity. Additionally, the passing of clouds is a phenomenon that importantly affects all the CSP facilities by reducing their efficiency. The behavior of the reactor under these conditions has been studied.

Investigation of alternative layouts for the supercritical carbon dioxide Brayton cycle for a sodium-cooled fast reactor.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Moisseytsev, A.; Sienicki, J. J.

2009-07-01

Analyses of supercritical carbon dioxide (S-CO{sub 2}) Brayton cycle performance have largely settled on the recompression supercritical cycle (or Feher cycle) incorporating a flow split between the main compressor downstream of heat rejection, a recompressing compressor providing direct compression without heat rejection, and high and low temperature recuperators to raise the effectiveness of recuperation and the cycle efficiency. Alternative cycle layouts have been previously examined by Angelino (Politecnico, Milan), by MIT (Dostal, Hejzlar, and Driscoll), and possibly others but not for sodium-cooled fast reactors (SFRs) operating at relatively low core outlet temperature. Thus, the present authors could not be suremore » that the recompression cycle is an optimal arrangement for application to the SFR. To ensure that an advantageous alternative layout has not been overlooked, several alternative cycle layouts have been investigated for a S-CO{sub 2} Brayton cycle coupled to the Advanced Burner Test Reactor (ABTR) SFR preconceptual design having a 510 C core outlet temperature and a 470 C turbine inlet temperature to determine if they provide any benefit in cycle performance (e.g., enhanced cycle efficiency). No such benefits were identified, consistent with the previous examinations, such that attention was devoted to optimizing the recompression supercritical cycle. The effects of optimizing the cycle minimum temperature and pressure are investigated including minimum temperatures and/or pressures below the critical values. It is found that improvements in the cycle efficiency of 1% or greater relative to previous analyses which arbitrarily fixed the minimum temperature and pressure can be realized through an optimal choice of the combination of the minimum cycle temperature and pressure (e.g., for a fixed minimum temperature there is an optimal minimum pressure). However, this leads to a requirement for a larger cooler for heat rejection which may impact

Liquid metal cooled nuclear reactor plant system

DOEpatents

Hunsbedt, Anstein; Boardman, Charles E.

1993-01-01

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting for fuel decay during reactor shutdown, or heat produced during a mishap. The reactor system is enhanced with sealing means for excluding external air from contact with the liquid metal coolant leaking from the reactor vessel during an accident. The invention also includes a silo structure which resists attack by leaking liquid metal coolant, and an added unique cooling means.

Comparison of the performance of MBBR and SBR systems for the treatment of anaerobic reactor biowaste effluent.

PubMed

Comett-Ambriz, I; Gonzalez-Martinez, S; Wilderer, P

2003-01-01

Anaerobic reactor biowaste effluent was treated with biofilm and activated sludge sequencing batch reactors to compare the performance of both systems. The treatment targets were organic carbon removal and nitrification. The pilot plant was operated in two phases. During the first phase, it was operated like a Moving Bed Biofilm Reactor (MBBR) with the Natrix media, with a specific surface area of 210 m2/m3. The MBBR was operated under Sequencing Batch Reactor (SBR) modality with three 8-hour cycles per day over 70 days. During the second phase of the experiment, the pilot plant was operated over 79 days as a SBR. In both phases the influent was fed to the reactor at a flow rate corresponding to a Hydraulic Retention Time (HRT) of 4 days. Both systems presented a good carbon removal for this specific wastewater. The Chemical Oxygen Demand (COD) total removal was 53% for MBBR and 55% for SBR. MBBR offered a higher dissolved COD removal (40%) than SBR (30%). The limited COD removal achieved is in agreement with the high COD to BOD5 ratio (1/3) of the influent wastewater. In both systems a complete nitrification was obtained. The different efficiencies in both systems are related to the different biomass concentrations.

75 FR 35001 - Blue Ribbon Commission on America's Nuclear Future, Reactor and Fuel Cycle Technologies Subcommittee

Federal Register 2010, 2011, 2012, 2013, 2014

2010-06-21

... DEPARTMENT OF ENERGY Blue Ribbon Commission on America's Nuclear Future, Reactor and Fuel Cycle Technologies Subcommittee AGENCY: Department of Energy, Office of Nuclear Energy. ACTION: Notice of Open... facsimile (202) 586- 0544; e-mail [email protected]nuclear.energy.gov . Additional information may also be...

75 FR 61139 - Blue Ribbon Commission on America's Nuclear Future, Reactor and Fuel Cycle Technology Subcommittee

Federal Register 2010, 2011, 2012, 2013, 2014

2010-10-04

... DEPARTMENT OF ENERGY Blue Ribbon Commission on America's Nuclear Future, Reactor and Fuel Cycle Technology Subcommittee AGENCY: Department of Energy, Office of Nuclear Energy. ACTION: Notice of Open...) 586- 0544; e-mail [email protected]nuclear.energy.gov . Additional information will be available at http...

Advantages of Production of New Fissionable Nuclides for the Nuclear Power Industry in Hybrid Fusion-Fission Reactors

NASA Astrophysics Data System (ADS)

Tsibulskiy, V. F.; Andrianova, E. A.; Davidenko, V. D.; Rodionova, E. V.; Tsibulskiy, S. V.

2017-12-01

A concept of a large-scale nuclear power engineering system equipped with fusion and fission reactors is presented. The reactors have a joint fuel cycle, which imposes the lowest risk of the radiation impact on the environment. The formation of such a system is considered within the framework of the evolution of the current nuclear power industry with the dominance of thermal reactors, gradual transition to the thorium fuel cycle, and integration into the system of the hybrid fusion-fission reactors for breeding nuclear fuel for fission reactors. Such evolution of the nuclear power engineering system will allow preservation of the existing structure with the dominance of thermal reactors, enable the reprocessing of the spent nuclear fuel (SNF) with low burnup, and prevent the dangerous accumulation of minor actinides. The proposed structure of the nuclear power engineering system minimizes the risk of radioactive contamination of the environment and the SNF reprocessing facilities, decreasing it by more than one order of magnitude in comparison with the proposed scheme of closing the uranium-plutonium fuel cycle based on the reprocessing of SNF with high burnup from fast reactors.

High efficiency Dual-Cycle Conversion System using Kr-85.

PubMed

Prelas, Mark A; Tchouaso, Modeste Tchakoua

2018-04-26

This paper discusses the use of one of the safest isotopes known isotopes, Kr-85, as a candidate fuel source for deep space missions. This isotope comes from 0.286% of fission events. There is a vast quantity of Kr-85 stored in spent fuel and it is continually being produced by nuclear reactors. In using Kr-85 with a novel Dual Cycle Conversion System (DCCS) it is feasible to boost the system efficiency from 26% to 45% over a single cycle device while only increasing the system mass by less than 1%. The Kr-85 isotope is the ideal fuel for a Photon Intermediate Direct Energy Conversion (PIDEC) system. PIDEC is an excellent choice for the top cycle in a DCCS. In the top cycle, ionization and excitation of the Kr-85:Cl gas mixture (99% Kr and 1% Cl) from beta particles creates KrCl* excimer photons which are efficiently absorbed by diamond photovoltaic cells on the walls of the pressure vessels. The benefit of using the DCCS is that Kr-85 is capable of operating at high temperatures in the primary cycle and the residual heat can then be converted into electrical power in the bottom cycle which uses a Stirling Engine. The design of the DCCS begins with a spherical pressure vessel of radius 13.7 cm with 3.7 cm thick walls and is filled with a Kr-85:Cl gas mixture. The inner wall has diamond photovoltaic cells attached to it and there is a sapphire window between the diamond photovoltaic cells and the Kr-85:Cl gas mixture which shields the photovoltaic cells from beta particles. The DCCS without a gamma ray shield has specific power of 6.49 W/kg. A removable 6 cm thick tungsten shield is used to safely limit the radiation exposure levels of personnel. A shadow shield remains in the payload to protect the radiation sensitive components in the flight package. The estimated specific power of the unoptimized system design in this paper is about 2.33 W/kg. The specific power of an optimized system should be higher. The Kr-85 isotope is relatively safe because it

Two-Dimensional Neutronic and Fuel Cycle Analysis of the Transatomic Power Molten Salt Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Betzler, Benjamin R.; Powers, Jeffrey J.; Worrall, Andrew

2017-01-15

This status report presents the results from the first phase of the collaboration between Transatomic Power Corporation (TAP) and Oak Ridge National Laboratory (ORNL) to provide neutronic and fuel cycle analysis of the TAP core design through the Department of Energy Gateway for Accelerated Innovation in Nuclear, Nuclear Energy Voucher program. The TAP design is a molten salt reactor using movable moderator rods to shift the neutron spectrum in the core from mostly epithermal at beginning of life to thermal at end of life. Additional developments in the ChemTriton modeling and simulation tool provide the critical moderator-to-fuel ratio searches andmore » time-dependent parameters necessary to simulate the continuously changing physics in this complex system. Results from simulations with these tools show agreement with TAP-calculated performance metrics for core lifetime, discharge burnup, and salt volume fraction, verifying the viability of reducing actinide waste production with this design. Additional analyses of time step sizes, mass feed rates and enrichments, and isotopic removals provide additional information to make informed design decisions. This work further demonstrates capabilities of ORNL modeling and simulation tools for analysis of molten salt reactor designs and strongly positions this effort for the upcoming three-dimensional core analysis.« less

Rapid starting methanol reactor system

DOEpatents

Chludzinski, Paul J.; Dantowitz, Philip; McElroy, James F.

1984-01-01

The invention relates to a methanol-to-hydrogen cracking reactor for use with a fuel cell vehicular power plant. The system is particularly designed for rapid start-up of the catalytic methanol cracking reactor after an extended shut-down period, i.e., after the vehicular fuel cell power plant has been inoperative overnight. Rapid system start-up is accomplished by a combination of direct and indirect heating of the cracking catalyst. Initially, liquid methanol is burned with a stoichiometric or slightly lean air mixture in the combustion chamber of the reactor assembly. The hot combustion gas travels down a flue gas chamber in heat exchange relationship with the catalytic cracking chamber transferring heat across the catalyst chamber wall to heat the catalyst indirectly. The combustion gas is then diverted back through the catalyst bed to heat the catalyst pellets directly. When the cracking reactor temperature reaches operating temperature, methanol combustion is stopped and a hot gas valve is switched to route the flue gas overboard, with methanol being fed directly to the catalytic cracking reactor. Thereafter, the burner operates on excess hydrogen from the fuel cells.

Comparing the new generation accelerator driven subcritical reactor system (ADS) to traditional critical reactors

NASA Astrophysics Data System (ADS)

Kemah, Elif; Akkaya, Recep; Tokgöz, Seyit Rıza

2017-02-01

In recent years, the accelerator driven subcritical reactors have taken great interest worldwide. The Accelerator Driven System (ADS) has been used to produce neutron in subcritical state by the external proton beam source. These reactors, which are hybrid systems, are important in production of clean and safe energy and conversion of radioactive waste. The ADS with the selection of reliability and robust target materials have been the new generation of fission reactors. In addition, in the ADS Reactors the problems of long-lived radioactive fission products and waste actinides encountered in the fission process of the reactor during incineration can be solved, and ADS has come to the forefront of thorium as fuel for the reactors.

Integral reactor system and method for fuel cells

DOEpatents

Fernandes, Neil Edward; Brown, Michael S.; Cheekatamaria, Praveen; Deng, Thomas; Dimitrakopoulos, James; Litka, Anthony F.

2017-03-07

A reactor system is integrated internally within an anode-side cavity of a fuel cell. The reactor system is configured to convert higher hydrocarbons to smaller species while mitigating the lower production of solid carbon. The reactor system may incorporate one or more of a pre-reforming section, an anode exhaust gas recirculation device, and a reforming section.

Integral reactor system and method for fuel cells

DOEpatents

Fernandes, Neil Edward; Brown, Michael S; Cheekatamarla, Praveen; Deng, Thomas; Dimitrakopoulos, James; Litka, Anthony F

2013-11-19

A reactor system is integrated internally within an anode-side cavity of a fuel cell. The reactor system is configured to convert hydrocarbons to smaller species while mitigating the lower production of solid carbon. The reactor system may incorporate one or more of a pre-reforming section, an anode exhaust gas recirculation device, and a reforming section.

Reactor power system deployment and startup

NASA Technical Reports Server (NTRS)

Wetch, J. R.; Nelin, C. J.; Britt, E. J.; Klein, G.

1985-01-01

This paper addresses issues that should receive further examination in the near-term as concept selection for development of a U.S. space reactor power system is approached. The issues include: the economics, practicality and system reliability associated with transfer of nuclear spacecraft from low earth shuttle orbits to operational orbits, via chemical propulsion versus nuclear electric propulsion; possible astronaut supervised reactor and nuclear electric propulsion startup in low altitude Shuttle orbit; potential deployment methods for nuclear powered spacecraft from Shuttle; the general public safety of low altitude startup and nuclear safe and disposal orbits; the question of preferred reactor power level; and the question of frozen versus molten alkali metal coolant during launch and deployment. These issues must be considered now because they impact the SP-100 concept selection, power level selection, weight and size limits, use of deployable radiators, reliability requirements, and economics, as well as the degree of need for and the urgency of developing space reactor power systems.

Optimization and Comparison of Direct and Indirect Supercritical Carbon Dioxide Power Plant Cycles for Nuclear Applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Edwin A. Harvego; Michael G. McKellar

2011-11-01

There have been a number of studies involving the use of gases operating in the supercritical mode for power production and process heat applications. Supercritical carbon dioxide (CO2) is particularly attractive because it is capable of achieving relatively high power conversion cycle efficiencies in the temperature range between 550 C and 750 C. Therefore, it has the potential for use with any type of high-temperature nuclear reactor concept, assuming reactor core outlet temperatures of at least 550 C. The particular power cycle investigated in this paper is a supercritical CO2 Recompression Brayton Cycle. The CO2 Recompression Brayton Cycle can bemore » used as either a direct or indirect power conversion cycle, depending on the reactor type and reactor outlet temperature. The advantage of this cycle when compared to the helium Brayton cycle is the lower required operating temperature; 550 C versus 850 C. However, the supercritical CO2 Recompression Brayton Cycle requires an operating pressure in the range of 20 MPa, which is considerably higher than the required helium Brayton cycle operating pressure of 8 MPa. This paper presents results of analyses performed using the UniSim process analyses software to evaluate the performance of both a direct and indirect supercritical CO2 Brayton Recompression cycle for different reactor outlet temperatures. The direct supercritical CO2 cycle transferred heat directly from a 600 MWt reactor to the supercritical CO2 working fluid supplied to the turbine generator at approximately 20 MPa. The indirect supercritical CO2 cycle assumed a helium-cooled Very High Temperature Reactor (VHTR), operating at a primary system pressure of approximately 7.0 MPa, delivered heat through an intermediate heat exchanger to the secondary indirect supercritical CO2 Brayton Recompression cycle, again operating at a pressure of about 20 MPa. For both the direct and indirect cycles, sensitivity calculations were performed for reactor outlet

Seismic attenuation system for a nuclear reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liszkai, Tamas; Cadell, Seth

A system for attenuating seismic forces includes a reactor pressure vessel containing nuclear fuel and a containment vessel that houses the reactor pressure vessel. Both the reactor pressure vessel and the containment vessel include a bottom head. Additionally, the system includes a base support to contact a support surface on which the containment vessel is positioned in a substantially vertical orientation. An attenuation device is located between the bottom head of the reactor pressure vessel and the bottom head of the containment vessel. Seismic forces that travel from the base support to the reactor pressure vessel via the containment vesselmore » are attenuated by the attenuation device in a direction that is substantially lateral to the vertical orientation of the containment vessel.« less

Mass tracking and material accounting in the Integral Fast Reactor (IFR)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Orechwa, Y.; Adams, C.H.; White, A.M.

1991-01-01

The Integral Fast Reactor (IFR) is a generic advanced liquid metal cooled reactor concept being developed at Argonne National Laboratory (ANL). There are a number of technical features of the IFR which contribute to its potential as a next-generation reactor. These are associated with large safety margins with regard to off-normal events involving the heat transport system, and the use of metallic fuel which makes possible the utilization of innovative fuel cycle processes. The latter feature permits fuel cycle closure the compact, low-cost reprocessing facilities, collocated with the reactor plant. These primary features are being demonstrated in the facilities atmore » ANL-West, utilizing Experimental Breeder Reactor 2 and the associated Fuel Cycle Facility (FCF) as an IFR prototype. The demonstration of this IFR prototype includes the design and implementation of the Mass-Tracking System (MTG). In this system, data from the operations of the FCF, including weights and batch-process parameters, are collected and maintained by the MTG running on distributed workstations. The components of the MTG System include: (1) an Oracle database manager with a Fortran interface, (2) a set of MTG Tasks'' which collect, manipulate and report data, (3) a set of MTG Terminal Sessions'' which provide some interactive control of the Tasks, and (4) a set of servers which manage the Tasks and which provide the communications link between the MTG System and Operator Control Stations, which control process equipment and monitoring devices within the FCF.« less

Trench fast reactor design using the microcomputer

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rohach, A.F.; Sankoorikal, J.T.; Schmidt, R.R.

1987-01-01

This project is a study of alternative liquid-metal-cooled fast power reactor system concepts. Specifically, an unconventional primary system is being conceptually designed and evaluated. The project design is based primarily on microcomputer analysis through the use of computational modules. The reactor system concept is a long, narrow pool with a long, narrow reactor called a trench-type pool reactor in it. The reactor consists of five core-blanket modules in a line. Specific power is to be modest, permitting long fuel residence time. Two fuel cycles are currently being considered. The reactor design philosophy is that of the inherently safe concept. Thismore » requires transient analysis dependent on reactivity coefficients: prompt fuel, including Doppler and expansion, fuel expansion, sodium temperature and void, and core expansion. Conceptual reactor design is done on a microcomputer. A part of the trench reactor project is to develop a microcomputer-based system that can be used by the user for scoping studies and design. Current development includes the neutronics and fuel management aspects of the design. Thermal-hydraulic analysis and economics are currently being incorporated into the microcomputer system. The system is menu-driven including preparation of program input data and of output data for displays in graphics form.« less

The Dynomak: An advanced spheromak reactor system with imposed-dynamo current drive and next-generation nuclear power technologies

NASA Astrophysics Data System (ADS)

Sutherland, D. A.; Jarboe, T. R.; Marklin, G.; Morgan, K. D.; Nelson, B. A.

2013-10-01

A high-beta spheromak reactor system has been designed with an overnight capital cost that is competitive with conventional power sources. This reactor system utilizes recently discovered imposed-dynamo current drive (IDCD) and a molten salt blanket system for first wall cooling, neutron moderation and tritium breeding. Currently available materials and ITER developed cryogenic pumping systems were implemented in this design on the basis of technological feasibility. A tritium breeding ratio of greater than 1.1 has been calculated using a Monte Carlo N-Particle (MCNP5) neutron transport simulation. High-temperature superconducting tapes (YBCO) were used for the equilibrium coil set, substantially reducing the recirculating power fraction when compared to previous spheromak reactor studies. Using zirconium hydride for neutron shielding, a limiting equilibrium coil lifetime of at least thirty full-power years has been achieved. The primary FLiBe loop was coupled to a supercritical carbon dioxide Brayton cycle due to attractive economics and high thermal efficiencies. With these advancements, an electrical output of 1000 MW from a thermal output of 2486 MW was achieved, yielding an overall plant efficiency of approximately 40%. A paper concerning the Dynomak reactor design is currently being reviewed for publication.

Process Model of A Fusion Fuel Recovery System for a Direct Drive IFE Power Reactor

NASA Astrophysics Data System (ADS)

Natta, Saswathi; Aristova, Maria; Gentile, Charles

2008-11-01

A task has been initiated to develop a detailed representative model for the fuel recovery system (FRS) in the prospective direct drive inertial fusion energy (IFE) reactor. As part of the conceptual design phase of the project, a chemical process model is developed in order to observe the interaction of system components. This process model is developed using FEMLAB Multiphysics software with the corresponding chemical engineering module (CEM). Initially, the reactants, system structure, and processes are defined using known chemical species of the target chamber exhaust. Each step within the Fuel recovery system is modeled compartmentally and then merged to form the closed loop fuel recovery system. The output, which includes physical properties and chemical content of the products, is analyzed after each step of the system to determine the most efficient and productive system parameters. This will serve to attenuate possible bottlenecks in the system. This modeling evaluation is instrumental in optimizing and closing the fusion fuel cycle in a direct drive IFE power reactor. The results of the modeling are presented in this paper.

Particle bed reactor modeling

NASA Technical Reports Server (NTRS)

Sapyta, Joe; Reid, Hank; Walton, Lew

1993-01-01

The topics are presented in viewgraph form and include the following: particle bed reactor (PBR) core cross section; PBR bleed cycle; fuel and moderator flow paths; PBR modeling requirements; characteristics of PBR and nuclear thermal propulsion (NTP) modeling; challenges for PBR and NTP modeling; thermal hydraulic computer codes; capabilities for PBR/reactor application; thermal/hydralic codes; limitations; physical correlations; comparison of predicted friction factor and experimental data; frit pressure drop testing; cold frit mask factor; decay heat flow rate; startup transient simulation; and philosophy of systems modeling.

The 5-kwe reactor thermoelectric system summary

NASA Technical Reports Server (NTRS)

Vanosdol, J. H. (Editor)

1973-01-01

Design of the 5-kwe reactor thermoelectric system was initiated in February 1972 and extended through the conceptual design phase into the preliminary design phase. Design effort was terminated in January, 1973. This report documents the system and component requirements, design approaches, and performance and design characteristics for the 5-kwe system. Included is summary information on the reactor, radiation shields, power conversion systems, thermoelectric pump, radiator/structure, liquid metal components, and the control system.

Fault-tolerant reactor protection system

DOEpatents

Gaubatz, Donald C.

1997-01-01

A reactor protection system having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Each division performs independently of the others (asynchronous operation). All communications between the divisions are asynchronous. Each chassis substitutes its own spare sensor reading in the 2/3 vote if a sensor reading from one of the other chassis is faulty or missing. Therefore the presence of at least two valid sensor readings in excess of a set point is required before terminating the output to the hardware logic of a scram inhibition signal even when one of the four sensors is faulty or when one of the divisions is out of service.

Moon base reactor system

NASA Technical Reports Server (NTRS)

Chavez, H.; Flores, J.; Nguyen, M.; Carsen, K.

1989-01-01

The objective of our reactor design is to supply a lunar-based research facility with 20 MW(e). The fundamental layout of this lunar-based system includes the reactor, power conversion devices, and a radiator. The additional aim of this reactor is a longevity of 12 to 15 years. The reactor is a liquid metal fast breeder that has a breeding ratio very close to 1.0. The geometry of the core is cylindrical. The metallic fuel rods are of beryllium oxide enriched with varying degrees of uranium, with a beryllium core reflector. The liquid metal coolant chosen was natural lithium. After the liquid metal coolant leaves the reactor, it goes directly into the power conversion devices. The power conversion devices are Stirling engines. The heated coolant acts as a hot reservoir to the device. It then enters the radiator to be cooled and reenters the Stirling engine acting as a cold reservoir. The engines' operating fluid is helium, a highly conductive gas. These Stirling engines are hermetically sealed. Although natural lithium produces a lower breeding ratio, it does have a larger temperature range than sodium. It is also corrosive to steel. This is why the container material must be carefully chosen. One option is to use an expensive alloy of cerbium and zirconium. The radiator must be made of a highly conductive material whose melting point temperature is not exceeded in the reactor and whose structural strength can withstand meteor showers.

Reactor core isolation cooling system

DOEpatents

Cooke, F.E.

1992-12-08

A reactor core isolation cooling system includes a reactor pressure vessel containing a reactor core, a drywell vessel, a containment vessel, and an isolation pool containing an isolation condenser. A turbine is operatively joined to the pressure vessel outlet steamline and powers a pump operatively joined to the pressure vessel feedwater line. In operation, steam from the pressure vessel powers the turbine which in turn powers the pump to pump makeup water from a pool to the feedwater line into the pressure vessel for maintaining water level over the reactor core. Steam discharged from the turbine is channeled to the isolation condenser and is condensed therein. The resulting heat is discharged into the isolation pool and vented to the atmosphere outside the containment vessel for removing heat therefrom. 1 figure.

Reactor core isolation cooling system

DOEpatents

Cooke, Franklin E.

1992-01-01

A reactor core isolation cooling system includes a reactor pressure vessel containing a reactor core, a drywell vessel, a containment vessel, and an isolation pool containing an isolation condenser. A turbine is operatively joined to the pressure vessel outlet steamline and powers a pump operatively joined to the pressure vessel feedwater line. In operation, steam from the pressure vessel powers the turbine which in turn powers the pump to pump makeup water from a pool to the feedwater line into the pressure vessel for maintaining water level over the reactor core. Steam discharged from the turbine is channeled to the isolation condenser and is condensed therein. The resulting heat is discharged into the isolation pool and vented to the atmosphere outside the containment vessel for removing heat therefrom.

Standardized verification of fuel cycle modeling

DOE PAGES

Feng, B.; Dixon, B.; Sunny, E.; ...

2016-04-05

A nuclear fuel cycle systems modeling and code-to-code comparison effort was coordinated across multiple national laboratories to verify the tools needed to perform fuel cycle analyses of the transition from a once-through nuclear fuel cycle to a sustainable potential future fuel cycle. For this verification study, a simplified example transition scenario was developed to serve as a test case for the four systems codes involved (DYMOND, VISION, ORION, and MARKAL), each used by a different laboratory participant. In addition, all participants produced spreadsheet solutions for the test case to check all the mass flows and reactor/facility profiles on a year-by-yearmore » basis throughout the simulation period. The test case specifications describe a transition from the current US fleet of light water reactors to a future fleet of sodium-cooled fast reactors that continuously recycle transuranic elements as fuel. After several initial coordinated modeling and calculation attempts, it was revealed that most of the differences in code results were not due to different code algorithms or calculation approaches, but due to different interpretations of the input specifications among the analysts. Therefore, the specifications for the test case itself were iteratively updated to remove ambiguity and to help calibrate interpretations. In addition, a few corrections and modifications were made to the codes as well, which led to excellent agreement between all codes and spreadsheets for this test case. Although no fuel cycle transition analysis codes matched the spreadsheet results exactly, all remaining differences in the results were due to fundamental differences in code structure and/or were thoroughly explained. As a result, the specifications and example results are provided so that they can be used to verify additional codes in the future for such fuel cycle transition scenarios.« less

sCO2 Brayton Cycle: Roadmap to sCO2 Power Cycles NE Commercial Applications.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mendez Cruz, Carmen Margarita; Rochau, Gary E.

The mission of the Energy Conversion (EC) area of the Advanced Reactor Technology (ART) program is to commercialize the sCO2 Brayton cycle for Advance Reactors and for the Supercritical Transformational Electric Production (STEP) program. The near-term objective of the EC team efforts is to support the development of a commercially scalable Recompression Closed Brayton Cycle (RCBC) to be constructed for the first STEP demonstration system with the lowest risk possible. This document details the status of technology, policy and market considerations, documentation of gaps and needs, and outlines the steps necessary for the successful development and deployment of commercial sCO2more » Brayton Power Systems along the path to nuclear reactor applications. Document Control Version Creation Date Revisions Created By Release Date 1.0 2/29/2016 Preliminary Draft Mendez, C. 3/2/2016 2.0 7/29/2016 Preliminaty/Partial Report -- updated Focus Area structure, added commercial path forward Mendez, C. 8/10/16 3.0 5/1/2018 Updated Roadmap supports timeline changes and inclusion of grid qualification goals Mendez, C. 6/6/18« less

Fission control system for nuclear reactor

DOEpatents

Conley, G.H.; Estes, G.P.

Control system for nuclear reactor comprises a first set of reactivity modifying rods fixed in a reactor core with their upper ends stepped in height across the core, and a second set of reactivity modifying rods movable vertically within the reactor core and having their lower ends stepped to correspond with the stepped arrangement of the first set of rods, pairs of the rods of the first and second sets being in coaxial alignment.

The mechanism of enhanced wastewater nitrogen removal by photo-sequencing batch reactors based on comprehensive analysis of system dynamics within a cycle.

PubMed

Ye, Jianfeng; Liang, Junyu; Wang, Liang; Markou, Giorgos

2018-07-01

To understand the mechanism of enhanced nitrogen removal by photo-sequencing batch reactors (photo-SBRs), which incorporated microalgal photosynthetic oxygenation into the aerobic phases of a conventional cycle, this study performed comprehensive analysis of one-cycle dynamics. Under a low aeration intensity (about 0.02 vvm), a photo-SBR, illuminated with light at 92.27 μ·mol·m -2 ·s -1 , could remove 99.45% COD, 99.93% NH 4 + -N, 90.39% TN, and 95.17% TP, while the control SBR could only remove 98.36% COD, 83.51% NH 4 + -N, 78.96% TN, and 97.75% TP, for a synthetic domestic sewage. The specific oxygen production rate (SOPR) of microalgae in the photo-SBR could reach 6.63 fmol O 2 ·cell -1 ·h -1 . One-cycle dynamics shows that the enhanced nitrogen removal by photo-SBRs is related to photosynthetic oxygenation, resulting in strengthened nitrification, instead of direct nutrient uptake by microalgae. A too high light or aeration intensity could deteriorate anoxic conditions and thus adversely affect the removal of TN and TP in photo-SBRs. Copyright © 2018 Elsevier Ltd. All rights reserved.

Nuclear electric propulsion reactor control systems status

NASA Technical Reports Server (NTRS)

Ferg, D. A.

1973-01-01

The thermionic reactor control system design studies conducted over the past several years for a nuclear electric propulsion system are described and summarized. The relevant reactor control system studies are discussed in qualitative terms, pointing out the significant advantages and disadvantages including the impact that the various control systems would have on the nuclear electric propulsion system design. A recommendation for the reference control system is made, and a program for future work leading to an engineering model is described.

METHOD AND APPARATUS FOR CONTROLLING DIRECT-CYCLE NEUTRONIC REACTORS

DOEpatents

Reed, G.A.

1961-01-10

A control arrangement is offered for a boiling-water reactor. Boric acid is maintained in the water in the reactor and the amount in the reactor is controlled by continuously removing a portion of the water from the reactor, concentrating the boric acid by evaporating the water therefrom, returning a controlled amount of the acid to the reactor, and simultaneously controlling the water level by varying the rate of spent steam return to the reactor.

Microchannel Reactor System for Catalytic Hydrogenation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Adeniyi Lawal; Woo Lee; Ron Besser

2010-12-22

We successfully demonstrated a novel process intensification concept enabled by the development of microchannel reactors, for energy efficient catalytic hydrogenation reactions at moderate temperature, and pressure, and low solvent levels. We designed, fabricated, evaluated, and optimized a laboratory-scale microchannel reactor system for hydrogenation of onitroanisole and a proprietary BMS molecule. In the second phase of the program, as a prelude to full-scale commercialization, we designed and developed a fully-automated skid-mounted multichannel microreactor pilot plant system for multiphase reactions. The system is capable of processing 1 – 10 kg/h of liquid substrate, and an industrially relevant immiscible liquid-liquid was successfully demonstratedmore » on the system. Our microreactor-based pilot plant is one-of-akind. We anticipate that this process intensification concept, if successfully demonstrated, will provide a paradigm-changing basis for replacing existing energy inefficient, cost ineffective, environmentally detrimental slurry semi-batch reactor-based manufacturing practiced in the pharmaceutical and fine chemicals industries.« less

Thorium Fuel Utilization Analysis on Small Long Life Reactor for Different Coolant Types

NASA Astrophysics Data System (ADS)

Permana, Sidik

2017-07-01

A small power reactor and long operation which can be deployed for less population and remote area has been proposed by the IAEA as a small and medium reactor (SMR) program. Beside uranium utilization, it can be used also thorium fuel resources for SMR as a part of optimalization of nuclear fuel as a “partner” fuel with uranium fuel. A small long-life reactor based on thorium fuel cycle for several reactor coolant types and several power output has been evaluated in the present study for 10 years period of reactor operation. Several key parameters are used to evaluate its effect to the reactor performances such as reactor criticality, excess reactivity, reactor burnup achievement and power density profile. Water-cooled types give higher criticality than liquid metal coolants. Liquid metal coolant for fast reactor system gives less criticality especially at beginning of cycle (BOC), which shows liquid metal coolant system obtains almost stable criticality condition. Liquid metal coolants are relatively less excess reactivity to maintain longer reactor operation than water coolants. In addition, liquid metal coolant gives higher achievable burnup than water coolant types as well as higher power density for liquid metal coolants.

JEN-1 Reactor Control System; SISTEMA DE CONTROL DEL REACTOR JEN-1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cantillo, M.F.; Nuno, C.M.; Andreu, J.L.M.

1963-01-01

ABS>The JEN-1 3Mw power swimming pool reactor electrical control circuits are described. Start-up, power generation in the core, and shutdown are controlled by the reactor control system. This control system guarantees in each moment the safety conditions during reactor operation. Each circuit was represented by a scheme, complemented with a description of its function, components, and operation theory. Components described include: scram circuit; fission counter control circuit; servo control circuit; control circuit of safety sheets; control circuits of primary, secondary, and clean-up pump motors and tower fan motor; primary valve motor circuit; center cubicle alarm circuit; and process alarm circuit.more » (auth)« less

Fault-tolerant reactor protection system

DOEpatents

Gaubatz, D.C.

1997-04-15

A reactor protection system is disclosed having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Each division performs independently of the others (asynchronous operation). All communications between the divisions are asynchronous. Each chassis substitutes its own spare sensor reading in the 2/3 vote if a sensor reading from one of the other chassis is faulty or missing. Therefore the presence of at least two valid sensor readings in excess of a set point is required before terminating the output to the hardware logic of a scram inhibition signal even when one of the four sensors is faulty or when one of the divisions is out of service. 16 figs.

Performance evaluation of two-stage fuel cycle from SFR to PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fei, T.; Hoffman, E.A.; Kim, T.K.

2013-07-01

One potential fuel cycle option being considered is a two-stage fuel cycle system involving the continuous recycle of transuranics in a fast reactor and the use of bred plutonium in a thermal reactor. The first stage is a Sodium-cooled Fast Reactor (SFR) fuel cycle with metallic U-TRU-Zr fuel. The SFRs need to have a breeding ratio greater than 1.0 in order to produce fissile material for use in the second stage. The second stage is a PWR fuel cycle with uranium and plutonium mixed oxide fuel based on the design and performance of the current state-of-the-art commercial PWRs with anmore » average discharge burnup of 50 MWd/kgHM. This paper evaluates the possibility of this fuel cycle option and discusses its fuel cycle performance characteristics. The study focuses on an equilibrium stage of the fuel cycle. Results indicate that, in order to avoid a positive coolant void reactivity feedback in the stage-2 PWR, the reactor requires high quality of plutonium from the first stage and minor actinides in the discharge fuel of the PWR needs to be separated and sent back to the stage-1 SFR. The electricity-sharing ratio between the 2 stages is 87.0% (SFR) to 13.0% (PWR) for a TRU inventory ratio (the mass of TRU in the discharge fuel divided by the mass of TRU in the fresh fuel) of 1.06. A sensitivity study indicated that by increasing the TRU inventory ratio to 1.13, The electricity generation fraction of stage-2 PWR is increased to 28.9%. The two-stage fuel cycle system considered in this study was found to provide a high uranium utilization (>80%). (authors)« less

Design of a 25-kWe Surface Reactor System Based on SNAP Reactor Technologies

NASA Astrophysics Data System (ADS)

Dixon, David D.; Hiatt, Matthew T.; Poston, David I.; Kapernick, Richard J.

2006-01-01

A Hastelloy-X clad, sodium-potassium (NaK-78) cooled, moderated spectrum reactor using uranium zirconium hydride (UZrH) fuel based on the SNAP program reactors is a promising design for use in surface power systems. This paper presents a 98 kWth reactor for a power system the uses multiple Stirling engines to produce 25 kWe-net for 5 years. The design utilizes a pin type geometry containing UZrHx fuel clad with Hastelloy-X and NaK-78 flowing around the pins as coolant. A compelling feature of this design is its use of 49.9% enriched U, allowing it to be classified as a category III-D attractiveness and reducing facility costs relative to highly-enriched space reactor concepts. Presented below are both the design and an analysis of this reactor's criticality under various safety and operations scenarios.

Assessment of Possible Cycle Lengths for Fully-Ceramic Micro-Encapsulated Fuel-Based Light Water Reactor Concepts

DOE Office of Scientific and Technical Information (OSTI.GOV)

R. Sonat Sen; Michael A. Pope; Abderrafi M. Ougouag

2012-04-01

The tri-isotropic (TRISO) fuel developed for High Temperature reactors is known for its extraordinary fission product retention capabilities [1]. Recently, the possibility of extending the use of TRISO particle fuel to Light Water Reactor (LWR) technology, and perhaps other reactor concepts, has received significant attention [2]. The Deep Burn project [3] currently focuses on once-through burning of transuranic fissile and fissionable isotopes (TRU) in LWRs. The fuel form for this purpose is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the TRISO fuel particle design from high temperature reactor technology, but uses SiC as a matrix material rather thanmore » graphite. In addition, FCM fuel may also use a cladding made of a variety of possible material, again including SiC as an admissible choice. The FCM fuel used in the Deep Burn (DB) project showed promising results in terms of fission product retention at high burnup values and during high-temperature transients. In the case of DB applications, the fuel loading within a TRISO particle is constituted entirely of fissile or fissionable isotopes. Consequently, the fuel was shown to be capable of achieving reasonable burnup levels and cycle lengths, especially in the case of mixed cores (with coexisting DB and regular LWR UO2 fuels). In contrast, as shown below, the use of UO2-only FCM fuel in a LWR results in considerably shorter cycle length when compared to current-generation ordinary LWR designs. Indeed, the constraint of limited space availability for heavy metal loading within the TRISO particles of FCM fuel and the constraint of low (i.e., below 20 w/0) 235U enrichment combine to result in shorter cycle lengths compared to ordinary LWRs if typical LWR power densities are also assumed and if typical TRISO particle dimensions and UO2 kernels are specified. The primary focus of this summary is on using TRISO particles with up to 20 w/0 enriched uranium kernels loaded in

Passive cooling safety system for liquid metal cooled nuclear reactors

DOEpatents

Hunsbedt, Anstein; Boardman, Charles E.; Hui, Marvin M.; Berglund, Robert C.

1991-01-01

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel. The passive cooling system includes a closed primary fluid circuit through the partitions surrounding the reactor vessel and a partially adjoining secondary open fluid circuit for carrying transferred heat out into the atmosphere.

Indirect passive cooling system for liquid metal cooled nuclear reactors

DOEpatents

Hunsbedt, Anstein; Boardman, Charles E.

1990-01-01

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel. The passive cooling system includes a closed primary fluid circuit through the partitions surrounding the reactor vessel and a partially adjoining secondary open fluid circuit for carrying transferred heat out into the atmosphere.

Uniaxial low cycle fatigue behavior for pre-corroded 16MND5 bainitic steel in simulated pressurized water reactor environment

NASA Astrophysics Data System (ADS)

Chen, Xu; Ren, Bin; Yu, Dunji; Xu, Bin; Zhang, Zhe; Chen, Gang

2018-06-01

The effects of uniaxial tension properties and low cycle fatigue behavior of 16MND5 bainitic steel cylinder pre-corroded in simulated pressurized water reactor (PWR) were investigated by fatigue at room temperature in air and immersion test system, scanning electron microscopy (SEM), energy disperse spectroscopy (EDS). The experimental results indicated that the corrosion fatigue lives of 16MND5 specimen were significantly affected by the strain amplitude and simulated PWR environments. The compositions of corrosion products were complexly formed in simulated PWR environments. The porous corrosion surface of pre-corroded materials tended to generate pits as a result of promoting contact area to the fresh metal, which promoted crack initiation. For original materials, the fatigue cracks initiated at inclusions imbedded in the micro-cracks. Moreover, the simulated PWR environments degraded the mechanical properties and low cycle fatigue behavior of 16MND5 specimens remarkably. Pre-corrosion of 16MND5 specimen mainly affected the plastic term of the Coffin-Manson equation.

Sensitivity Analysis and Optimization of the Nuclear Fuel Cycle: A Systematic Approach

NASA Astrophysics Data System (ADS)

Passerini, Stefano

For decades, nuclear energy development was based on the expectation that recycling of the fissionable materials in the used fuel from today's light water reactors into advanced (fast) reactors would be implemented as soon as technically feasible in order to extend the nuclear fuel resources. More recently, arguments have been made for deployment of fast reactors in order to reduce the amount of higher actinides, hence the longevity of radioactivity, in the materials destined to a geologic repository. The cost of the fast reactors, together with concerns about the proliferation of the technology of extraction of plutonium from used LWR fuel as well as the large investments in construction of reprocessing facilities have been the basis for arguments to defer the introduction of recycling technologies in many countries including the US. In this thesis, the impacts of alternative reactor technologies on the fuel cycle are assessed. Additionally, metrics to characterize the fuel cycles and systematic approaches to using them to optimize the fuel cycle are presented. The fuel cycle options of the 2010 MIT fuel cycle study are re-examined in light of the expected slower rate of growth in nuclear energy today, using the CAFCA (Code for Advanced Fuel Cycle Analysis). The Once Through Cycle (OTC) is considered as the base-line case, while advanced technologies with fuel recycling characterize the alternative fuel cycle options available in the future. The options include limited recycling in L WRs and full recycling in fast reactors and in high conversion LWRs. Fast reactor technologies studied include both oxide and metal fueled reactors. Additional fuel cycle scenarios presented for the first time in this work assume the deployment of innovative recycling reactor technologies such as the Reduced Moderation Boiling Water Reactors and Uranium-235 initiated Fast Reactors. A sensitivity study focused on system and technology parameters of interest has been conducted to test

Modeling and analysis of tritium dynamics in a DT fusion fuel cycle

NASA Astrophysics Data System (ADS)

Kuan, William

1998-11-01

A number of crucial design issues have a profound effect on the dynamics of the tritium fuel cycle in a DT fusion reactor, where the development of appropriate solutions to these issues is of particular importance to the introduction of fusion as a commercial system. Such tritium-related issues can be classified according to their operational, safety, and economic impact to the operation of the reactor during its lifetime. Given such key design issues inherent in next generation fusion devices using the DT fuel cycle development of appropriate models can then lead to optimized designs of the fusion fuel cycle for different types of DT fusion reactors. In this work, two different types of modeling approaches are developed and their application to solving key tritium issues presented. For the first approach, time-dependent inventories, concentrations, and flow rates characterizing the main subsystems of the fuel cycle are simulated with a new dynamic modular model of a fusion reactor's fuel cycle, named X-TRUFFLES (X-Windows TRitiUm Fusion Fuel cycLE dynamic Simulation). The complex dynamic behavior of the recycled fuel within each of the modeled subsystems is investigated using this new integrated model for different reactor scenarios and design approaches. Results for a proposed fuel cycle design taking into account current technologies are presented, including sensitivity studies. Ways to minimize the tritium inventory are also assessed by examining various design options that could be used to minimize local and global tritium inventories. The second modeling approach involves an analytical model to be used for the calculation of the required tritium breeding ratio, i.e., a primary design issue which relates directly to the feasibility and economics of DT fusion systems. A time-integrated global tritium balance scheme is developed and appropriate analytical expressions are derived for tritium self-sufficiency relevant parameters. The easy exploration of the large

Hybrid Plasma Reactor/Filter for Transportable Collective Protection Systems

DTIC Science & Technology

2011-03-01

protection. The key premise of the hybrid system is to couple a nonthermal plasma (NTP) reactor with reactive adsorption to provide a broader envelope of...conventional methods for collective protection. The key premise of the hybrid system is to couple a nonthermal plasma (NTP) reactor with reactive adsorption to...protection. The key premise of the hybrid system is to couple a nonthermal plasma (NTP) reactor with reactive adsorption to provide a broader

Evaluation and Optimization of a Supercritical Carbon Dioxide Power Conversion Cycle for Nuclear Applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Edwin A. Harvego; Michael G. McKellar

2011-05-01

There have been a number of studies involving the use of gases operating in the supercritical mode for power production and process heat applications. Supercritical carbon dioxide (CO2) is particularly attractive because it is capable of achieving relatively high power conversion cycle efficiencies in the temperature range between 550°C and 750°C. Therefore, it has the potential for use with any type of high-temperature nuclear reactor concept, assuming reactor core outlet temperatures of at least 550°C. The particular power cycle investigated in this paper is a supercritical CO2 Recompression Brayton Cycle. The CO2 Recompression Brayton Cycle can be used as eithermore » a direct or indirect power conversion cycle, depending on the reactor type and reactor outlet temperature. The advantage of this cycle when compared to the helium Brayton Cycle is the lower required operating temperature; 550°C versus 850°C. However, the supercritical CO2 Recompression Brayton Cycle requires an operating pressure in the range of 20 MPa, which is considerably higher than the required helium Brayton cycle operating pressure of 8 MPa. This paper presents results of analyses performed using the UniSim process analyses software to evaluate the performance of the supercritical CO2 Brayton Recompression Cycle for different reactor outlet temperatures. The UniSim model assumed a 600 MWt reactor power source, which provides heat to the power cycle at a maximum temperature of between 550°C and 750°C. The UniSim model used realistic component parameters and operating conditions to model the complete power conversion system. CO2 properties were evaluated, and the operating range for the cycle was adjusted to take advantage of the rapidly changing conditions near the critical point. The UniSim model was then optimized to maximize the power cycle thermal efficiency at the different maximum power cycle operating temperatures. The results of the analyses showed that power cycle thermal

Reactor control rod timing system

DOEpatents

Wu, Peter T. K.

1982-01-01

A fluid driven jet-edge whistle timing system for control rods of a nuclear reactor for producing real-time detection of the timing of each control rod in its scram operation. An important parameter in reactor safety, particularly for liquid metal fast breeder reactors (LMFBR), is the time deviation between the time the control rod is released and the time the rod actually reaches the down position. The whistle has a nearly pure tone signal with center frequency (above 100 kHz) far above the frequency band in which the energy of the background noise is concentrated. Each control rod can be fitted with a whistle with a different frequency so that there is no ambiguity in differentiating the signal from each control rod.

Small reactor power system for space application

NASA Technical Reports Server (NTRS)

Shirbacheh, M.

1987-01-01

A development history and comparative performance capability evaluation is presented for spacecraft nuclear powerplant Small Reactor Power System alternatives. The choice of power conversion technology depends on the reactor's operating temperature; thermionic, thermoelectric, organic Rankine, and Alkali metal thermoelectric conversion are the primary power conversion subsystem technology alternatives. A tabulation is presented for such spacecraft nuclear reactor test histories as those of SNAP-10A, SP-100, and NERVA.

Nuclear reactor sealing system

DOEpatents

McEdwards, James A.

1983-01-01

A liquid metal-cooled nuclear reactor sealing system. The nuclear reactor includes a vessel sealed at its upper end by a closure head. The closure head comprises at least two components, one of which is rotatable; and the two components define an annulus therebetween. The sealing system includes at least a first and second inflatable seal disposed in series in an upper portion of the annulus. The system further includes a dip seal extending into a body of insulation located adjacent a bottom portion of the closure head. The dip seal comprises a trough formed by a lower portion of one of the components, and a seal blade pendently supported from the other component and extending downwardly into the trough. A body of liquid metal is contained in the trough which submerges a portion of the seal blade. The seal blade is provided with at least one aperture located above the body of liquid metal for providing fluid communication between the annulus intermediate the dip seal and the inflatable seals, and a body of cover gas located inside the vessel. There also is provided means for introducing a purge gas into the annulus intermediate the inflatable seals and the seal blade. The purge gas is introduced in an amount sufficient to substantially reduce diffusion of radioactive cover gas or sodium vapor up to the inflatable seals. The purge gas mixes with the cover gas in the reactor vessel where it can be withdrawn from the vessel for treatment and recycle to the vessel.

Effect of temperature and cycle length on microbial competition in PHB-producing sequencing batch reactor.

PubMed

Jiang, Yang; Marang, Leonie; Kleerebezem, Robbert; Muyzer, Gerard; van Loosdrecht, Mark C M

2011-05-01

The impact of temperature and cycle length on microbial competition between polyhydroxybutyrate (PHB)-producing populations enriched in feast-famine sequencing batch reactors (SBRs) was investigated at temperatures of 20 °C and 30 °C, and in a cycle length range of 1-18 h. In this study, the microbial community structure of the PHB-producing enrichments was found to be strongly dependent on temperature, but not on cycle length. Zoogloea and Plasticicumulans acidivorans dominated the SBRs operated at 20 °C and 30 °C, respectively. Both enrichments accumulated PHB more than 75% of cell dry weight. Short-term temperature change experiments revealed that P. acidivorans was more temperature sensitive as compared with Zoogloea. This is particularly true for the PHB degradation, resulting in incomplete PHB degradation in P. acidivorans at 20 °C. Incomplete PHB degradation limited biomass growth and allowed Zoogloea to outcompete P. acidivorans. The PHB content at the end of the feast phase correlated well with the cycle length at a constant solid retention time (SRT). These results suggest that to establish enrichment with the capacity to store a high fraction of PHB, the number of cycles per SRT should be minimized independent of the temperature.

REACTOR - a Concept for establishing a System-of-Systems

NASA Astrophysics Data System (ADS)

Haener, Rainer; Hammitzsch, Martin; Wächter, Joachim

2014-05-01

REACTOR is a working title for activities implementing reliable, emergent, adaptive, and concurrent collaboration on the basis of transactional object repositories. It aims at establishing federations of autonomous yet interoperable systems (Systems-of-Systems), which are able to expose emergent behaviour. Following the principles of event-driven service-oriented architectures (SOA 2.0), REACTOR enables adaptive re-organisation by dynamic delegation of responsibilities and novel yet coherent monitoring strategies by combining information from different domains. Thus it allows collaborative decision-processes across system, discipline, and administrative boundaries. Interoperability is based on two approaches that implement interconnection and communication between existing heterogeneous infrastructures and information systems: Coordinated (orchestration-based) communication and publish/subscribe (choreography-based) communication. Choreography-based communication ensures the autonomy of the participating systems to the highest possible degree but requires the implementation of adapters, which provide functional access to information (publishing/consuming events) via a Message Oriented Middleware (MOM). Any interconnection of the systems (composition of service and message cascades) is established on the basis of global conversations that are enacted by choreographies specifying the expected behaviour of the participating systems with respect to agreed Service Level Agreements (SLA) required by e.g. national authorities. The specification of conversations, maintained in commonly available repositories also enables the utilisation of systems for purposes (evolving) other than initially intended. Orchestration-based communication additionally requires a central component that controls the information transfer via service requests or event processing and also takes responsibility of managing business processes. Commonly available transactional object repositories are

Modeling and simulation of CANDU reactor and its regulating system

NASA Astrophysics Data System (ADS)

Javidnia, Hooman

Analytical computer codes are indispensable tools in design, optimization, and control of nuclear power plants. Numerous codes have been developed to perform different types of analyses related to the nuclear power plants. A large number of these codes are designed to perform safety analyses. In the context of safety analyses, the control system is often neglected. Although there are good reasons for such a decision, that does not mean that the study of control systems in the nuclear power plants should be neglected altogether. In this thesis, a proof of concept code is developed as a tool that can be used in the design. optimization. and operation stages of the control system. The main objective in the design of this computer code is providing a tool that is easy to use by its target audience and is capable of producing high fidelity results that can be trusted to design the control system and optimize its performance. Since the overall plant control system covers a very wide range of processes, in this thesis the focus has been on one particular module of the the overall plant control system, namely, the reactor regulating system. The center of the reactor regulating system is the CANDU reactor. A nodal model for the reactor is used to represent the spatial neutronic kinetics of the core. The nodal model produces better results compared to the point kinetics model which is often used in the design and analysis of control system for nuclear reactors. The model can capture the spatial effects to some extent. although it is not as detailed as the finite difference methods. The criteria for choosing a nodal model of the core are: (1) the model should provide more detail than point kinetics and capture spatial effects, (2) it should not be too complex or overly detailed to slow down the simulation and provide details that are extraneous or unnecessary for a control engineer. Other than the reactor itself, there are auxiliary models that describe dynamics of different

Lunar Regolith Simulant Feed System for a Hydrogen Reduction Reactor System

NASA Technical Reports Server (NTRS)

Mueller, R. P.; Townsend, Ivan I., III

2009-01-01

One of the goals of In-Situ Resource Utilization (ISRU) on the moon is to produce oxygen from the lunar regolith which is present in the form of Ilmenite (FeTi03) and other compounds. A reliable and attainable method of extracting some of the oxygen from the lunar regolith is to use the hydrogen reduction process in a hot reactor to create water vapor which is then condensed and electrolyzed to obtain oxygen for use as a consumable. One challenge for a production system is to reliably acquire the regolith with an excavator hauler mobility platform and then introduce it into the reactor inlet tube which is raised from the surface and above the reactor itself. After the reaction, the hot regolith (-1000 C) must be expelled from the reactor for disposal by the excavator hauler mobility system. In addition, the reactor regolith inlet and outlet tubes must be sealed by valves during the reaction in order to allow collection of the water vapor by the chemical processing sub-system. These valves must be able to handle abrasive regolith passing through them as well as the heat conduction from the hot reactor. In 2008, NASA has designed and field tested a hydrogen reduction system called ROxygen in order to demonstrate the feasibility of extracting oxygen from lunar regolith. The field test was performed with volcanic ash known as Tephra on Mauna Kea volcano on the Big Island of Hawai'i. The tephra has similar properties to lunar regolith, so that it is regarded as a good simulant for the hydrogen reduction process. This paper will discuss the design, fabrication, operation, test results and lessons learned with the ROxygen regolith feed system as tested on Mauna Kea in November 2008.

Effect of cycle time on polyhydroxybutyrate (PHB) production in aerobic mixed cultures.

PubMed

Ozdemir, Sebnem; Akman, Dilek; Cirik, Kevser; Cinar, Ozer

2014-03-01

The aim of this study was to investigate the effect of cycle time on polyhydroxybutyrate (PHB) production under aerobic dynamic feeding system. The acetate-fed feast and famine sequencing batch reactor was used to enrich PHB accumulating microorganism. Sequencing batch reactor (SBR) was operated in four different cycle times (12, 8, 4, and 2 h) fed with a synthetic wastewater. The system performance was determined by monitoring total dissolved organic carbon, dissolved oxygen, oxidation-reduction potential, and PHB concentration. In this study, under steady-state conditions, the feast period of the SBR was found to allow the PHB storage while a certain part of stored PHB was used for continued growth in famine period. The percentage PHB storages by aerobic microorganism were at 16, 18, 42, and 55% for the 12, 8, 4, and 2-h cycle times, respectively. The PHB storage was increased as the length of the cycle time was decreased, and the ratio of the feast compared to the total cycle length was increased from around 13 to 33% for the 12 and 2-h cycle times, respectively.

Novel, Integrated Reactor / Power Conversion System (LMR-AMTEC)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pablo Rubiolo, Principal Investigator

2003-03-21

The main features of this project were the development of a long life (up to 10 years) Liquid Metal Reactor (LMR) and a static conversion subsystem comprising an Alkali Metal Thermal-to-Electric (AMTEC) topping cycle and a ThermoElectric (TE) Bottom cycle. Various coupling options of the LMR with the energy conversion subsystem were explored and, base in the performances found in this analysis, an Indirect Coupling (IC) between the LMR and the AMTEC/TE converters with Alkali Metal Boilers (AMB) was chosen as the reference design. The performance model of the fully integrated sodium-and potassium-AMTEC/TE converters shows that a combined conversion efficiencymore » in excess of 30% could be achieved by the plant. (B204)« less

Development of toroid-type HTS DC reactor series for HVDC system

NASA Astrophysics Data System (ADS)

Kim, Kwangmin; Go, Byeong-Soo; Park, Hea-chul; Kim, Sung-kyu; Kim, Seokho; Lee, Sangjin; Oh, Yunsang; Park, Minwon; Yu, In-Keun

2015-11-01

This paper describes design specifications and performance of a toroid-type high-temperature superconducting (HTS) DC reactor. The first phase operation targets of the HTS DC reactor were 400 mH and 400 A. The authors have developed a real HTS DC reactor system during the last three years. The HTS DC reactor was designed using 2G GdBCO HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. The total system has been successfully developed and tested in connection with LCC type HVDC system. Now, the authors are studying a 400 mH, kA class toroid-type HTS DC reactor for the next phase research. The 1500 A class DC reactor system was designed using layered 13 mm GdBCO 2G HTS wire. The expected operating temperature is under 30 K. These fundamental data obtained through both works will usefully be applied to design a real toroid-type HTS DC reactor for grid application.

Microbial ureolysis in the seawater-catalysed urine phosphorus recovery system: Kinetic study and reactor verification.

PubMed

Tang, Wen-Tao; Dai, Ji; Liu, Rulong; Chen, Guang-Hao

2015-12-15

Our previous study has confirmed the feasibility of using seawater as an economical precipitant for urine phosphorus (P) precipitation. However, we still understand very little about the ureolysis in the Seawater-based Urine Phosphorus Recovery (SUPR) system despite its being a crucial step for urine P recovery. In this study, batch experiments were conducted to investigate the kinetics of microbial ureolysis in the seawater-urine system. Indigenous bacteria from urine and seawater exhibited relatively low ureolytic activity, but they adapted quickly to the urine-seawater mixture during batch cultivation. During cultivation, both the abundance and specific ureolysis rate of the indigenous bacteria were greatly enhanced as confirmed by a biomass-dependent Michaelis-Menten model. The period for fully ureolysis was decreased from 180 h to 2.5 h after four cycles of cultivation. Based on the successful cultivation, a lab-scale SUPR reactor was set up to verify the fast ureolysis and efficient P recovery in the SUPR system. Nearly complete urine P removal was achieved in the reactor in 6 h without adding any chemicals. Terminal Restriction Fragment Length Polymorphism (TRFLP) analysis revealed that the predominant groups of bacteria in the SUPR reactor likely originated from seawater rather than urine. Moreover, batch tests confirmed the high ureolysis rates and high phosphorus removal efficiency induced by cultivated bacteria in the SUPR reactor under seawater-to-urine mixing ratios ranging from 1:1 to 9:1. This study has proved that the enrichment of indigenous bacteria in the SUPR system can lead to sufficient ureolytic activity for phosphate precipitation, thus providing an efficient and economical method for urine P recovery. Copyright © 2015 Elsevier Ltd. All rights reserved.

Off-design performance of a chemical looping combustion (CLC) combined cycle: effects of ambient temperature

NASA Astrophysics Data System (ADS)

Chi, Jinling; Wang, Bo; Zhang, Shijie; Xiao, Yunhan

2010-02-01

The present work investigates the influence of ambient temperature on the steady-state off-design thermodynamic performance of a chemical looping combustion (CLC) combined cycle. A sensitivity analysis of the CLC reactor system was conducted, which shows that the parameters that influence the temperatures of the CLC reactors most are the flow rate and temperature of air entering the air reactor. For the ambient temperature variation, three off-design control strategies have been assumed and compared: 1) without any Inlet Guide Vane (IGV) control, 2) IGV control to maintain air reactor temperature and 3) IGV control to maintain constant fuel reactor temperature, aside from fuel flow rate adjusting. Results indicate that, compared with the conventional combined cycle, due to the requirement of pressure balance at outlet of the two CLC reactors, CLC combined cycle shows completely different off-design thermodynamic characteristics regardless of the control strategy adopted. For the first control strategy, temperatures of the two CLC reactors both rise obviously as ambient temperature increases. IGV control adopted by the second and the third strategy has the effect to maintain one of the two reactors' temperatures at design condition when ambient temperature is above design point. Compare with the second strategy, the third would induce more severe decrease of efficiency and output power of the CLC combined cycle.

Liquid fuel molten salt reactors for thorium utilization

DOE PAGES

Gehin, Jess C.; Powers, Jeffrey J.

2016-04-08

Molten salt reactors (MSRs) represent a class of reactors that use liquid salt, usually fluoride- or chloride-based, as either a coolant with a solid fuel (such as fluoride salt-cooled high temperature reactors) or as a combined coolant and fuel with fuel dissolved in a carrier salt. For liquid-fuelled MSRs, the salt can be processed online or in a batch mode to allow for removal of fission products as well as introduction of fissile fuel and fertile materials during reactor operation. The MSR is most commonly associated with the 233U/thorium fuel cycle, as the nuclear properties of 233U combined with themore » online removal of parasitic absorbers allow for the ability to design a thermal-spectrum breeder reactor; however, MSR concepts have been developed using all neutron energy spectra (thermal, intermediate, fast, and mixed-spectrum zoned concepts) and with a variety of fuels including uranium, thorium, plutonium, and minor actinides. Early MSR work was supported by a significant research and development (R&D) program that resulted in two experimental systems operating at ORNL in the 1960s, the Aircraft Reactor Experiment and the Molten Salt Reactor Experiment. Subsequent design studies in the 1970s focusing on thermal-spectrum thorium-fueled systems established reference concepts for two major design variants: (1) a molten salt breeder reactor (MSBR), with multiple configurations that could breed additional fissile material or maintain self-sustaining operation; and (2) a denatured molten salt reactor (DMSR) with enhanced proliferation-resistance. T MSRs has been selected as one of six most promising Generation IV systems and development activities have been seen in fast-spectrum MSRs, waste-burning MSRs, MSRs fueled with low-enriched uranium (LEU), as well as more traditional thorium fuel cycle-based MSRs. This study provides an historical background of MSR R&D efforts, surveys and summarizes many of the recent development, and provides analysis comparing

New reactor technology: safety improvements in nuclear power systems.

PubMed

Corradini, M L

2007-11-01

Almost 450 nuclear power plants are currently operating throughout the world and supplying about 17% of the world's electricity. These plants perform safely, reliably, and have no free-release of byproducts to the environment. Given the current rate of growth in electricity demand and the ever growing concerns for the environment, nuclear power can only satisfy the need for electricity and other energy-intensive products if it can demonstrate (1) enhanced safety and system reliability, (2) minimal environmental impact via sustainable system designs, and (3) competitive economics. The U.S. Department of Energy with the international community has begun research on the next generation of nuclear energy systems that can be made available to the market by 2030 or earlier, and that can offer significant advances toward these challenging goals; in particular, six candidate reactor system designs have been identified. These future nuclear power systems will require advances in materials, reactor physics, as well as thermal-hydraulics to realize their full potential. However, all of these designs must demonstrate enhanced safety above and beyond current light water reactor systems if the next generation of nuclear power plants is to grow in number far beyond the current population. This paper reviews the advanced Generation-IV reactor systems and the key safety phenomena that must be considered to guarantee that enhanced safety can be assured in future nuclear reactor systems.

Nuclear reactor cavity floor passive heat removal system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Edwards, Tyler A.; Neeley, Gary W.; Inman, James B.

A nuclear reactor includes a reactor core disposed in a reactor pressure vessel. A radiological containment contains the nuclear reactor and includes a concrete floor located underneath the nuclear reactor. An ex vessel corium retention system includes flow channels embedded in the concrete floor located underneath the nuclear reactor, an inlet in fluid communication with first ends of the flow channels, and an outlet in fluid communication with second ends of the flow channels. In some embodiments the inlet is in fluid communication with the interior of the radiological containment at a first elevation and the outlet is in fluidmore » communication with the interior of the radiological containment at a second elevation higher than the first elevation. The radiological containment may include a reactor cavity containing a lower portion of the pressure vessel, wherein the concrete floor located underneath the nuclear reactor is the reactor cavity floor.« less

The Rockwell SR-100G reactor turboelectric space power system

NASA Technical Reports Server (NTRS)

Anderson, R. V.

1985-01-01

During FY 1982 and 1983, Rockwell International performed system and subsystem studies for space reactor power systems. These studies drew on the expertise gained from the design and flight of the SNAP-10A space nuclear reactor system. These studies, performed for the SP-100 Program, culminated in the selection of a reactor-turboelectric (gas Brayton) system for the SP-100 application; this system is called the SR-100G. This paper describes the features of the system and provides references where more detailed information can be obtained.

Fast Flux Test Facility thermal and pressure transient events during Cycle 11

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ahrens, D. M.

1992-03-01

This report documents the thermal and pressure transients experienced by the Reactor Heat Transport System (RHTS) during Cycle 11 which included Cycles 11A, 11B-1, 11B-2 and 11C (i.e. 4 startups and 4 shutdowns). Cycle 11 consisted of a refueling period that began on March 14, 1989 and power operation which began on May 3, 1989 and ended on October 27, 1990. Transients resulted from secondary pump starts/stops while at refueling conditions. The major causes of transients at power were five unplanned reactor scrams from 100% power and problems with Loop 2 DHX Fan Controls During 11A.

SP-100 program: Space reactor system and subsystem investigations

NASA Astrophysics Data System (ADS)

Harty, R. B.

1983-09-01

For a space reactor power system, a comprehensive safety program will be required to assure that no undue risk is present. The nuclear safety review/approval process that is required for a space reactor system is summarized. The documentation requirements are presented along with a summary of the required contents of key documents. Finally, the aerospace safety program conducted for the SNAP-10A reactor system is summarized. The results of this program are presented to show the type of program that is expected and to provide information that could be usable in future programs.

Weld monitor and failure detector for nuclear reactor system

DOEpatents

Sutton, Jr., Harry G.

1987-01-01

Critical but inaccessible welds in a nuclear reactor system are monitored throughout the life of the reactor by providing small aperture means projecting completely through the reactor vessel wall and also through the weld or welds to be monitored. The aperture means is normally sealed from the atmosphere within the reactor. Any incipient failure or cracking of the weld will cause the environment contained within the reactor to pass into the aperture means and thence to the outer surface of the reactor vessel where its presence is readily detected.

Automatic reactor control system for transient operation

NASA Astrophysics Data System (ADS)

Lipinski, Walter C.; Bhattacharyya, Samit K.; Hanan, Nelson A.

Various programmatic considerations have delayed the upgrading of the TREAT reactor and the performance of the control system is not yet experimentally verified. The current schedule calls for the upgrading activities to occur last in the calendar year 1987. Detailed simulation results, coupled with earlier validation of individual components of the control strategy in TREAT, verify the performance of the algorithms. The control system operates within the safety envelope provided by a protection system designed to ensure reactor safety under conditions of spurious reactivity additions. The approach should be directly applicable to MMW systems, with appropriate accounting of temperature rate limitations of key components and of the inertia of the secondary system components.

Emergency heat removal system for a nuclear reactor

DOEpatents

Dunckel, Thomas L.

1976-01-01

A heat removal system for nuclear reactors serving as a supplement to an Emergency Core Cooling System (ECCS) during a Loss of Coolant Accident (LOCA) comprises a plurality of heat pipes having one end in heat transfer relationship with either the reactor pressure vessel, the core support grid structure or other in-core components and the opposite end located in heat transfer relationship with a heat exchanger having heat transfer fluid therein. The heat exchanger is located external to the pressure vessel whereby excessive core heat is transferred from the above reactor components and dissipated within the heat exchanger fluid.

Thorium fueled reactor

NASA Astrophysics Data System (ADS)

Sipaun, S.

2017-01-01

Current development in thorium fueled reactors shows that they can be designed to operate in the fast or thermal spectrum. The thorium/uranium fuel cycle converts fertile thorium-232 into fissile uranium-233, which fissions and releases energy. This paper analyses the characteristics of thorium fueled reactors and discusses the thermal reactor option. It is found that thorium fuel can be utilized in molten salt reactors through many configurations and designs. A balanced assessment on the feasibility of adopting one reactor technology versus another could lead to optimized benefits of having thorium resource.

Microprocessor tester for the treat upgrade reactor trip system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lenkszus, F.R.; Bucher, R.G.

1984-01-01

The upgrading of the Transient Reactor Test (TREAT) Facility at ANL-Idaho has been designed to provide additional experimental capabilities for the study of core disruptive accident (CDA) phenomena. In addition, a programmable Automated Reactor Control System (ARCS) will permit high-power transients up to 11,000 MW having a controlled reactor period of from 15 to 0.1 sec. These modifications to the core neutronics will improve simulation of LMFBR accident conditions. Finally, a sophisticated, multiply-redundant safety system, the Reactor Trip System (RTS), will provide safe operation for both steady state and transient production operating modes. To insure that this complex safety systemmore » is functioning properly, a Dedicated Microprocessor Tester (DMT) has been implemented to perform a thorough checkout of the RTS prior to all TREAT operations.« less

Exhaust system with emissions storage device and plasma reactor

DOEpatents

Hoard, John W.

1998-01-01

An exhaust system for a combustion system, comprising a storage device for collecting NO.sub.x, hydrocarbon, or particulate emissions, or mixture of these emissions, and a plasma reactor for destroying the collected emissions is described. After the emission is collected in by the storage device for a period of time, the emission is then destroyed in a non-thermal plasma generated by the plasma reactor. With respect to the direction of flow of the exhaust stream, the storage device must be located before the terminus of the plasma reactor, and it may be located wholly before, overlap with, or be contained within the plasma reactor.

A liquid-metal filling system for pumped primary loop space reactors

NASA Astrophysics Data System (ADS)

Crandall, D. L.; Reed, W. C.

Some concepts for the SP-100 space nuclear power reactor use liquid metal as the primary coolant in a pumped loop. Prior to filling ground engineering test articles or reactor systems, the liquid metal must be purified and circulated through the reactor primary system to remove contaminants. If not removed, these contaminants enhance corrosion and reduce reliability. A facility was designed and built to support Department of Energy Liquid Metal Fast Breeder Reactor tests conducted at the Idaho National Engineering Laboratory. This test program used liquid sodium to cool nuclear fuel in in-pile experiments; thus, a system was needed to store and purify sodium inventories and fill the experiment assemblies. This same system, with modifications and potential changeover to lithium or sodium-potassium (NaK), can be used in the Space Nuclear Power Reactor Program. This paper addresses the requirements, description, modifications, operation, and appropriateness of using this liquid-metal system to support the SP-100 space reactor program.

SP-100 Program: space reactor system and subsystem investigations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harty, R.B.

1983-09-30

For a space reactor power system, a comprehensive safety program will be required to assure that no undue risk is present. This report summarizes the nuclear safety review/approval process that will be required for a space reactor system. The documentation requirements are presented along with a summary of the required contents of key documents. Finally, the aerospace safety program conducted for the SNAP-10A reactor system is summarized. The results of this program are presented to show the type of program that can be expected and to provide information that could be usable in future programs.

Small reactor power systems for manned planetary surface bases

NASA Technical Reports Server (NTRS)

Bloomfield, Harvey S.

1987-01-01

A preliminary feasibility study of the potential application of small nuclear reactor space power systems to manned planetary surface base missions was conducted. The purpose of the study was to identify and assess the technology, performance, and safety issues associated with integration of reactor power systems with an evolutionary manned planetary surface exploration scenario. The requirements and characteristics of a variety of human-rated modular reactor power system configurations selected for a range of power levels from 25 kWe to hundreds of kilowatts is described. Trade-off analyses for reactor power systems utilizing both man-made and indigenous shielding materials are provided to examine performance, installation and operational safety feasibility issues. The results of this study have confirmed the preliminary feasibility of a wide variety of small reactor power plant configurations for growth oriented manned planetary surface exploration missions. The capability for power level growth with increasing manned presence, while maintaining safe radiation levels, was favorably assessed for nominal 25 to 100 kWe modular configurations. No feasibility limitations or technical barriers were identified and the use of both distance and indigenous planetary soil material for human rated radiation shielding were shown to be viable and attractive options.

Static, Dynamic, and Fatigue Analysis of the Mechanical System of Ultrasonic Scanner for Inservice Inspection of Research Reactors

NASA Astrophysics Data System (ADS)

Awwaluddin, Muhammad; Kristedjo, K.; Handono, Khairul; Ahmad, H.

2018-02-01

This analysis is conducted to determine the effects of static and dynamic loads of the structure of mechanical system of Ultrasonic Scanner i.e., arm, column, and connection systems for inservice inspection of research reactors. The analysis is performed using the finite element method with 520 N static load. The correction factor of dynamic loads used is the Gerber mean stress correction (stress life). The results of the analysis show that the value of maximum equivalent von Mises stress is 1.3698E8 Pa for static loading and value of the maximum equivalent alternating stress is 1.4758E7 Pa for dynamic loading. These values are below the upper limit allowed according to ASTM A240 standards i.e. 2.05E8 Pa. The result analysis of fatigue life cycle are at least 1E6 cycle, so it can be concluded that the structure is in the high life cycle category.

Deployment history and design considerations for space reactor power systems

NASA Astrophysics Data System (ADS)

El-Genk, Mohamed S.

2009-05-01

The history of the deployment of nuclear reactors in Earth orbits is reviewed with emphases on lessons learned and the operation and safety experiences. The former Soviet Union's "BUK" power systems, with SiGe thermoelectric conversion and fast neutron energy spectrum reactors, powered a total of 31 Radar Ocean Reconnaissance Satellites (RORSATs) from 1970 to 1988 in 260 km orbit. Two of the former Soviet Union's TOPAZ reactors, with in-core thermionic conversion and epithermal neutron energy spectrum, powered two Cosmos missions launched in 1987 in ˜800 km orbit. The US' SNAP-10A system, with SiGe energy conversion and a thermal neutron energy spectrum reactor, was launched in 1965 in 1300 km orbit. The three reactor systems used liquid NaK-78 coolant, stainless steel structure and highly enriched uranium fuel (90-96 wt%) and operated at a reactor exit temperature of 833-973 K. The BUK reactors used U-Mo fuel rods, TOPAZ used UO 2 fuel rods and four ZrH moderator disks, and the SNAP-10A used moderated U-ZrH fuel rods. These low power space reactor systems were designed for short missions (˜0.5 kW e and ˜1 year for SNAP-10A, <3.0 kW e and <6 months for BUK, and ˜5.5 kW e and up to 1 year for TOPAZ). The deactivated BUK reactors at the end of mission, which varied in duration from a few hours to ˜4.5 months, were boosted into ˜800 km storage orbit with a decay life of more than 600 year. The ejection of the last 16 BUK reactor fuel cores caused significant contamination of Earth orbits with NaK droplets that varied in sizes from a few microns to 5 cm. Power systems to enhance or enable future interplanetary exploration, in-situ resources utilization on Mars and the Moon, and civilian missions in 1000-3000 km orbits would generate significantly more power of 10's to 100's kW e for 5-10 years, or even longer. A number of design options to enhance the operation reliability and safety of these high power space reactor power systems are presented and discussed.

Scanning tunneling microscope assembly, reactor, and system

DOEpatents

Tao, Feng; Salmeron, Miquel; Somorjai, Gabor A

2014-11-18

An embodiment of a scanning tunneling microscope (STM) reactor includes a pressure vessel, an STM assembly, and three spring coupling objects. The pressure vessel includes a sealable port, an interior, and an exterior. An embodiment of an STM system includes a vacuum chamber, an STM reactor, and three springs. The three springs couple the STM reactor to the vacuum chamber and are operable to suspend the scanning tunneling microscope reactor within the interior of the vacuum chamber during operation of the STM reactor. An embodiment of an STM assembly includes a coarse displacement arrangement, a piezoelectric fine displacement scanning tube coupled to the coarse displacement arrangement, and a receiver. The piezoelectric fine displacement scanning tube is coupled to the coarse displacement arrangement. The receiver is coupled to the piezoelectric scanning tube and is operable to receive a tip holder, and the tip holder is operable to receive a tip.

Reference Reactor Module for the Affordable Fission Surface Power System

NASA Astrophysics Data System (ADS)

Poston, David I.; Kapernick, Richard J.; Dixon, David D.; Amiri, Benjamin W.; Marcille, Thomas F.

2008-01-01

Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The requirements of many surface power applications allow the consideration of systems with much less development risk than most other space reactor applications, because of modest power (10s of kWe) and no driving need for minimal mass (allowing temperatures <1000 K). The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. This paper describes the reference AFSPS reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based, UO2-fueled, liquid metal-cooled fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. One of the important ``affordability'' attributes is that the concept has been designed to minimize both the technical and programmatic safety risk.

Control system for a small fission reactor

DOEpatents

Burelbach, J.P.; Kann, W.J.; Saiveau, J.G.

1985-02-08

A system for controlling the reactivity of a small fission reactor includes an elongated, flexible hollow tube in the general form of a helical coiled spring axially positioned around and outside of the reactor vessel in an annular space between the reactor vessel and a surrounding cylindrical-shaped neutron reflector. A neutron absorbing material is provided within the hollow tube with the rate of the reaction controlled by the extension and compression of the hollow tube, e.g., extension of the tube increases reactivity while its compression reduces reactivity, in varying the amount of neutron absorbing material disposed between the reactor vessel and the neutron reflector. Conventional mechanical displacement means may be employed to control the coil density of the hollow tube as desired.

Increasing the reliability of the shutdown of 500 - 750-kV overhead lines equipped with shunt reactors in an unsuccessful three-phase automatic repeated closure cycle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kuz'micheva, K. I.; Merzlyakov, A. S.; Fokin, G. G.

2013-05-15

The reasons for circuit-breaker failures during repeated disconnection of 500 - 750 kV overhead lines with shunt reactors in a cycle of unsuccessful three-phase automatic reconnection (TARC) are analyzed. Recommendations are made for increasing the operating reliability of power transmission lines with shunt reactors when there is unsuccessful reconnection.

CONCEPTUAL DESIGN OF A LUNAR REGOLITH CLUSTERED-REACTOR SYSTEM

DOE Office of Scientific and Technical Information (OSTI.GOV)

John Darrell Bess

2009-06-01

It is proposed that a fast-fission, heatpipe-cooled, lunar-surface power reactor system be divided into subcritical units that could be launched safely without the incorporation of additional spectral shift absorbers or other complex means of control. The reactor subunits are to be emplaced directly into the lunar regolith utilizing the regolith not just for shielding but as the reflector material to increase the neutron economy of the system. While a single subunit cannot achieve criticality by itself, coordinated placement of additional subunits will provide a critical reactor system for lunar surface power generation. A lunar regolith clustered-reactor system promotes reliability, safety,more » and ease of manufacture and testing at the cost of a slight increase in launch mass per rated power level and an overall reduction in neutron economy when compared to a single-reactor system. Additional subunits may be launched with future missions to increase the cluster size and power according to desired lunar base power demand and lifetime. The results address the potential uncertainties associated with the lunar regolith material and emplacement of the subunit systems. Physical distance between subunits within the clustered emplacement exhibits the most significant feedback regarding changes in overall system reactivity. Narrow, deep holes will be the most effective in reducing axial neutron leakage from the core. The variation in iron concentration in the lunar regolith can directly influence the overall system reactivity although its effects are less than the more dominant factors of subunit emplacement.« less

Reactor control rod timing system. [LMFBR

DOEpatents

Wu, P.T.K.

1980-03-18

A fluid driven jet-edge whistle timing system is described for control rods of a nuclear reactor for producing real-time detection of the timing of each control rod in its scram operation. An important parameter in reactor safety, particularly for liquid metal fast breeder reactors (LMFBR), is the time deviation between the time the control rod is released and the time the rod actually reaches the down position. The whistle has a nearly pure tone signal with center frequency (above 100 kHz) far above the frequency band in which the energy of the background noise is concentrated. Each control rod can be fitted with a whistle with a different frequency so that there is no ambiguity in differentiating the signal from each control rod.

Nuclear reactor pressure vessel support system

DOEpatents

Sepelak, George R.

1978-01-01

A support system for nuclear reactor pressure vessels which can withstand all possible combinations of stresses caused by a postulated core disrupting accident during reactor operation. The nuclear reactor pressure vessel is provided with a flange around the upper periphery thereof, and the flange includes an annular vertical extension formed integral therewith. A support ring is positioned atop of the support ledge and the flange vertical extension, and is bolted to both members. The plug riser is secured to the flange vertical extension and to the top of a radially outwardly extension of the rotatable plug. This system eliminates one joint through which fluids contained in the vessel could escape by making the fluid flow path through the joint between the flange and the support ring follow the same path through which fluid could escape through the plug risers. In this manner, the sealing means to prohibit the escape of contained fluids through the plug risers can also prohibit the escape of contained fluid through the securing joint.

A comparison of production system life cycle models

NASA Astrophysics Data System (ADS)

Attri, Rajesh; Grover, Sandeep

2012-09-01

Companies today need to keep up with the rapidly changing market conditions to stay competitive. The main issues in this paper are related to a company's market and its competitors. The prediction of market behavior is helpful for a manufacturing enterprise to build efficient production systems. However, these predictions are usually not reliable. A production system is required to adapt to changing markets, but such requirement entails higher cost. Hence, analyzing different life cycle models of the production system is necessary. In this paper, different life cycle models of the production system are compared to evaluate the distinctive features and the limitations of each model. Furthermore, the difference between product life cycle and production life cycle is summarized, and the effect of product life cycle on production life cycle is explained. Finally, a production system life cycle model, along with key activities to be performed in each stage, is proposed specifically for the manufacturing sector.

Importance of the (n,gamma) Cm-247 Evaluation on Neutron Emission in Fast Reactor Fuel Cycle Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Benoit Forget; Mehdi Asgari; Rodolfo M. Ferrer

2007-11-01

As part of the GNEP program, it is envisioned to build a fast reactor for the transmutation of minor actinides. The spent nuclear fuel from the current fleet of light water reactors would be recycled, the current baseline is the UREX+1a process, and would act as a feed for the fast reactor. As the fuel is irradiated in a fast reactor a certain quantity of minor actinides would thus build up in the fuel stream creating possible concerns with the neutron emission of these minor actinides for fuel transportation, handling and fabrication. Past neutronic analyses had not tracked minor actinidesmore » above Cm-246 in the transmutation chain, because of the small influence on the overall reactor performance and cycle parameters. However, when trying to quantify the neutron emission from the recycled fuel with high minor actinide content, these higher isotopes play an essential role and should be included in the analysis. In this paper, the influence of tracking these minor actinides on the calculated neutron emission is presented. Also presented is the particular influence of choosing a different evaluated cross section data set to represent the minor actinides above Cm-246. The first representation uses the cross-sections provided by MC2-2 for all isotopes, while the second representation uses infinitely diluted ENDF/BVII.0 cross-sections for Cm-247 to Cf-252 and MC2-2 for all other isotopes.« less

Exploratory study of several advanced nuclear-MHD power plant systems.

NASA Technical Reports Server (NTRS)

Williams, J. R.; Clement, J. D.; Rosa, R. J.; Yang, Y. Y.

1973-01-01

In order for efficient multimegawatt closed cycle nuclear-MHD systems to become practical, long-life gas cooled reactors with exit temperatures of about 2500 K or higher must be developed. Four types of nuclear reactors which have the potential of achieving this goal are the NERVA-type solid core reactor, the colloid core (rotating fluidized bed) reactor, the 'light bulb' gas core reactor, and the 'coaxial flow' gas core reactor. Research programs aimed at developing these reactors have progressed rapidly in recent years so that prototype power reactors could be operating by 1980. Three types of power plant systems which use these reactors have been analyzed to determine the operating characteristics, critical parameters and performance of these power plants. Overall thermal efficiencies as high as 80% are projected, using an MHD turbine-compressor cycle with steam bottoming, and slightly lower efficiencies are projected for an MHD motor-compressor cycle.

High efficiency Brayton cycles using LNG

DOEpatents

Morrow, Charles W [Albuquerque, NM

2006-04-18

A modified, closed-loop Brayton cycle power conversion system that uses liquefied natural gas as the cold heat sink media. When combined with a helium gas cooled nuclear reactor, achievable efficiency can approach 68 76% (as compared to 35% for conventional steam cycle power cooled by air or water). A superheater heat exchanger can be used to exchange heat from a side-stream of hot helium gas split-off from the primary helium coolant loop to post-heat vaporized natural gas exiting from low and high-pressure coolers. The superheater raises the exit temperature of the natural gas to close to room temperature, which makes the gas more attractive to sell on the open market. An additional benefit is significantly reduced costs of a LNG revaporization plant, since the nuclear reactor provides the heat for vaporization instead of burning a portion of the LNG to provide the heat.

Zirconium hydride reactor control reflector systems: summary report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Horton, P.H.; Kurzeka, W.J.

1972-06-30

The beryllium reflector control system development for SNAP reactors is documented, from the initial SNAP 10A System through the current 5-kW(e) Thermoelectric System. Described are the various reflector concepts used in these systems for shadowshielded and 4 pi -shielded nuclear systems. The development of the key components, such as the actuators, bearings, and drive mechanisms for these systems, is also traced from the SNAP 10A concept through to the current system. Developmental test results are outlined, showing the performance capability improvements made throughout the life of the SNAP programs. Component development was highly successful, as proven by a number ofmore » reactor systems tests, including the launch and operation of the SNAP 10A flight. 46 references. (auth)« less

Liquid metal cooled nuclear reactors with passive cooling system

DOEpatents

Hunsbedt, Anstein; Fanning, Alan W.

1991-01-01

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of cooling medium flow circuits which cooperate to remove and carry heat away from the fuel core upon loss of the normal cooling flow circuit to areas external thereto.

Method for passive cooling liquid metal cooled nuclear reactors, and system thereof

DOEpatents

Hunsbedt, Anstein; Busboom, Herbert J.

1991-01-01

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel.

Multi-reactor power system configurations for multimegawatt nuclear electric propulsion

NASA Technical Reports Server (NTRS)

George, Jeffrey A.

1991-01-01

A modular, multi-reactor power system and vehicle configuration for piloted nuclear electric propulsion (NEP) missions to Mars is presented. Such a design could provide enhanced system and mission reliability, allowing a comfortable safety margin for early manned flights, and would allow a range of piloted and cargo missions to be performed with a single power system design. Early use of common power modules for cargo missions would also provide progressive flight experience and validation of standardized systems for use in later piloted applications. System and mission analysis are presented to compare single and multi-reactor configurations for piloted Mars missions. A conceptual design for the Hydra modular multi-reactor NEP vehicle is presented.

Reference reactor module for NASA's lunar surface fission power system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Poston, David I; Kapernick, Richard J; Dixon, David D

Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. The AFSPS concept is now being further developed within the Fission Surface Power (FSP) Project, which is a near-term technology program to demonstrate system-level TRL-6 by 2013. This paper describes the reference FSP reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on themore » lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based. UO{sub 2}-fueled, pumped-NaK fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. The ultimate goal of this work is to provide a 'workhorse' power system that NASA can utilize in near-term and future Lunar and Martian mission architectures, with the eventual capability to evolve to very high power, low mass systems, for either surface, deep space, and/or orbital missions.« less

Corrosion of Structural Materials for Advanced Supercritical Carbon- Dioxide Brayton Cycle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sridharan, Kumar

The supercritical carbon-dioxide (referred to as SC-CO 2 hereon) Brayton cycle is being considered for power conversion systems for a number of nuclear reactor concepts, including the sodium fast reactor (SFR), fluoride saltcooled high temperature reactor (FHR), and high temperature gas reactor (HTGR), and several types of small modular reactors (SMR). The SC-CO 2 direct cycle gas fast reactor has also been recently proposed. The SC-CO 2 Brayton cycle (discussed in Chapter 1) provides higher efficiencies compared to the Rankine steam cycle due to less compression work stemming from higher SC-CO 2 densities, and allows for smaller components size, fewermore » components, and simpler cycle layout. For example, in the case of a SFR using a SC-CO 2 Brayton cycle instead of a steam cycle would also eliminate the possibility of sodium-water interactions. The SC-CO 2 cycle has a higher efficiency than the helium Brayton cycle, with the additional advantage of being able to operate at lower temperatures and higher pressures. In general, the SC-CO 2 Brayton cycle is well-suited for any type of nuclear reactor (including SMR) with core outlet temperature above ~ 500°C in either direct or indirect versions. In all the above applications, materials corrosion in high temperature SC-CO 2 is an important consideration, given their expected lifetimes of 20 years or longer. Our discussions with National Laboratories and private industry early on in this project indicated materials corrosion to be one of the significant gaps in the implementation of SC-CO 2 Brayton cycle. Corrosion can lead to a loss of effective load-bearing wall thickness of a component and can potentially lead to the generation of oxide particulate debris which can lead to three-body wear in turbomachinery components. Another environmental degradation effect that is rather unique to CO 2 environment is the possibility for simultaneous occurrence of carburization during oxidation of the material. Carburization

Dual annular rotating "windowed" nuclear reflector reactor control system

DOEpatents

Jacox, Michael G.; Drexler, Robert L.; Hunt, Robert N. M.; Lake, James A.

1994-01-01

A nuclear reactor control system is provided in a nuclear reactor having a core operating in the fast neutron energy spectrum where criticality control is achieved by neutron leakage. The control system includes dual annular, rotatable reflector rings. There are two reflector rings: an inner reflector ring and an outer reflector ring. The reflectors are concentrically assembled, surround the reactor core, and each reflector ring includes a plurality of openings. The openings in each ring are capable of being aligned or non-aligned with each other. Independent driving means for each of the annular reflector rings is provided so that reactor criticality can be initiated and controlled by rotation of either reflector ring such that the extent of alignment of the openings in each ring controls the reflection of neutrons from the core.

Fast breeder reactor protection system

DOEpatents

van Erp, J.B.

1973-10-01

Reactor protection is provided for a liquid-metal-fast breeder reactor core by measuring the coolant outflow temperature from each of the subassemblies of the core. The outputs of the temperature sensors from a subassembly region of the core containing a plurality of subassemblies are combined in a logic circuit which develops a scram alarm if a predetermined number of the sensors indicate an over temperature condition. The coolant outflow from a single subassembly can be mixed with the coolant outflow from adjacent subassemblies prior to the temperature sensing to increase the sensitivity of the protection system to a single subassembly failure. Coherence between the sensors can be required to discriminate against noise signals. (Official Gazette)

Integrated intelligent systems in advanced reactor control rooms

DOE Office of Scientific and Technical Information (OSTI.GOV)

Beckmeyer, R.R.

1989-01-01

An intelligent, reactor control room, information system is designed to be an integral part of an advanced control room and will assist the reactor operator's decision making process by continuously monitoring the current plant state and providing recommended operator actions to improve that state. This intelligent system is an integral part of, as well as an extension to, the plant protection and control systems. This paper describes the interaction of several functional components (intelligent information data display, technical specifications monitoring, and dynamic procedures) of the overall system and the artificial intelligence laboratory environment assembled for testing the prototype. 10 refs.,more » 5 figs.« less

Control system for a small fission reactor

DOEpatents

Burelbach, James P.; Kann, William J.; Saiveau, James G.

1986-01-01

A system for controlling the reactivity of a small fission reactor includes an elongated, flexible hollow tube in the general form of a helical coiled spring axially positioned around and outside of the reactor vessel in an annular space between the reactor vessel and a surrounding cylindrical-shaped neutron reflector. A neutron absorbing material is provided within the hollow tube with the rate of the reaction controlled by the extension and compression of the hollow tube, e.g., extension of the tube increases reactivity while its compression reduces reactivity, in varying the amount of neutron absorbing material disposed between the reactor vessel and the neutron reflector. Conventional mechanical displacement means may be employed to control the coil density of the hollow tube as desired. In another embodiment, a plurality of flexible hollow tubes each containing a neutron absorber are positioned adjacent to one another in spaced relation around the periphery of the reactor vessel and inside the outer neutron reflector with reactivity controlled by the extension and compression of all or some of the coiled hollow tubes. Yet another embodiment of the invention envisions the neutron reflector in the form of an expandable coil spring positioned in an annular space between the reactor vessel and an outer neutron absorbing structure for controlling the neutron flux reflected back into the reactor vessel.

Reactor technology assessment and selection utilizing systems engineering approach

NASA Astrophysics Data System (ADS)

Zolkaffly, Muhammed Zulfakar; Han, Ki-In

2014-02-01

The first Nuclear power plant (NPP) deployment in a country is a complex process that needs to consider technical, economic and financial aspects along with other aspects like public acceptance. Increased interest in the deployment of new NPPs, both among newcomer countries and those with expanding programs, necessitates the selection of reactor technology among commercially available technologies. This paper reviews the Systems Decision Process (SDP) of Systems Engineering and applies it in selecting the most appropriate reactor technology for the deployment in Malaysia. The integrated qualitative and quantitative analyses employed in the SDP are explored to perform reactor technology assessment and to select the most feasible technology whose design has also to comply with the IAEA standard requirements and other relevant requirements that have been established in this study. A quick Malaysian case study result suggests that the country reside with PWR (pressurized water reactor) technologies with more detailed study to be performed in the future for the selection of the most appropriate reactor technology for Malaysia. The demonstrated technology assessment also proposes an alternative method to systematically and quantitatively select the most appropriate reactor technology.

Fuel Sustainability And Actinide Production Of Doping Minor Actinide In Water-Cooled Thorium Reactor

NASA Astrophysics Data System (ADS)

Permana, Sidik

2017-07-01

Fuel sustainability of nuclear energy is coming from an optimum fuel utilization of the reactor and fuel breeding program. Fuel cycle option becomes more important for fuel cycle utilization as well as fuel sustainability capability of the reactor. One of the important issues for recycle fuel option is nuclear proliferation resistance issue due to production plutonium. To reduce the proliferation resistance level, some barriers were used such as matrial barrier of nuclear fuel based on isotopic composition of even mass number of plutonium isotope. Analysis on nuclear fuel sustainability and actinide production composition based on water-cooled thorium reactor system has been done and all actinide composition are recycled into the reactor as a basic fuel cycle scheme. Some important parameters are evaluated such as doping composition of minor actinide (MA) and volume ratio of moderator to fuel (MFR). Some feasible parameters of breeding gains have been obtained by additional MA doping and some less moderation to fuel ratios (MFR). The system shows that plutonium and MA are obtained low compositions and it obtains some higher productions of even mass plutonium, which is mainly Pu-238 composition, as a control material to protect plutonium to be used as explosive devices.

SP-100 reactor with Brayton conversion for lunar surface applications

NASA Technical Reports Server (NTRS)

Mason, Lee S.; Rodriguez, Carlos D.; Mckissock, Barbara I.; Hanlon, James C.; Mansfield, Brian C.

1992-01-01

Examined here is the potential for integrating Brayton-cycle power conversion with the SP-100 reactor for lunar surface power system applications. Two designs were characterized and modeled. The first design integrates a 100-kWe SP-100 Brayton power system with a lunar lander. This system is intended to meet early lunar mission power needs while minimizing on-site installation requirements. Man-rated radiation protection is provided by an integral multilayer, cylindrical lithium hydride/tungsten (LiH/W) shield encircling the reactor vessel. Design emphasis is on ease of deployment, safety, and reliability, while utilizing relatively near-term technology. The second design combines Brayton conversion with the SP-100 reactor in a erectable 550-kWe powerplant concept intended to satisfy later-phase lunar base power requirements. This system capitalizes on experience gained from operating the initial 100-kWe module and incorporates some technology improvements. For this system, the reactor is emplaced in a lunar regolith excavation to provide man-rated shielding, and the Brayton engines and radiators are mounted on the lunar surface and extend radially from the central reactor. Design emphasis is on performance, safety, long life, and operational flexibility.

Titer-plate formatted continuous flow thermal reactors: Design and performance of a nanoliter reactor

PubMed Central

Chen, Pin-Chuan; Park, Daniel S.; You, Byoung-Hee; Kim, Namwon; Park, Taehyun; Soper, Steven A.; Nikitopoulos, Dimitris E.; Murphy, Michael C.

2010-01-01

Arrays of continuous flow thermal reactors were designed, configured, and fabricated in a 96-device (12 × 8) titer-plate format with overall dimensions of 120 mm × 96 mm, with each reactor confined to a 8 mm × 8 mm footprint. To demonstrate the potential, individual 20-cycle (740 nL) and 25-cycle (990 nL) reactors were used to perform the continuous flow polymerase chain reaction (CFPCR) for amplification of DNA fragments of different lengths. Since thermal isolation of the required temperature zones was essential for optimal biochemical reactions, three finite element models, executed with ANSYS (v. 11.0, Canonsburg, PA), were used to characterize the thermal performance and guide system design: (1) a single device to determine the dimensions of the thermal management structures; (2) a single CFPCR device within an 8 mm × 8 mm area to evaluate the integrity of the thermostatic zones; and (3) a single, straight microchannel representing a single loop of the spiral CFPCR device, accounting for all of the heat transfer modes, to determine whether the PCR cocktail was exposed to the proper temperature cycling. In prior work on larger footprint devices, simple grooves between temperature zones provided sufficient thermal resistance between zones. For the small footprint reactor array, 0.4 mm wide and 1.2 mm high fins were necessary within the groove to cool the PCR cocktail efficiently, with a temperature gradient of 15.8°C/mm, as it flowed from the denaturation zone to the renaturation zone. With temperature tolerance bands of ±2°C defined about the nominal temperatures, more than 72.5% of the microchannel length was located within the desired temperature bands. The residence time of the PCR cocktail in each temperature zone decreased and the transition times between zones increased at higher PCR cocktail flow velocities, leading to less time for the amplification reactions. Experiments demonstrated the performance of the CFPCR devices as a function of flow

Status report on the Small Secure Transportable Autonomous Reactor (SSTAR) /Lead-cooled Fast Reactor (LFR) and supporting research and development.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sienicki, J. J.; Moisseytsev, A.; Yang, W. S.

2008-06-23

This report provides an update on development of a pre-conceptual design for the Small Secure Transportable Autonomous Reactor (SSTAR) Lead-Cooled Fast Reactor (LFR) plant concept and supporting research and development activities. SSTAR is a small, 20 MWe (45 MWt), natural circulation, fast reactor plant for international deployment concept incorporating proliferation resistance for deployment in non-fuel cycle states and developing nations, fissile self-sufficiency for efficient utilization of uranium resources, autonomous load following making it suitable for small or immature grid applications, and a high degree of passive safety further supporting deployment in developing nations. In FY 2006, improvements have been mademore » at ANL to the pre-conceptual design of both the reactor system and the energy converter which incorporates a supercritical carbon dioxide Brayton cycle providing higher plant efficiency (44 %) and improved economic competitiveness. The supercritical CO2 Brayton cycle technology is also applicable to Sodium-Cooled Fast Reactors providing the same benefits. One key accomplishment has been the development of a control strategy for automatic control of the supercritical CO2 Brayton cycle in principle enabling autonomous load following over the full power range between nominal and essentially zero power. Under autonomous load following operation, the reactor core power adjusts itself to equal the heat removal from the reactor system to the power converter through the large reactivity feedback of the fast spectrum core without the need for motion of control rods, while the automatic control of the power converter matches the heat removal from the reactor to the grid load. The report includes early calculations for an international benchmarking problem for a LBE-cooled, nitride-fueled fast reactor core organized by the IAEA as part of a Coordinated Research Project on Small Reactors without Onsite Refueling; the calculations use the same

NON-CORROSIVE REACTOR FUEL SYSTEM

DOEpatents

Herrick, C.C.

1962-08-14

A non-corrosive nuclear reactor fuel system was developed utilizing a molten plutonium-- iron alloy fuel having about 2 at.% carbon and contained in a tantalum vessel. This carbon reacts with the interior surface of the tantalum vessel to form a plutonium resistant self-healing tantalum carbide film. (AEC)

Airbreathing combined cycle engine systems

NASA Technical Reports Server (NTRS)

Rohde, John

1992-01-01

The Air Force and NASA share a common interest in developing advanced propulsion systems for commercial and military aerospace vehicles which require efficient acceleration and cruise operation in the Mach 4 to 6 flight regime. The principle engine of interest is the turboramjet; however, other combined cycles such as the turboscramjet, air turborocket, supercharged ejector ramjet, ejector ramjet, and air liquefaction based propulsion are also of interest. Over the past months careful planning and program implementation have resulted in a number of development efforts that will lead to a broad technology base for those combined cycle propulsion systems. Individual development programs are underway in thermal management, controls materials, endothermic hydrocarbon fuels, air intake systems, nozzle exhaust systems, gas turbines and ramjet ramburners.

ORIGEN-based Nuclear Fuel Inventory Module for Fuel Cycle Assessment: Final Project Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Skutnik, Steven E.

and output processing, and depletion/decay solvers) can be self-contained into a single executable sequence. Further, to embed this capability into other software environments (such as the Cyclus fuel cycle simulator) requires that Origen’s capabilities be encapsulated into a portable, self-contained library which other codes can then call directly through function calls, thereby directly accessing the solver and data processing capabilities of Origen. Additional components relevant to this work include modernization of the reactor data libraries used by Origen for conducting nuclear fuel depletion calculations. This work has included the development of new fuel assembly lattices not previously available (such as for CANDU heavy-water reactor assemblies) as well as validation of updated lattices for light-water reactors updated to employ modern nuclear data evaluations. The CyBORG reactor analysis module as-developed under this workscope is fully capable of dynamic calculation of depleted fuel compositions from all commercial U.S. reactor assembly types as well as a number of international fuel types, including MOX, VVER, MAGNOX, and PHWR CANDU fuel assemblies. In addition, the Origen-based depletion engine allows for CyBORG to evaluate novel fuel assembly and reactor design types via creation of Origen reactor data libraries via SCALE. The establishment of this new modeling capability affords fuel cycle modelers a substantially improved ability to model dynamically-changing fuel cycle and reactor conditions, including recycled fuel compositions from fuel cycle scenarios involving material recycle into thermal-spectrum systems.« less

Expert systems for fault diagnosis in nuclear reactor control

NASA Astrophysics Data System (ADS)

Jalel, N. A.; Nicholson, H.

1990-11-01

An expert system for accident analysis and fault diagnosis for the Loss Of Fluid Test (LOFT) reactor, a small scale pressurized water reactor, was developed for a personal computer. The knowledge of the system is presented using a production rule approach with a backward chaining inference engine. The data base of the system includes simulated dependent state variables of the LOFT reactor model. Another system is designed to assist the operator in choosing the appropriate cooling mode and to diagnose the fault in the selected cooling system. The response tree, which is used to provide the link between a list of very specific accident sequences and a set of generic emergency procedures which help the operator in monitoring system status, and to differentiate between different accident sequences and select the correct procedures, is used to build the system knowledge base. Both systems are written in TURBO PROLOG language and can be run on an IBM PC compatible with 640k RAM, 40 Mbyte hard disk and color graphics.

Attrition reactor system

DOEpatents

Scott, Charles D.; Davison, Brian H.

1993-01-01

A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur.

Safety and Regulatory Issues of the Thorium Fuel Cycle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ade, Brian; Worrall, Andrew; Powers, Jeffrey

2014-02-01

Thorium has been widely considered an alternative to uranium fuel because of its relatively large natural abundance and its ability to breed fissile fuel (233U) from natural thorium (232Th). Possible scenarios for using thorium in the nuclear fuel cycle include use in different nuclear reactor types (light water, high temperature gas cooled, fast spectrum sodium, molten salt, etc.), advanced accelerator-driven systems, or even fission-fusion hybrid systems. The most likely near-term application of thorium in the United States is in currently operating light water reactors (LWRs). This use is primarily based on concepts that mix thorium with uranium (UO2 + ThO2),more » add fertile thorium (ThO2) fuel pins to LWR fuel assemblies, or use mixed plutonium and thorium (PuO2 + ThO2) fuel assemblies. The addition of thorium to currently operating LWRs would result in a number of different phenomenological impacts on the nuclear fuel. Thorium and its irradiation products have nuclear characteristics that are different from those of uranium. In addition, ThO2, alone or mixed with UO2 fuel, leads to different chemical and physical properties of the fuel. These aspects are key to reactor safety-related issues. The primary objectives of this report are to summarize historical, current, and proposed uses of thorium in nuclear reactors; provide some important properties of thorium fuel; perform qualitative and quantitative evaluations of both in-reactor and out-of-reactor safety issues and requirements specific to a thorium-based fuel cycle for current LWR reactor designs; and identify key knowledge gaps and technical issues that need to be addressed for the licensing of thorium LWR fuel in the United States.« less

Passive decay heat removal system for water-cooled nuclear reactors

DOEpatents

Forsberg, Charles W.

1991-01-01

A passive decay-heat removal system for a water-cooled nuclear reactor employs a closed heat transfer loop having heat-exchanging coils inside an open-topped, insulated box located inside the reactor vessel, below its normal water level, in communication with a condenser located outside of containment and exposed to the atmosphere. The heat transfer loop is located such that the evaporator is in a position where, when the water level drops in the reactor, it will become exposed to steam. Vapor produced in the evaporator passes upward to the condenser above the normal water level. In operation, condensation in the condenser removes heat from the system, and the condensed liquid is returned to the evaporator. The system is disposed such that during normal reactor operations where the water level is at its usual position, very little heat will be removed from the system, but during emergency, low water level conditions, substantial amounts of decay heat will be removed.

Life cycle assessment of hydrogen production from S-I thermochemical process coupled to a high temperature gas reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Giraldi, M. R.; Francois, J. L.; Castro-Uriegas, D.

The purpose of this paper is to quantify the greenhouse gas (GHG) emissions associated to the hydrogen produced by the sulfur-iodine thermochemical process, coupled to a high temperature nuclear reactor, and to compare the results with other life cycle analysis (LCA) studies on hydrogen production technologies, both conventional and emerging. The LCA tool was used to quantify the impacts associated with climate change. The product system was defined by the following steps: (i) extraction and manufacturing of raw materials (upstream flows), (U) external energy supplied to the system, (iii) nuclear power plant, and (iv) hydrogen production plant. Particular attention wasmore » focused to those processes where there was limited information from literature about inventory data, as the TRISO fuel manufacture, and the production of iodine. The results show that the electric power, supplied to the hydrogen plant, is a sensitive parameter for GHG emissions. When the nuclear power plant supplied the electrical power, low GHG emissions were obtained. These results improve those reported by conventional hydrogen production methods, such as steam reforming. (authors)« less

Thorium Fuel Cycle Option Screening in the United States

DOE Office of Scientific and Technical Information (OSTI.GOV)

Taiwo, Temitope A.; Kim, Taek K.; Wigeland, Roald A.

2016-05-01

As part of a nuclear fuel cycle Evaluation and Screening (E&S) study, a wide-range of thorium fuel cycle options were evaluated and their performance characteristics and challenges to implementation were compared to those of other nuclear fuel cycle options based on criteria specified by the Nuclear Energy Office of the U.S. Department of Energy (DOE). The evaluated nuclear fuel cycles included the once-through, limited, and continuous recycle options using critical or externally-driven nuclear energy systems. The E&S study found that the continuous recycle of 233U/Th in fuel cycles using either thermal or fast reactors is an attractive promising fuel cyclemore » option with high effective fuel resource utilization and low waste generation, but did not perform quite as well as the continuous recycle of Pu/U using a fast critical system, which was identified as one of the most promising fuel cycle options in the E&S study. This is because compared to their uranium counterparts the thorium-based systems tended to have higher radioactivity in the short term (about 100 years post irradiation) because of differences in the fission product yield curves, and in the long term (100,000 years post irradiation) because of the decay of 233U and daughters, and because of higher mass flow rates due to lower discharge burnups. Some of the thorium-based systems also require enriched uranium support, which tends to be detrimental to resource utilization and waste generation metrics. Finally, similar to the need for developing recycle fuel fabrication, fuels separations and fast reactors for the most promising options using Pu/U recycle, the future thorium-based fuel cycle options with continuous recycle would also require such capabilities, although their deployment challenges are expected to be higher since such facilities have not been developed in the past to a comparable level of maturity for Th-based systems.« less

NEUTRONIC REACTOR FUEL ELEMENT AND CORE SYSTEM

DOEpatents

Moore, W.T.

1958-09-01

This patent relates to neutronic reactors and in particular to an improved fuel element and a novel reactor core system for facilitating removal of contaminating fission products, as they are fermed, from association with the flssionable fuel, so as to mitigate the interferent effects of such fission products during reactor operation. The fuel elements are comprised of tubular members impervious to fluid and contatning on their interior surfaces a thin layer of fissionable material providing a central void. The core structure is comprised of a plurality of the tubular fuel elements arranged in parallel and a closed manifold connected to their ends. In the reactor the core structure is dispersed in a water moderator and coolant within a pressure vessel, and a means connected to said manifuld is provided for withdrawing and disposing of mobile fission product contamination from the interior of the feel tubes and manifold.

Systems aspects of a space nuclear reactor power system

NASA Technical Reports Server (NTRS)

Jaffe, L.; Fujita, T.; Beatty, R.; Bhandari, P.; Chow, E.; Deininger, W.; Ewell, R.; Grossman, M.; Bloomfield, H.; Heller, J.

1988-01-01

Various system aspects of a 300-kW nuclear reactor power system for spacecraft have been investigated. Special attention is given to the cases of a reusable OTV and a space-based radar. It is demonstrated that the stowed length of the power system is important to mission design, and that orbital storage for months to years may be needed for missions involving orbital assembly.

High Efficiency Nuclear Power Plants using Liquid Fluoride Thorium Reactor Technology

NASA Technical Reports Server (NTRS)

Juhasz, Albert J.; Rarick, Richard A.; Rangarajan, Rajmohan

2009-01-01

An overall system analysis approach is used to propose potential conceptual designs of advanced terrestrial nuclear power plants based on Oak Ridge National Laboratory (ORNL) Molten Salt Reactor (MSR) experience and utilizing Closed Cycle Gas Turbine (CCGT) thermal-to-electric energy conversion technology. In particular conceptual designs for an advanced 1 GWe power plant with turbine reheat and compressor intercooling at a 950 K turbine inlet temperature (TIT), as well as near term 100 MWe demonstration plants with TITS of 950 K and 1200 K are presented. Power plant performance data were obtained for TITS ranging from 650 to 1300 K by use of a Closed Brayton Cycle (CBC) systems code which considered the interaction between major sub-systems, including the Liquid Fluoride Thorium Reactor (LFTR), heat source and heat sink heat exchangers, turbo -generator machinery, and an electric power generation and transmission system. Optional off-shore submarine installation of the power plant is a major consideration.

High Efficiency Nuclear Power Plants Using Liquid Fluoride Thorium Reactor Technology

NASA Technical Reports Server (NTRS)

Juhasz, Albert J.; Rarick, Richard A.; Rangarajan, Rajmohan

2009-01-01

An overall system analysis approach is used to propose potential conceptual designs of advanced terrestrial nuclear power plants based on Oak Ridge National Laboratory (ORNL) Molten Salt Reactor (MSR) experience and utilizing Closed Cycle Gas Turbine (CCGT) thermal-to-electric energy conversion technology. In particular conceptual designs for an advanced 1 GWe power plant with turbine reheat and compressor intercooling at a 950 K turbine inlet temperature (TIT), as well as near term 100 MWe demonstration plants with TITs of 950 and 1200 K are presented. Power plant performance data were obtained for TITs ranging from 650 to 1300 K by use of a Closed Brayton Cycle (CBC) systems code which considered the interaction between major sub-systems, including the Liquid Fluoride Thorium Reactor (LFTR), heat source and heat sink heat exchangers, turbo-generator machinery, and an electric power generation and transmission system. Optional off-shore submarine installation of the power plant is a major consideration.

Attrition reactor system

DOEpatents

Scott, C.D.; Davison, B.H.

1993-09-28

A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur. 2 figures.

Decay Heat Removal in GEN IV Gas-Cooled Fast Reactors

DOE PAGES

Cheng, Lap-Yan; Wei, Thomas Y. C.

2009-01-01

The safety goal of the current designs of advanced high-temperature thermal gas-cooled reactors (HTRs) is that no core meltdown would occur in a depressurization event with a combination of concurrent safety system failures. This study focused on the analysis of passive decay heat removal (DHR) in a GEN IV direct-cycle gas-cooled fast reactor (GFR) which is based on the technology developments of the HTRs. Given the different criteria and design characteristics of the GFR, an approach different from that taken for the HTRs for passive DHR would have to be explored. Different design options based on maintaining core flow weremore » evaluated by performing transient analysis of a depressurization accident using the system code RELAP5-3D. The study also reviewed the conceptual design of autonomous systems for shutdown decay heat removal and recommends that future work in this area should be focused on the potential for Brayton cycle DHRs.« less

DynMo: Dynamic Simulation Model for Space Reactor Power Systems

NASA Astrophysics Data System (ADS)

El-Genk, Mohamed; Tournier, Jean-Michel

2005-02-01

A Dynamic simulation Model (DynMo) for space reactor power systems is developed using the SIMULINK® platform. DynMo is modular and could be applied to power systems with different types of reactors, energy conversion, and heat pipe radiators. This paper presents a general description of DynMo-TE for a space power system powered by a Sectored Compact Reactor (SCoRe) and that employs off-the-shelf SiGe thermoelectric converters. SCoRe is liquid metal cooled and designed for avoidance of a single point failure. The reactor core is divided into six equal sectors that are neutronically, but not thermal-hydraulically, coupled. To avoid a single point failure in the power system, each reactor sector has its own primary and secondary loops, and each loop is equipped with an electromagnetic (EM) pump. A Power Conversion assembly (PCA) and a Thermoelectric Conversion Assembly (TCA) of the primary and secondary EM pumps thermally couple each pair of a primary and a secondary loop. The secondary loop transports the heat rejected by the PCA and the pumps TCA to a rubidium heat pipes radiator panel. The primary loops transport the thermal power from the reactor sector to the PCAs for supplying a total of 145-152 kWe to the load at 441-452 VDC, depending on the selections of the primary and secondary liquid metal coolants. The primary and secondary coolant combinations investigated are lithium (Li)/Li, Li/sodium (Na), Na-Na, Li/NaK-78 and Na/NaK-78, for which the reactor exit temperature is kept below 1250 K. The results of a startup transient of the system from an initial temperature of 500 K are compared and discussed.

Supply of enriched uranium for research reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mueller, H.

1997-08-01

Since the RERTR-meeting In Newport/USA in 1990 the author delivered a series of papers in connection with the fuel cycle for research reactors dealing with its front-end. In these papers the author underlined the need for unified specifications for enriched uranium metal suitable for the production of fuel elements and made proposals with regard to the re-use of in Europe reprocessed highly enriched uranium. With regard to the fuel cycle of research reactors the research reactor community was since 1989 more concentrating on the problems of its back-end since the USA stopped the acceptance of spent research reactor fuel onmore » December 31, 1988. Now, since it is apparent that these back-end problem have been solved by AEA`s ability to reprocess and the preparedness of the USA to again accept physically spent research reactor fuel the author is focusing with this paper again on the front-end of the fuel cycle on the question whether there is at all a safe supply of low and high enriched uranium for research reactors in the future.« less

System Study: Reactor Core Isolation Cooling 1998-2014

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schroeder, John Alton

2015-12-01

This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant trends were identified in the RCIC results.

Reactor protection system with automatic self-testing and diagnostic

DOEpatents

Gaubatz, Donald C.

1996-01-01

A reactor protection system having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Automatic detection and discrimination against failed sensors allows the reactor protection system to automatically enter a known state when sensor failures occur. Cross communication of sensor readings allows comparison of four theoretically "identical" values. This permits identification of sensor errors such as drift or malfunction. A diagnostic request for service is issued for errant sensor data. Automated self test and diagnostic monitoring, sensor input through output relay logic, virtually eliminate the need for manual surveillance testing. This provides an ability for each division to cross-check all divisions and to sense failures of the hardware logic.

Reactor protection system with automatic self-testing and diagnostic

DOEpatents

Gaubatz, D.C.

1996-12-17

A reactor protection system is disclosed having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Automatic detection and discrimination against failed sensors allows the reactor protection system to automatically enter a known state when sensor failures occur. Cross communication of sensor readings allows comparison of four theoretically ``identical`` values. This permits identification of sensor errors such as drift or malfunction. A diagnostic request for service is issued for errant sensor data. Automated self test and diagnostic monitoring, sensor input through output relay logic, virtually eliminate the need for manual surveillance testing. This provides an ability for each division to cross-check all divisions and to sense failures of the hardware logic. 16 figs.

Study of Pu consumption in advanced light water reactors: Evaluation of GE advanced boiling water reactor plants - compilation of Phase 1B task reports

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

1993-09-15

This report contains an extensive evaluation of GE advanced boiling water reactor plants prepared for United State Department of Energy. The general areas covered in this report are: core and system performance; fuel cycle; infrastructure and deployment; and safety and environmental approval.

Nuclear reactor shutdown system

DOEpatents

Bhate, Suresh K.; Cooper, Martin H.; Riffe, Delmar R.; Kinney, Calvin L.

1981-01-01

An inherent shutdown system for a nuclear reactor having neutron absorbing rods affixed to an armature which is held in an upper position by a magnetic flux flowing through a Curie temperature material. The Curie temperature material is fixedly positioned about the exterior of an inner duct in an annular region through which reactor coolant flows. Elongated fuel rods extending from within the core upwardly toward the Curie temperature material are preferably disposed within the annular region. Upon abnormal conditions which result in high neutron flux and coolant temperature, the Curie material loses its magnetic permeability, breaking the magnetic flux path and allowing the armature and absorber rods to drop into the core, thus shutting down the fissioning reaction. The armature and absorber rods are retrieved by lowering the housing for the electromagnet forming coils which create a magnetic flux path which includes the inner duct wall. The coil housing then is raised, resetting the armature.

Design of virtual SCADA simulation system for pressurized water reactor

NASA Astrophysics Data System (ADS)

Wijaksono, Umar; Abdullah, Ade Gafar; Hakim, Dadang Lukman

2016-02-01

The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.

Solar Metal Sulfate-Ammonia Based Thermochemical Water Splitting Cycle for Hydrogen Production

NASA Technical Reports Server (NTRS)

T-Raissi, Ali (Inventor); Muradov, Nazim (Inventor); Huang, Cunping (Inventor)

2014-01-01

Two classes of hybrid/thermochemical water splitting processes for the production of hydrogen and oxygen have been proposed based on (1) metal sulfate-ammonia cycles (2) metal pyrosulfate-ammonia cycles. Methods and systems for a metal sulfate MSO.sub.4--NH3 cycle for producing H2 and O2 from a closed system including feeding an aqueous (NH3)(4)SO3 solution into a photoctalytic reactor to oxidize the aqueous (NH3)(4)SO3 into aqueous (NH3)(2)SO4 and reduce water to hydrogen, mixing the resulting aqueous (NH3)(2)SO4 with metal oxide (e.g. ZnO) to form a slurry, heating the slurry of aqueous (NH4)(2)SO4 and ZnO(s) in the low temperature reactor to produce a gaseous mixture of NH3 and H2O and solid ZnSO4(s), heating solid ZnSO4 at a high temperature reactor to produce a gaseous mixture of SO2 and O2 and solid product ZnO, mixing the gaseous mixture of SO2 and O2 with an NH3 and H2O stream in an absorber to form aqueous (NH4)(2)SO3 solution and separate O2 for aqueous solution, recycling the resultant solution back to the photoreactor and sending ZnO to mix with aqueous (NH4)(2)SO4 solution to close the water splitting cycle wherein gaseous H2 and O2 are the only products output from the closed ZnSO4--NH3 cycle.

Life cycle assessment of nuclear-based hydrogen production via thermochemical water splitting using a copper-chlorine (Cu-Cl) cycle

NASA Astrophysics Data System (ADS)

Ozbilen, Ahmet Ziyaettin

The energy carrier hydrogen is expected to solve some energy challenges. Since its oxidation does not emit greenhouse gases (GHGs), its use does not contribute to climate change, provided that it is derived from clean energy sources. Thermochemical water splitting using a Cu-Cl cycle, linked with a nuclear super-critical water cooled reactor (SCWR), which is being considered as a Generation IV nuclear reactor, is a promising option for hydrogen production. In this thesis, a comparative environmental study is reported of the three-, four- and five-step Cu-Cl thermochemical water splitting cycles with various other hydrogen production methods. The investigation uses life cycle assessment (LCA), which is an analytical tool to identify and quantify environmentally critical phases during the life cycle of a system or a product and/or to evaluate and decrease the overall environmental impact of the system or product. The LCA results for the hydrogen production processes indicate that the four-step Cu-Cl cycle has lower environmental impacts than the three- and five-step Cu-Cl cycles due to its lower thermal energy requirement. Parametric studies show that acidification potentials (APs) and global warming potentials (GWPs) for the four-step Cu-Cl cycle can be reduced from 0.0031 to 0.0028 kg SO2-eq and from 0.63 to 0.55 kg CO2-eq, respectively, if the lifetime of the system increases from 10 to 100 years. Moreover, the comparative study shows that the nuclear-based S-I and the four-step Cu-Cl cycles are the most environmentally benign hydrogen production methods in terms of AP and GWP. GWPs of the S-I and the four-step Cu-Cl cycles are 0.412 and 0.559 kg CO2-eq for reference case which has a lifetime of 60 years. Also, the corresponding APs of these cycles are 0.00241 and 0.00284 kg SO2-eq. It is also found that an increase in hydrogen plant efficiency from 0.36 to 0.65 decreases the GWP from 0.902 to 0.412 kg CO 2-eq and the AP from 0.00459 to 0.00209 kg SO2-eq for the

System and method for temperature control in an oxygen transport membrane based reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kelly, Sean M.

A system and method for temperature control in an oxygen transport membrane based reactor is provided. The system and method involves introducing a specific quantity of cooling air or trim air in between stages in a multistage oxygen transport membrane based reactor or furnace to maintain generally consistent surface temperatures of the oxygen transport membrane elements and associated reactors. The associated reactors may include reforming reactors, boilers or process gas heaters.

Development of the ANL plant dynamics code and control strategies for the supercritical carbon dioxide Brayton cycle and code validation with data from the Sandia small-scale supercritical carbon dioxide Brayton cycle test loop.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Moisseytsev, A.; Sienicki, J. J.

2011-11-07

Significant progress has been made in the ongoing development of the Argonne National Laboratory (ANL) Plant Dynamics Code (PDC), the ongoing investigation and development of control strategies, and the analysis of system transient behavior for supercritical carbon dioxide (S-CO{sub 2}) Brayton cycles. Several code modifications have been introduced during FY2011 to extend the range of applicability of the PDC and to improve its calculational stability and speed. A new and innovative approach was developed to couple the Plant Dynamics Code for S-CO{sub 2} cycle calculations with SAS4A/SASSYS-1 Liquid Metal Reactor Code System calculations for the transient system level behavior onmore » the reactor side of a Sodium-Cooled Fast Reactor (SFR) or Lead-Cooled Fast Reactor (LFR). The new code system allows use of the full capabilities of both codes such that whole-plant transients can now be simulated without additional user interaction. Several other code modifications, including the introduction of compressor surge control, a new approach for determining the solution time step for efficient computational speed, an updated treatment of S-CO{sub 2} cycle flow mergers and splits, a modified enthalpy equation to improve the treatment of negative flow, and a revised solution of the reactor heat exchanger (RHX) equations coupling the S-CO{sub 2} cycle to the reactor, were introduced to the PDC in FY2011. All of these modifications have improved the code computational stability and computational speed, while not significantly affecting the results of transient calculations. The improved PDC was used to continue the investigation of S-CO{sub 2} cycle control and transient behavior. The coupled PDC-SAS4A/SASSYS-1 code capability was used to study the dynamic characteristics of a S-CO{sub 2} cycle coupled to a SFR plant. Cycle control was investigated in terms of the ability of the cycle to respond to a linear reduction in the electrical grid demand from 100% to 0% at a rate

The Conceptual Design of an Integrated Nuclearhydrogen Production Plant Using the Sulfur Cycle Water Decomposition System

NASA Technical Reports Server (NTRS)

Farbman, G. H.

1976-01-01

A hydrogen production plant was designed based on a hybrid electrolytic-thermochemical process for decomposing water. The sulfur cycle water decomposition system is driven by a very high temperature nuclear reactor that provides 1,283 K helium working gas. The plant is sized to approximately ten million standard cubic meters per day of electrolytically pure hydrogen and has an overall thermal efficiently of 45.2 percent. The economics of the plant were evaluated using ground rules which include a 1974 cost basis without escalation, financing structure and other economic factors. Taking into account capital, operation, maintenance and nuclear fuel cycle costs, the cost of product hydrogen was calculated at $5.96/std cu m for utility financing. These values are significantly lower than hydrogen costs from conventional water electrolysis plants and competitive with hydrogen from coal gasification plants.

Isotopic signature of atmospheric xenon released from light water reactors.

PubMed

Kalinowski, Martin B; Pistner, Christoph

2006-01-01

A global monitoring system for atmospheric xenon radioactivity is being established as part of the International Monitoring System to verify compliance with the Comprehensive Nuclear-Test-Ban Treaty (CTBT). The isotopic activity ratios of (135)Xe, (133m)Xe, (133)Xe and (131m)Xe are of interest for distinguishing nuclear explosion sources from civilian releases. Simulations of light water reactor (LWR) fuel burn-up through three operational reactor power cycles are conducted to explore the possible xenon isotopic signature of nuclear reactor releases under different operational conditions. It is studied how ratio changes are related to various parameters including the neutron flux, uranium enrichment and fuel burn-up. Further, the impact of diffusion and mixing on the isotopic activity ratio variability are explored. The simulations are validated with reported reactor emissions. In addition, activity ratios are calculated for xenon isotopes released from nuclear explosions and these are compared to the reactor ratios in order to determine whether the discrimination of explosion releases from reactor effluents is possible based on isotopic activity ratios.

Advanced Reactor Passive System Reliability Demonstration Analysis for an External Event

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bucknor, Matthew D.; Grabaskas, David; Brunett, Acacia J.

2016-01-01

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

Advanced Reactor Passive System Reliability Demonstration Analysis for an External Event

DOE PAGES

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

2017-01-24

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

Reactor safeguards system assessment and design. Volume I

DOE Office of Scientific and Technical Information (OSTI.GOV)

Varnado, G.B.; Ericson, D.M. Jr.; Daniel, S.L.

1978-06-01

This report describes the development and application of a methodology for evaluating the effectiveness of nuclear power reactor safeguards systems. Analytic techniques are used to identify the sabotage acts which could lead to release of radioactive material from a nuclear power plant, to determine the areas of a plant which must be protected to assure that significant release does not occur, to model the physical plant layout, and to evaluate the effectiveness of various safeguards systems. The methodology was used to identify those aspects of reactor safeguards systems which have the greatest effect on overall system performance and which, therefore,more » should be emphasized in the licensing process. With further refinements, the methodology can be used by the licensing reviewer to aid in assessing proposed or existing safeguards systems.« less

Summary of space nuclear reactor power systems, 1983 - 1992

NASA Astrophysics Data System (ADS)

Buden, D.

1993-08-01

This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressed from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987-88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power.

Design of virtual SCADA simulation system for pressurized water reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wijaksono, Umar, E-mail: umar.wijaksono@student.upi.edu; Abdullah, Ade Gafar; Hakim, Dadang Lukman

The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles ofmore » energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.« less

Cermet-fueled reactors for advanced space applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cowan, C.L.; Palmer, R.S.; Taylor, I.N.

Cermet-fueled nuclear reactors are attractive candidates for high-performance advanced space power systems. The cermet consists of a hexagonal matrix of a refractory metal and a ceramic fuel, with multiple tubular flow channels. The high performance characteristics of the fuel matrix come from its high strength at elevated temperatures and its high thermal conductivity. The cermet fuel concept evolved in the 1960s with the objective of developing a reactor design that could be used for a wide range of mobile power generating sytems, including both Brayton and Rankine power conversion cycles. High temperature thermal cycling tests for the cermet fuel weremore » carried out by General Electric as part of the 710 Project (General Electric 1966), and by Argonne National Laboratory in the Direct Nuclear Rocket Program (1965). Development programs for cermet fuel are currently under way at Argonne National Laboratory and Pacific Northwest Laboratory. The high temperature qualification tests from the 1960s have provided a base for the incorporation of cermet fuel in advanced space applications. The status of the cermet fuel development activities and descriptions of the key features of the cermet-fueled reactor design are summarized in this paper.« less

LIQUID METAL REACTOR COOLING SYSTEMS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Aberdam, M.; Gros, G.

1965-02-01

This report is part of a series of bibliographies. The specific purpose of this report is to describe the various elements of the cooling systems in the principal liquid-metal-cooled reactors now operating, being contsructed, or in the design stage. The information given is drawn from reports or publicatios received during or before September 1964.

Novel online monitoring and alert system for anaerobic digestion reactors.

PubMed

Dong, Fang; Zhao, Quan-Bao; Li, Wen-Wei; Sheng, Guo-Ping; Zhao, Jin-Bao; Tang, Yong; Yu, Han-Qing; Kubota, Kengo; Li, Yu-You; Harada, Hideki

2011-10-15

Effective monitoring and diagnosis of anaerobic digestion processes is a great challenge for anaerobic digestion reactors, which limits their stable operation. In this work, an online monitoring and alert system for upflow anaerobic sludge blanket (UASB) reactors is developed on the basis of a set of novel evaluating indexes. The two indexes, i.e., stability index S and auxiliary index a, which incorporate both gas- and liquid-phase parameters for UASB, enable a quantitative and comprehensive evaluation of reactor status. A series of shock tests is conducted to evaluate the response of the monitoring and alert system to organic overloading, hydraulic, temperature, and toxicant shocks. The results show that this system enables an accurate and rapid monitoring and diagnosis of the reactor status, and offers reliable early warnings on the potential risks. As the core of this system, the evaluating indexes are demonstrated to be of high accuracy and sensitivity in process evaluation and good adaptability to the artificial intelligence and automated control apparatus. This online monitoring and alert system presents a valuable effort to promote the automated monitoring and control of anaerobic digestion process, and holds a high promise for application.

Developments and Tendencies in Fission Reactor Concepts

NASA Astrophysics Data System (ADS)

Adamov, E. O.; Fuji-Ie, Y.

This chapter describes, in two parts, new-generation nuclear energy systems that are required to be in harmony with nature and to make full use of nuclear resources. The issues of transmutation and containment of radioactive waste will also be addressed. After a short introduction to the first part, Sect. 58.1.2 will detail the requirements these systems must satisfy on the basic premise of peaceful use of nuclear energy. The expected designs themselves are described in Sect. 58.1.3. The subsequent sections discuss various types of advanced reactor systems. Section 58.1.4 deals with the light water reactor (LWR) whose performance is still expected to improve, which would extend its application in the future. The supercritical-water-cooled reactor (SCWR) will also be shortly discussed. Section 58.1.5 is mainly on the high temperature gas-cooled reactor (HTGR), which offers efficient and multipurpose use of nuclear energy. The gas-cooled fast reactor (GFR) is also included. Section 58.1.6 focuses on the sodium-cooled fast reactor (SFR) as a promising concept for advanced nuclear reactors, which may help both to achieve expansion of energy sources and environmental protection thus contributing to the sustainable development of mankind. The molten-salt reactor (MSR) is shortly described in Sect. 58.1.7. The second part of the chapter deals with reactor systems of a new generation, which are now found at the research and development (R&D) stage and in the medium term of 20-30 years can shape up as reliable, economically efficient, and environmentally friendly energy sources. They are viewed as technologies of cardinal importance, capable of resolving the problems of fuel resources, minimizing the quantities of generated radioactive waste and the environmental impacts, and strengthening the regime of nonproliferation of the materials suitable for nuclear weapons production. Particular attention has been given to naturally safe fast reactors with a closed fuel cycle (CFC

Technical, hygiene, economic, and life cycle assessment of full-scale moving bed biofilm reactors for wastewater treatment in India.

PubMed

Singh, Anju; Kamble, Sheetal Jaisingh; Sawant, Megha; Chakravarthy, Yogita; Kazmi, Absar; Aymerich, Enrique; Starkl, Markus; Ghangrekar, Makarand; Philip, Ligy

2018-01-01

Moving bed biofilm reactor (MBBR) is a highly effective biological treatment process applied to treat both urban and industrial wastewaters in developing countries. The present study investigated the technical performance of ten full-scale MBBR systems located across India. The biochemical oxygen demand, chemical oxygen demand, total suspended solid, pathogens, and nutrient removal efficiencies were low as compared to the values claimed in literature. Plant 1 was considered for evaluation of environmental impacts using life cycle assessment approach. CML 2 baseline 2000 methodology was adopted, in which 11 impact categories were considered. The life cycle impact assessment results revealed that the main environmental hot spot of this system was energy consumption. Additionally, two scenarios were compared: scenario 1 (direct discharge of treated effluent, i.e., no reuse) and scenario 2 (effluent reuse and tap water replacement). The results showed that scenario 2 significantly reduce the environmental impact in all the categories ultimately decreasing the environmental burden. Moreover, significant economic and environmental benefits can be obtained in scenario 2 by replacing the freshwater demand for non-potable uses. To enhance the performance of wastewater treatment plant (WWTP), there is a need to optimize energy consumption and increase wastewater collection efficiency to maximize the operating capacity of plant and minimize overall environmental footprint. It was concluded that MBBR can be a good alternative for upgrading and optimizing existing municipal wastewater treatment plants with appropriate tertiary treatment. Graphical abstract ᅟ.

Computer optimization of reactor-thermoelectric space power systems

NASA Technical Reports Server (NTRS)

Maag, W. L.; Finnegan, P. M.; Fishbach, L. H.

1973-01-01

A computer simulation and optimization code that has been developed for nuclear space power systems is described. The results of using this code to analyze two reactor-thermoelectric systems are presented.

Data acquisition system for segmented reactor antineutrino detector

NASA Astrophysics Data System (ADS)

Hons, Z.; Vlášek, J.

2017-01-01

This paper describes the data acquisition system used for data readout from the PMT channels of a segmented detector of reactor antineutrinos with active shielding. Theoretical approach to the data acquisition is described and two possible solutions using QDCs and digitizers are discussed. Also described are the results of the DAQ performance during routine data taking operation of DANSS. DANSS (Detector of the reactor AntiNeutrino based on Solid Scintillator) is a project aiming to measure a spectrum of reactor antineutrinos using inverse beta decay (IBD) in a plastic scintillator. The detector is located close to an industrial nuclear reactor core and is covered by passive and active shielding. It is expected to have about 15000 IBD interactions per day. Light from the detector is sensed by PMT and SiPM.

Modeling Chilled-Water Storage System Components for Coupling to a Small Modular Reactor in a Nuclear Hybrid Energy System

NASA Astrophysics Data System (ADS)

Misenheimer, Corey Thomas

The intermittency of wind and solar power puts strain on electric grids, often forcing carbonbased and nuclear sources of energy to operate in a load-follow mode. Operating nuclear reactors in a load-follow fashion is undesirable due to the associated thermal and mechanical stresses placed on the fuel and other reactor components. Various Thermal Energy Storage (TES) elements and ancillary energy applications can be coupled to nuclear (or renewable) power sources to help absorb grid instabilities caused by daily electric demand changes and renewable intermittency, thereby forming the basis of a candidate Nuclear Hybrid Energy System (NHES). During the warmer months of the year in many parts of the country, facility air-conditioning loads are significant contributors to the increase in the daily peak electric demand. Previous research demonstrated that a stratified chilled-water storage tank can displace peak cooling loads to off-peak hours. Based on these findings, the objective of this work is to evaluate the prospect of using a stratified chilled-water storage tank as a potential TES reservoir for a nuclear reactor in a NHES. This is accomplished by developing time-dependent models of chilled-water system components, including absorption chillers, cooling towers, a storage tank, and facility cooling loads appropriate for a large office space or college campus, as a callable FORTRAN subroutine. The resulting TES model is coupled to a high-fidelity mPower-sized Small Modular Reactor (SMR) Simulator, with the goal of utilizing excess reactor capacity to operate several sizable chillers in order to keep reactor power constant. Chilled-water production via single effect, lithium bromide (LiBr) absorption chillers is primarily examined in this study, although the use of electric chillers is briefly explored. Absorption chillers use hot water or low-pressure steam to drive an absorption-refrigeration cycle. The mathematical framework for a high-fidelity dynamic

UF6 breeder reactor power plants for electric power generation

NASA Technical Reports Server (NTRS)

Rust, J. H.; Clement, J. D.; Hohl, F.

1976-01-01

The reactor concept analyzed is a U-233F6 core surrounded by a molten salt (Li(7)F, BeF2, ThF4) blanket. Nuclear survey calculations were carried out for both spherical and cylindrical geometries. Thermodynamic cycle calculations were performed for a variety of Rankine cycles. A conceptual design is presented along with a system layout for a 1000 MW stationary power plant. Advantages of the gas core breeder reactor (GCBR) are as follows: (1) high efficiency; (2) simplified on-line reprocessing; (3) inherent safety considerations; (4) high breeding ratio; (5) possibility of burning all or most of the long-lived nuclear waste actinides; and (6) possibility of extrapolating the technology to higher temperatures and MHD direct conversion.

Estimates of power requirements for a Manned Mars Rover powered by a nuclear reactor

NASA Technical Reports Server (NTRS)

Morley, Nicholas J.; El-Genk, Mohamed S.; Cataldo, Robert; Bloomfield, Harvey

1991-01-01

This paper assesses the power requirement for a Manned Mars Rover vehicle. Auxiliary power needs are fulfilled using a hybrid solar photovoltaic/regenerative fuel cell system, while the primary power needs are meet using an SP-100 type reactor. The primary electric power needs, which include 30-kW(e) net user power, depend on the reactor thermal power and the efficiency of the power conversion system. Results show that an SP-100 type reactor coupled to a Free Piston Stirling Engine yields the lowest total vehicle mass and lowest specific mass for the power system. The second lowest mass was for a SP-100 reactor coupled to a Closed Brayton Cycle using He/Xe as the working fluid. The specific mass of the nuclear reactor power system, including a man-rated radiation shield, ranged from 150-kg/kW(e) to 190-kg/KW(e) and the total mass of the Rover vehicle varied depend upon the cruising speed.

An assessment of the attractiveness of material associated with thorium/uranium and uranium closed fuel cycles from a safeguards perspective

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bathke, Charles Gary; Wallace, Richard K; Hase, Kevin R

2010-01-01

This paper reports the continued evaluation of the attractiveness of materials mixtures containing special nuclear materials (SNM) associated with various proposed nuclear fuel cycles. Specifically, this paper examines two closed fuel cycles. The first fuel cycle examined is a thorium fuel cycle in which a pressurized heavy water reactor (PHWR) is fueled with mixtures of plutonium/thorium and {sup 233}U/thorium. The used fuel is then reprocessed using the THOREX process and the actinides are recycled. The second fuel cycle examined consists of conventional light water reactors (LWR) whose fuel is reprocessed for actinides that are then fed to and recycled untilmore » consumed in fast-spectrum reactors: fast reactors and accelerator driven systems (ADS). As reprocessing of LWR fuel has already been examined, this paper will focus on the reprocessing of the scheme's fast-spectrum reactors' fuel. This study will indicate what is required to render these materials as having low utility for use in nuclear weapons. Nevertheless, the results of this paper suggest that all reprocessing products evaluated so far need to be rigorously safeguarded and provided high levels of physical protection. These studies were performed at the request of the United States Department of Energy (DOE). The methodology and key findings will be presented.« less

An Integrated Fuel Depletion Calculator for Fuel Cycle Options Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schneider, Erich; Scopatz, Anthony

2016-04-25

Bright-lite is a reactor modeling software developed at the University of Texas Austin to expand upon the work done with the Bright [1] reactor modeling software. Originally, bright-lite was designed to function as a standalone reactor modeling software. However, this aim was refocused t couple bright-lite with the Cyclus fuel cycle simulator [2] to make it a module for the fuel cycle simulator.

Cooling Performance Analysis of ThePrimary Cooling System ReactorTRIGA-2000Bandung

NASA Astrophysics Data System (ADS)

Irianto, I. D.; Dibyo, S.; Bakhri, S.; Sunaryo, G. R.

2018-02-01

The conversion of reactor fuel type will affect the heat transfer process resulting from the reactor core to the cooling system. This conversion resulted in changes to the cooling system performance and parameters of operation and design of key components of the reactor coolant system, especially the primary cooling system. The calculation of the operating parameters of the primary cooling system of the reactor TRIGA 2000 Bandung is done using ChemCad Package 6.1.4. The calculation of the operating parameters of the cooling system is based on mass and energy balance in each coolant flow path and unit components. Output calculation is the temperature, pressure and flow rate of the coolant used in the cooling process. The results of a simulation of the performance of the primary cooling system indicate that if the primary cooling system operates with a single pump or coolant mass flow rate of 60 kg/s, it will obtain the reactor inlet and outlet temperature respectively 32.2 °C and 40.2 °C. But if it operates with two pumps with a capacity of 75% or coolant mass flow rate of 90 kg/s, the obtained reactor inlet, and outlet temperature respectively 32.9 °C and 38.2 °C. Both models are qualified as a primary coolant for the primary coolant temperature is still below the permitted limit is 49.0 °C.

The thermodynamics of pyrochemical processes for liquid metal reactor fuel cycles

DOE Office of Scientific and Technical Information (OSTI.GOV)

Johnson, I.

1987-01-01

The thermodynamic basis for pyrochemical processes for the recovery and purification of fuel for the liquid metal reactor fuel cycle is described. These processes involve the transport of the uranium and plutonium from one liquid alloy to another through a molten salt. The processes discussed use liquid alloys of cadmium, zinc, and magnesium and molten chloride salts. The oxidation-reduction steps are done either chemically by the use of an auxiliary redox couple or electrochemically by the use of an external electrical supply. The same basic thermodynamics apply to both the salt transport and the electrotransport processes. Large deviations from idealmore » solution behavior of the actinides and lanthanides in the liquid alloys have a major influence on the solubilities and the performance of both the salt transport and electrotransport processes. Separation of plutonium and uranium from each other and decontamination from the more noble fission product elements can be achieved using both transport processes. The thermodynamic analysis is used to make process design computations for different process conditions.« less

Dilution cycle control for an absorption refrigeration system

DOEpatents

Reimann, Robert C.

1984-01-01

A dilution cycle control system for an absorption refrigeration system is disclosed. The control system includes a time delay relay for sensing shutdown of the absorption refrigeration system and for generating a control signal only after expiration of a preselected time period measured from the sensed shutdown of the absorption refrigeration system, during which the absorption refrigeration system is not restarted. A dilution cycle for the absorption refrigeration system is initiated in response to generation of a control signal by the time delay relay. This control system is particularly suitable for use with an absorption refrigeration system which is frequently cycled on and off since the time delay provided by the control system prevents needless dilution of the absorption refrigeration system when the system is turned off for only a short period of time and then is turned back on.

Introduction to D-He(3) fusion reactors

NASA Technical Reports Server (NTRS)

Vlases, G. C.; Steinhauer, L. C.

1989-01-01

A review and evaluation of D-He(3) fusion reactor technology is presented. The advantages and disadvantages of the D-He(3) and D-T reactor cycles are outlined and compared. In addition, the general design features of D-He(3) tokamaks and field reversed configuration (FRC) reactors are described and the relative merits of each are compared. It is concluded that both tokamaks and FRC's offer certain advantages, and that the ultimate decision as to which to persue for terrestrial power generation will depend heavily on how the physics performance of each of them develops over the next few years. It is clear that the D-He(3) fuel cycle offers marked advantages over the D-T cycle. Although the physics requirements for D-He(3) are more demanding, the overwhelming advantages resulting from the two order of magnitude reduction of neutron flux are expected to lead to a shorter time to commercialization than for the D-T cycle.

Introduction to D-He(3) fusion reactors

NASA Astrophysics Data System (ADS)

Vlases, G. C.; Steinhauer, L. C.

1989-07-01

A review and evaluation of D-He(3) fusion reactor technology is presented. The advantages and disadvantages of the D-He(3) and D-T reactor cycles are outlined and compared. In addition, the general design features of D-He(3) tokamaks and field reversed configuration (FRC) reactors are described and the relative merits of each are compared. It is concluded that both tokamaks and FRC's offer certain advantages, and that the ultimate decision as to which to persue for terrestrial power generation will depend heavily on how the physics performance of each of them develops over the next few years. It is clear that the D-He(3) fuel cycle offers marked advantages over the D-T cycle. Although the physics requirements for D-He(3) are more demanding, the overwhelming advantages resulting from the two order of magnitude reduction of neutron flux are expected to lead to a shorter time to commercialization than for the D-T cycle.

Summary of space nuclear reactor power systems, 1983--1992

DOE Office of Scientific and Technical Information (OSTI.GOV)

Buden, D.

1993-08-11

This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts:were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressedmore » from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987--88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power.« less

Hollow Fiber Membrane Bioreactor Systems for Wastewater Processing: Effects of Environmental Stresses Including Dormancy Cycling and Antibiotic Dosing

NASA Technical Reports Server (NTRS)

Coutts, Janelle L.; Hummerick, Mary E.; Lunn, Griffin M.; Larson, Brian D.; Spencer, LaShelle E.; Kosiba, Michael L.; Khodadad, Christina L.; Catechis, John A.; Birmele, Michele N.; Wheeler, Raymond M.

2016-01-01

Membrane-aerated biofilm reactors (MABRs) have been studied for a number of years as an alternate approach for treating wastewater streams during space exploration. While the technology provides a promising pre-treatment for lowering organic carbon and nitrogen content without the need for harsh stabilization chemicals, several challenges must be addressed before adoption of the technology in future missions. One challenge is the transportation of bioreactors containing intact, active biofilms as a means for rapid start-up on the International Space Station or beyond. Similarly, there could be a need for placing these biological systems into a dormant state for extended periods when the system is not in use, along with the ability for rapid restart. Previous studies indicated that there was little influence of storage condition (4 or 25 C, with or without bulk fluid) on recovery of bioreactors with immature biofilms (48 days old), but that an extensive recovery time was required (20+ days). Bioreactors with fully established biofilms (13 months) were able to recover from a 7-month dormancy within 4 days (approximately 1 residence). Further dormancy and recovery testing is presented here that examines the role of biofilm age on recovery requirements, repeated dormancy cycle capabilities, and effects of long-duration dormancy cycles (8-9 months) on HFMB systems. Another challenge that must be addressed is the possibility of antibiotics entering the wastewater stream. Currently, for most laboratory tests of biological water processors, donors providing urine may not contribute to the study when taking antibiotics because the effects on the system are yet uncharacterized. A simulated urinary tract infection event, where an opportunistic, pathogenic organism, E. coli, was introduced to the HFMBs followed by dosing with an antibiotic, ciprofloxacin, was completed to study the effect of the antibiotic on reactor performance and to also examine the development of

The slightly-enriched spectral shift control reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Martin, W.R.; Lee, J.C.; Larsen, E.W.

1991-11-01

An advanced converter reactor design utilizing mechanical spectral shift control rods in a conventional pressurized water reactor configuration is under investigation. The design is based on the principle that a harder spectrum during the early part of the fuel cycle will result in large neutron captures in fertile {sup 238}U, which can then be burned in situ in a softer spectrum later in the cycle. Preliminary design calculations performed during FY 89 showed that the slightly-enriched spectral shift reactor design offers the benefit of substantially increased fuel resource utilization with the proven safety characteristics of the pressurized water reactor technologymore » retained. Optimization of the fuel design and development of fuel management strategies were carried out in FY 90, along with effort to develop and validate neutronic methodology for tight-lattice configurations with hard spectra. During FY 91, the final year of the grant, the final Slightly-Enriched Spectral Shift Reactor (SESSR) design was determined, and reference design analyses were performed for the assemblies as well as the global core configuration, both at the beginning of cycle (BOC) and with depletion. The final SESSR design results in approximately a 20% increase in the utilization of uranium resources, based on equilibrium fuel cycle analyses. Acceptable pin power peaking is obtained with the final core design, with assembly peaking factors equal to less than 1.04 for spectral shift control rods both inserted and withdrawn, and global peaking factors at BOC predicted to be 1.4. In addition, a negative Moderation Temperature Coefficient (MTC) is maintained for BOC, which is difficult to achieve with conventional advanced converter designs based on a closed fuel cycle. The SESSR design avoids the need for burnable poison absorber, although they could be added if desired to increase the cycle length while maintaining a negative MTC.« less

Production assurance program strategy for N Reactor balance of plant systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

House, R.D.; Bitten, E.J.; Keenan, J.P.

1986-03-18

A production assurance program has been established for N Reactor, a dual purpose reactor plant, operated to produce special nuclear materials and steam for electricity. N Reactor, which began operation in December 1963, is now approaching the end of its design life. This paper describes the two phase program for Balance of Plant (BOP) systems. The Phase I evaluation has been completed and indications are that the lifetime of systems and components could be extended by implementing appropriate surveillance, operations and maintenance strategies. In Phase II, a thorough evaluation of components and systems is underway and action items are beingmore » identified which will allow component and system extended operation.« less

POWER GENERATING NEUTRONIC REACTOR SYSTEM

DOEpatents

Vernon, H.C.

1958-03-01

This patent relates to reactor systems of the type wherein the cooiing medium is a liquid which is converted by the heat of the reaction to steam which is conveyed directly to a pnime mover such as a steam turbine driving a generatore after which it is condensed and returred to the coolant circuit. In this design, the reactor core is disposed within a tank for containing either a slurry type fuel or an aggregation of solid fuel elements such as elongated rods submerged in a liquid moderator such as heavy water. The top of the tank is provided with a nozzle which extends into an expansion chamber connected with the upper end of the tank, the coolant being maintained in the expansion chamber at a level above the nozzle and the steam being formed in the expansion chamber.

Economic Analysis of Complex Nuclear Fuel Cycles with NE-COST

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ganda, Francesco; Dixon, Brent; Hoffman, Edward

The purpose of this work is to present a new methodology, and associated computational tools, developed within the U.S. Department of Energy (U.S. DOE) Fuel Cycle Option Campaign to quantify the economic performance of complex nuclear fuel cycles. The levelized electricity cost at the busbar is generally chosen to quantify and compare the economic performance of different baseload generating technologies, including of nuclear: it is the cost of electricity which renders the risk-adjusted discounted net present value of the investment cash flow equal to zero. The work presented here is focused on the calculation of the levelized cost of electricitymore » of fuel cycles at mass balance equilibrium, which is termed LCAE (Levelized Cost of Electricity at Equilibrium). To alleviate the computational issues associated with the calculation of the LCAE for complex fuel cycles, a novel approach has been developed, which has been called the “island approach” because of its logical structure: a generic complex fuel cycle is subdivided into subsets of fuel cycle facilities, called islands, each containing one and only one type of reactor or blanket and an arbitrary number of fuel cycle facilities. A nuclear economic software tool, NE-COST, written in the commercial programming software MATLAB®, has been developed to calculate the LCAE of complex fuel cycles with the “island” computational approach. NE-COST has also been developed with the capability to handle uncertainty: the input parameters (both unit costs and fuel cycle characteristics) can have uncertainty distributions associated with them, and the output can be computed in terms of probability density functions of the LCAE. In this paper NE-COST will be used to quantify, as examples, the economic performance of (1) current Light Water Reactors (LWR) once-through systems; (2) continuous plutonium recycling in Fast Reactors (FR) with driver and blanket; (3) Recycling of plutonium bred in FR into LWR. For each

Containment system for supercritical water oxidation reactor

DOEpatents

Chastagner, Philippe

1994-01-01

A system for containment of a supercritical water oxidation reactor in the event of a rupture of the reactor. The system includes a containment for housing the reaction vessel and a communicating chamber for holding a volume of coolant, such as water. The coolant is recirculated and sprayed to entrain and cool any reactants that might have escaped from the reaction vessel. Baffles at the entrance to the chamber prevent the sprayed coolant from contacting the reaction vessel. An impact-absorbing layer is positioned between the vessel and the containment to at least partially absorb momentum of any fragments propelled by the rupturing vessel. Remote, quick-disconnecting fittings exterior to the containment, in cooperation with shut-off valves, enable the vessel to be isolated and the system safely taken off-line. Normally-closed orifices throughout the containment and chamber enable decontamination of interior surfaces when necessary.

Containment system for supercritical water oxidation reactor

DOEpatents

Chastagner, P.

1994-07-05

A system is described for containment of a supercritical water oxidation reactor in the event of a rupture of the reactor. The system includes a containment for housing the reaction vessel and a communicating chamber for holding a volume of coolant, such as water. The coolant is recirculated and sprayed to entrain and cool any reactants that might have escaped from the reaction vessel. Baffles at the entrance to the chamber prevent the sprayed coolant from contacting the reaction vessel. An impact-absorbing layer is positioned between the vessel and the containment to at least partially absorb momentum of any fragments propelled by the rupturing vessel. Remote, quick-disconnecting fittings exterior to the containment, in cooperation with shut-off valves, enable the vessel to be isolated and the system safely taken off-line. Normally-closed orifices throughout the containment and chamber enable decontamination of interior surfaces when necessary. 2 figures.

High-temperature Gas Reactor (HTGR)

NASA Astrophysics Data System (ADS)

Abedi, Sajad

2011-05-01

General Atomics (GA) has over 35 years experience in prismatic block High-temperature Gas Reactor (HTGR) technology design. During this period, the design has recently involved into a modular have been performed to demonstrate its versatility. This versatility is directly related to refractory TRISO coated - particle fuel that can contain any type of fuel. This paper summarized GA's fuel cycle studies individually and compares each based upon its cycle sustainability, proliferation-resistance capabilities, and other performance data against pressurized water reactor (PWR) fuel cycle data. Fuel cycle studies LEU-NV;commercial HEU-Th;commercial LEU-Th;weapons-grade plutonium consumption; and burning of LWR waste including plutonium and minor actinides in the MHR. results show that all commercial MHR options, with the exception of HEU-TH, are more sustainable than a PWR fuel cycle. With LEU-NV being the most sustainable commercial options. In addition, all commercial MHR options out perform the PWR with regards to its proliferation-resistance, with thorium fuel cycle having the best proliferation-resistance characteristics.

Rodded shutdown system for a nuclear reactor

DOEpatents

Golden, Martin P.; Govi, Aldo R.

1978-01-01

A top mounted nuclear reactor diverse rodded shutdown system utilizing gas fed into a pressure bearing bellows region sealed at the upper extremity to an armature. The armature is attached to a neutron absorber assembly by a series of shafts and connecting means. The armature is held in an uppermost position by an electromagnet assembly or by pressurized gas in a second embodiment. Deenergizing the electromagnet assembly, or venting the pressurized gas, causes the armature to fall by the force of gravity, thereby lowering the attached absorber assembly into the reactor core.

Computer study of emergency shutdowns of a 60-kilowatt reactor Brayton space power system

NASA Technical Reports Server (NTRS)

Tew, R. C.; Jefferies, K. S.

1974-01-01

A digital computer study of emergency shutdowns of a 60-kWe reactor Brayton power system was conducted. Malfunctions considered were (1) loss of reactor coolant flow, (2) loss of Brayton system gas flow, (3)turbine overspeed, and (4) a reactivity insertion error. Loss of reactor coolant flow was the most serious malfunction for the reactor. Methods for moderating the reactor transients due to this malfunction are considered.

FLOW SYSTEM FOR REACTOR

DOEpatents

Zinn, W.H.

1963-06-11

A reactor is designed with means for terminating the reaction when returning coolant is below a predetermined temperature. Coolant flowing from the reactor passes through a heat exchanger to a lower reservoir, and then circulates between the lower reservoir and an upper reservoir before being returned to the reactor. Means responsive to the temperature of the coolant in the return conduit terminate the chain reaction when the temperature reaches a predetermined minimum value. (AEC)

D-He-3 spherical torus fusion reactor system study

NASA Astrophysics Data System (ADS)

Macon, William A., Jr.

1992-04-01

This system study extrapolates present physics knowledge and technology to predict the anticipated characteristics of D-He3 spherical torus fusion reactors and their sensitivity to uncertainties in important parameters. Reference cases for steady-state 1000 MWe reactors operating in H-mode in both the 1st stability regime and the 2nd stability regime were developed and assessed quantitatively. These devices would a very small aspect ratio (A=1,2), a major radius of about 2.0 m, an on-axis magnetic field less than 2 T, a large plasma current (80-120 MA) dominated by the bootstrap effect, and high plasma beta (greater than O.6). The estimated cost of electricity is in the range of 60-90 mills/kW-hr, assuming the use of a direct energy conversion system. The inherent safety and environmental advantages of D-He3 fusion indicate that this reactor concept could be competitive with advanced fission breeder reactors and large-scale solar electric plants by the end of the 21st century if research and development can produce the anticipated physics and technology advances.

Research in Support of the Use of Rankine Cycle Energy Conversion Systems for Space Power and Propulsion

NASA Technical Reports Server (NTRS)

Lahey, Richard T., Jr.; Dhir, Vijay

2004-01-01

This is the report of a Scientific Working Group (SWG) formed by NASA to determine the feasibility of using a liquid metal cooled nuclear reactor and Rankine energy conversion cycle for dual purpose power and propulsion in space. This is a high level technical report which is intended for use by NASA management in program planning. The SWG was composed of a team of specialists in nuclear energy and multiphase flow and heat transfer technology from academia, national laboratories, NASA and industry. The SWG has identified the key technology issues that need to be addressed and have recommended an integrated short term (approx. 2 years) and a long term (approx. 10 year) research and development (R&D) program to qualify a Rankine cycle power plant for use in space. This research is ultimately intended to give NASA and its contractors the ability to reliably predict both steady and transient multiphase flow and heat transfer phenomena at reduced gravity, so they can analyze and optimize designs and scale-up experimental data on Rankine cycle components and systems. In addition, some of these results should also be useful for the analysis and design of various multiphase life support and thermal management systems being considered by NASA.

A Characteristics-Based Approach to Radioactive Waste Classification in Advanced Nuclear Fuel Cycles

NASA Astrophysics Data System (ADS)

Djokic, Denia

The radioactive waste classification system currently used in the United States primarily relies on a source-based framework. This has lead to numerous issues, such as wastes that are not categorized by their intrinsic risk, or wastes that do not fall under a category within the framework and therefore are without a legal imperative for responsible management. Furthermore, in the possible case that advanced fuel cycles were to be deployed in the United States, the shortcomings of the source-based classification system would be exacerbated: advanced fuel cycles implement processes such as the separation of used nuclear fuel, which introduce new waste streams of varying characteristics. To be able to manage and dispose of these potential new wastes properly, development of a classification system that would assign appropriate level of management to each type of waste based on its physical properties is imperative. This dissertation explores how characteristics from wastes generated from potential future nuclear fuel cycles could be coupled with a characteristics-based classification framework. A static mass flow model developed under the Department of Energy's Fuel Cycle Research & Development program, called the Fuel-cycle Integration and Tradeoffs (FIT) model, was used to calculate the composition of waste streams resulting from different nuclear fuel cycle choices: two modified open fuel cycle cases (recycle in MOX reactor) and two different continuous-recycle fast reactor recycle cases (oxide and metal fuel fast reactors). This analysis focuses on the impact of waste heat load on waste classification practices, although future work could involve coupling waste heat load with metrics of radiotoxicity and longevity. The value of separation of heat-generating fission products and actinides in different fuel cycles and how it could inform long- and short-term disposal management is discussed. It is shown that the benefits of reducing the short-term fission

Modeling and Depletion Simulations for a High Flux Isotope Reactor Cycle with a Representative Experiment Loading

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chandler, David; Betzler, Ben; Hirtz, Gregory John

2016-09-01

The purpose of this report is to document a high-fidelity VESTA/MCNP High Flux Isotope Reactor (HFIR) core model that features a new, representative experiment loading. This model, which represents the current, high-enriched uranium fuel core, will serve as a reference for low-enriched uranium conversion studies, safety-basis calculations, and other research activities. A new experiment loading model was developed to better represent current, typical experiment loadings, in comparison to the experiment loading included in the model for Cycle 400 (operated in 2004). The new experiment loading model for the flux trap target region includes full length 252Cf production targets, 75Se productionmore » capsules, 63Ni production capsules, a 188W production capsule, and various materials irradiation targets. Fully loaded 238Pu production targets are modeled in eleven vertical experiment facilities located in the beryllium reflector. Other changes compared to the Cycle 400 model are the high-fidelity modeling of the fuel element side plates and the material composition of the control elements. Results obtained from the depletion simulations with the new model are presented, with a focus on time-dependent isotopic composition of irradiated fuel and single cycle isotope production metrics.« less

System and method for air temperature control in an oxygen transport membrane based reactor

DOEpatents

Kelly, Sean M

2016-09-27

A system and method for air temperature control in an oxygen transport membrane based reactor is provided. The system and method involves introducing a specific quantity of cooling air or trim air in between stages in a multistage oxygen transport membrane based reactor or furnace to maintain generally consistent surface temperatures of the oxygen transport membrane elements and associated reactors. The associated reactors may include reforming reactors, boilers or process gas heaters.

Plasma generators, reactor systems and related methods

DOEpatents

Kong, Peter C [Idaho Falls, ID; Pink, Robert J [Pocatello, ID; Lee, James E [Idaho Falls, ID

2007-06-19

A plasma generator, reactor and associated systems and methods are provided in accordance with the present invention. A plasma reactor may include multiple sections or modules which are removably coupled together to form a chamber. Associated with each section is an electrode set including three electrodes with each electrode being coupled to a single phase of a three-phase alternating current (AC) power supply. The electrodes are disposed about a longitudinal centerline of the chamber and are arranged to provide and extended arc and generate an extended body of plasma. The electrodes are displaceable relative to the longitudinal centerline of the chamber. A control system may be utilized so as to automatically displace the electrodes and define an electrode gap responsive to measure voltage or current levels of the associated power supply.

Passive cooling system for top entry liquid metal cooled nuclear reactors

DOEpatents

Boardman, Charles E.; Hunsbedt, Anstein; Hui, Marvin M.

1992-01-01

A liquid metal cooled nuclear fission reactor plant having a top entry loop joined satellite assembly with a passive auxiliary safety cooling system for removing residual heat resulting from fuel decay during shutdown, or heat produced during a mishap. This satellite type reactor plant is enhanced by a backup or secondary passive safety cooling system which augments the primary passive auxiliary cooling system when in operation, and replaces the primary cooling system when rendered inoperative.

Performance Analyses of 38 kWe Turbo-Machine Unit for Space Reactor Power Systems

NASA Astrophysics Data System (ADS)

Gallo, Bruno M.; El-Genk, Mohamed S.

2008-01-01

This paper developed a design and investigated the performance of 38 kWe turbo-machine unit for space nuclear reactor power systems with Closed Brayton Cycle (CBC) energy conversion. The compressor and turbine of this unit are scaled versions of the NASA's BRU developed in the sixties and seventies. The performance results of turbo-machine unit are calculated for rotational speed up to 45 krpm, variable reactor thermal power and system pressure, and fixed turbine and compressor inlet temperatures of 1144 K and 400 K. The analyses used a detailed turbo-machine model developed at the University of New Mexico that accounts for the various energy losses in the compressor and turbine and the effect of compressibility of the He-Xe (40 mole/g) working fluid with increased flow rate. The model also accounts for the changes in the physical and transport properties of the working fluid with temperature and pressure. Results show that a unit efficiency of 24.5% is achievable at rotation speed of 45 krpm and system pressure of 0.75 MPa, assuming shaft and electrical generator efficiencies of 86.7% and 90%. The corresponding net electric power output of the unit is 38.5 kWe, the flow rate of the working fluid is 1.667 kg/s, the pressure ratio and polytropic efficiency for the compressor are 1.60 and 83.1%, and 1.51 and 88.3% for the turbine.

THE COOLING REQUIREMENTS AND PROCESS SYSTEMS OF THE SOUTH AFRICAN RESEARCH REACTOR, SAFARI 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Colley, J.R.

1962-12-01

The SAFARI 1 research reactor is cooled and moderated by light water. There are three process systems, a primary water system which cools the reactor core and surroundings, a pool water system, and a secondary water system which removes the heat from the primary and pool systems. The cooling requirements for the reactor core and experimental facilities are outlined, and the cooling and purification functions of the three process systems are described. (auth)

Information system life-cycle and documentation standards, volume 1

NASA Technical Reports Server (NTRS)

Callender, E. David; Steinbacher, Jody

1989-01-01

The Software Management and Assurance Program (SMAP) Information System Life-Cycle and Documentation Standards Document describes the Version 4 standard information system life-cycle in terms of processes, products, and reviews. The description of the products includes detailed documentation standards. The standards in this document set can be applied to the life-cycle, i.e., to each phase in the system's development, and to the documentation of all NASA information systems. This provides consistency across the agency as well as visibility into the completeness of the information recorded. An information system is software-intensive, but consists of any combination of software, hardware, and operational procedures required to process, store, or transmit data. This document defines a standard life-cycle model and content for associated documentation.

Passive cooling system for liquid metal cooled nuclear reactors with backup coolant flow path

DOEpatents

Hunsbedt, Anstein; Boardman, Charles E.

1993-01-01

A liquid metal cooled nuclear fission reactor plant having a passive auxiliary safety cooling system for removing residual heat resulting from fuel decay during reactor shutdown, or heat produced during a mishap. This reactor plant is enhanced by a backup or secondary passive safety cooling system which augments the primary passive auxiliary cooling system when in operation, and replaces the primary system when rendered inoperable.

NUCLEAR REACTOR FUEL SYSTEMS

DOEpatents

Thamer, B.J.; Bidwell, R.M.; Hammond, R.P.

1959-09-15

Homogeneous reactor fuel solutions are reported which provide automatic recombination of radiolytic gases and exhibit large thermal expansion characteristics, thereby providing stability at high temperatures and enabling reactor operation without the necessity of apparatus to recombine gases formed by the radiolytic dissociation of water in the fuel and without the necessity of liquid fuel handling outside the reactor vessel except for recovery processes. The fuels consist of phosphoric acid and water solutions of enriched uranium, wherein the uranium is in either the hexavalent or tetravalent state.

Open-cycle systems performance analysis programming guide

DOE Office of Scientific and Technical Information (OSTI.GOV)

Olson, D.A.

1981-12-01

The Open-Cycle OTEC Systems Performance Analysis Program is an algorithm programmed on SERI's CDC Cyber 170/720 computer to predict the performance of a Claude-cycle, open-cycle OTEC plant. The algorithm models the Claude-cycle system as consisting of an evaporator, a turbine, a condenser, deaerators, a condenser gas exhaust, a cold water pipe and cold and warm seawater pumps. Each component is a separate subroutine in the main program. A description is given of how to write Fortran subroutines to fit into the main program for the components of the OTEC plant. An explanation is provided of how to use the algorithm.more » The main program and existing component subroutines are described. Appropriate common blocks and input and output variables are listed. Preprogrammed thermodynamic property functions for steam, fresh water, and seawater are described.« less

Passive cooling system for nuclear reactor containment structure

DOEpatents

Gou, Perng-Fei; Wade, Gentry E.

1989-01-01

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

Evaluation of some candidate materials for automobile thermal reactors in engine-dynamometer screening tests

NASA Technical Reports Server (NTRS)

Oldrieve, R. E.

1971-01-01

Fourteen materials were evaluated in engine screening tests on full-size thermal reactors for automobile engine pollution control systems. Cyclic test-stand engine operation provided 2 hours at 1040 C and a 20-minute air-cool to 70 C each test cycle. Each reactor material was exposed to 83 cycles in 200 hours of engine testing. On the basis of resistance to oxidation and distortion, the best materials included two ferritic iron alloys (Ge 1541 and Armco 18S/R), several commercial oxidation-resistant coatings on AlSl 651 (19-9 DL), and possibly uncoated AISI 310. The best commercial coatings were Cr-Al, Ni-Cr, and a glass ceramic.

Analysis of reactor trips originating in balance of plant systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

1990-09-01

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

Nuclear reactor insulation and preheat system

DOEpatents

Wampole, Nevin C.

1978-01-01

An insulation and preheat system for preselected components of a fluid cooled nuclear reactor. A gas tight barrier or compartment of thermal insulation surrounds the selected components and includes devices to heat the internal atmosphere of the compartment. An external surface of the compartment or enclosure is cooled, such as by a circulating fluid. The heating devices provide for preheating of the components, as well as maintenance of a temperature sufficient to ensure that the reactor coolant fluid will not solidify during shutdown. The external cooling limits the heat transferred to other plant structures, such as supporting concrete and steel. The barrier is spaced far enough from the surrounded components so as to allow access for remote or manual inspection, maintenance, and repair.

Advantages of liquid fluoride thorium reactor in comparison with light water reactor

NASA Astrophysics Data System (ADS)

Bahri, Che Nor Aniza Che Zainul; Majid, Amran Ab.; Al-Areqi, Wadeeah M.

2015-04-01

Liquid Fluoride Thorium Reactor (LFTR) is an innovative design for the thermal breeder reactor that has important potential benefits over the traditional reactor design. LFTR is fluoride based liquid fuel, that use the thorium dissolved in salt mixture of lithium fluoride and beryllium fluoride. Therefore, LFTR technology is fundamentally different from the solid fuel technology currently in use. Although the traditional nuclear reactor technology has been proven, it has perceptual problems with safety and nuclear waste products. The aim of this paper is to discuss the potential advantages of LFTR in three aspects such as safety, fuel efficiency and nuclear waste as an alternative energy generator in the future. Comparisons between LFTR and Light Water Reactor (LWR), on general principles of fuel cycle, resource availability, radiotoxicity and nuclear weapon proliferation shall be elaborated.

ADVANCED NUCLEAR FUEL CYCLE EFFECTS ON THE TREATMENT OF UNCERTAINTY IN THE LONG-TERM ASSESSMENT OF GEOLOGIC DISPOSAL SYSTEMS - EBS INPUT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sutton, M; Blink, J A; Greenberg, H R

2012-04-25

encapsulated in borosilicate glass. Because the heat load of the glass was much less than the PWR and BWR assemblies, the glass waste form was able to be co-disposed with the open cycle waste, by interspersing glass waste packages among the spent fuel assembly waste packages. In addition, the Yucca Mountain repository was designed to include some research reactor spent fuel and naval reactor spent fuel, within the envelope that was set using the commercial reactor assemblies as the design basis waste form. This milestone report supports Sandia National Laboratory milestone M2FT-12SN0814052, and is intended to be a chapter in that milestone report. The independent technical review of this LLNL milestone was performed at LLNL and is documented in the electronic Information Management (IM) system at LLNL. The objective of this work is to investigate what aspects of quantifying, characterizing, and representing the uncertainty associated with the engineered barrier are affected by implementing different advanced nuclear fuel cycles (e.g., partitioning and transmutation scenarios) together with corresponding designs and thermal constraints.« less

Rankine cycle waste heat recovery system

DOEpatents

Ernst, Timothy C.; Nelson, Christopher R.

2016-05-10

This disclosure relates to a waste heat recovery (WHR) system and to a system and method for regulation of a fluid inventory in a condenser and a receiver of a Rankine cycle WHR system. Such regulation includes the ability to regulate the pressure in a WHR system to control cavitation and energy conversion.

Rankine cycle waste heat recovery system

DOEpatents

Ernst, Timothy C.; Nelson, Christopher R.

2014-08-12

This disclosure relates to a waste heat recovery (WHR) system and to a system and method for regulation of a fluid inventory in a condenser and a receiver of a Rankine cycle WHR system. Such regulation includes the ability to regulate the pressure in a WHR system to control cavitation and energy conversion.

Enhancing power cycle efficiency for a supercritical Brayton cycle power system using tunable supercritical gas mixtures

DOEpatents

Wright, Steven A.; Pickard, Paul S.; Vernon, Milton E.; Radel, Ross F.

2017-08-29

Various technologies pertaining to tuning composition of a fluid mixture in a supercritical Brayton cycle power generation system are described herein. Compounds, such as Alkanes, are selectively added or removed from an operating fluid of the supercritical Brayton cycle power generation system to cause the critical temperature of the fluid to move up or down, depending upon environmental conditions. As efficiency of the supercritical Brayton cycle power generation system is substantially optimized when heat is rejected near the critical temperature of the fluid, dynamically modifying the critical temperature of the fluid based upon sensed environmental conditions improves efficiency of such a system.

A CAMAC based real-time noise analysis system for nuclear reactors

NASA Astrophysics Data System (ADS)

Ciftcioglu, Özer

1987-05-01

A CAMAC based real-time noise analysis system was designed for the TRIGA MARK II nuclear reactor at the Institute for Nuclear Energy, Istanbul. The input analog signals obtained from the radiation detectors are introduced to the system through CAMAC interface. The signals converted into digital form are processed by a PDP-11 computer. The fast data processing based on auto/cross power spectral density computations is carried out by means of assembly written FFT algorithms in real-time and the spectra obtained are displayed on a CAMAC driven display system as an additional monitoring device. The system has the advantage of being software programmable and controlled by a CAMAC system so that it is operated under program control for reactor surveillance, anomaly detection and diagnosis. The system can also be used for the identification of nonstationary operational characteristics of the reactor in long term by comparing the noise power spectra with the corresponding reference noise patterns prepared in advance.

PARTIAL ECONOMIC STUDY OF STEAM COOLED HEAVY WATER MODERATED REACTORS

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1960-04-01

Steam-cooled reactors are compared with CAHDU for costs of Calandria tubes, pressure tubes. heavy water moderator, heavy water reflector, fuel supply, heat exchanger, and turbine generator. A direct-cycle lightsteam-cooled heavy- water-moderated pressure-tube reactor formed the basic reactor design for the study. Two methods of steam circulation through the reactor were examined. In both cases the steam was generated outside the reactor and superheated in the reactor core. One method consisted of a series of reactor and steam generator passes. The second method consisted of the Loeffler cycle and its modifications. The fuel was assumed to be natural cylindrical UO/sub 2/more » pellets sheathed in a hypothetical material with the nuclear properties of Zircaloy, but able to function at temperatures to 900 deg F. For the conditions assumed, the longer the rod, the higher the outlet temperature and therefore the higher the efficiency. The turbine cycle efficiency was calculated on the assumption that suitable steam generators are available. As the neutron losses to the pressure tubes were significant, an economic analysis of insulated pressure tubes is included. A description of the physics program for steam-cooled reactors is included. Results indicated that power from the steam-cooled reactor would cost 1.4 mills/ kwh compared with 1.25 mills/kwh for CANDU. (M.C.G.)« less

Multi-Purpose Thermal Hydraulic Loop: Advanced Reactor Technology Integral System Test (ARTIST) Facility for Support of Advanced Reactor Technologies

DOE Office of Scientific and Technical Information (OSTI.GOV)

James E. O'Brien; Piyush Sabharwall; SuJong Yoon

2001-11-01

Effective and robust high temperature heat transfer systems are fundamental to the successful deployment of advanced reactors for both power generation and non-electric applications. Plant designs often include an intermediate heat transfer loop (IHTL) with heat exchangers at either end to deliver thermal energy to the application while providing isolation of the primary reactor system. In order to address technical feasibility concerns and challenges a new high-temperature multi-fluid, multi-loop test facility “Advanced Reactor Technology Integral System Test facility” (ARTIST) is under development at the Idaho National Laboratory. The facility will include three flow loops: high-temperature helium, molten salt, and steam/water.more » Details of some of the design aspects and challenges of this facility, which is currently in the conceptual design phase, are discussed« less

Sustainability Efficiency Factor: Measuring Sustainability in Advanced Energy Systems through Exergy, Exergoeconomic, Life Cycle, and Economic Analyses

NASA Astrophysics Data System (ADS)

Boldon, Lauren

The Encyclopedia of Life Support Systems defines sustainability or industrial ecology as "the wise use of resources through critical attention to policy, social, economic, technological, and ecological management of natural and human engineered capital so as to promote innovations that assure a higher degree of human needs fulfilment, or life support, across all regions of the world, while at the same time ensuring intergenerational equity" (Encyclopedia of Life Support Systems 1998). Developing and integrating sustainable energy systems to meet growing energy demands is a daunting task. Although the technology to utilize renewable energies is well understood, there are limited locations which are ideally suited for renewable energy development. Even in areas with significant wind or solar availability, backup or redundant energy supplies are still required during periods of low renewable generation. This is precisely why it would be difficult to make the switch directly from fossil fuel to renewable energy generation. A transition period in which a base-load generation supports renewables is required, and nuclear energy suits this need well with its limited life cycle emissions and fuel price stability. Sustainability is achieved by balancing environmental, economic, and social considerations, such that energy is produced without detriment to future generations through loss of resources, harm to the environment, etcetera. In essence, the goal is to provide future generations with the same opportunities to produce energy that the current generation has. This research explores sustainability metrics as they apply to a small modular reactor (SMR)-hydrogen production plant coupled with wind energy and storage technologies to develop a new quantitative sustainability metric, the Sustainability Efficiency Factor (SEF), for comparison of energy systems. The SEF incorporates the three fundamental aspects of sustainability and provides SMR or nuclear hybrid energy system

Advanced Small Modular Reactor Economics Status Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harrison, Thomas J.

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 andmore » 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

A microprocessor tester for the treat upgrade reactor trip system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lenkszus, F.R.; Bucher, R.G.

1985-02-01

The upgrading of the Transient Reactor Test (TREAT) Facility at ANL-Idaho has been designed to provide additional experimental capabilities for the study of core disruptive accident (CDA) phenomena. To improve the analytical extrapolation of test results to full-size assembly bundles, the facility upgrade will increase the maximum size of the test bundle from 7 to 37 fuel pins. By creating a core convertor zone around the test location, the neutron spectrum incident on the test assembly will be hardened and the maximum energy deposited in the sample will be increased. In addition, a programmable Automated Reactor Control System (ARCS) willmore » permit high-power transients up to 11,000 MW having a controlled reactor period of from 15 to 0.1 sec. These modifications to the core neutronics will improve simulation of LMFBR accident conditions. Finally, a sophisticated, multiply-redundant safety system, the Reactor Trip System (RTS), will provide safe operation for both steady state and transient production operating modes. To insure that this complex safety system is functioning properly, a Dedicated Microprocessor Tester (DMT) has been implemented to perform a thorough checkout of the RTS prior to all TREAT operations. A quantitative reliability analysis of the RTS shows that the unreliability, that is, the probability of failure, is acceptable for a 10 hour mission time or risk interval.« less

Variable pressure power cycle and control system

DOEpatents

Goldsberry, Fred L.

1984-11-27

A variable pressure power cycle and control system that is adjustable to a variable heat source is disclosed. The power cycle adjusts itself to the heat source so that a minimal temperature difference is maintained between the heat source fluid and the power cycle working fluid, thereby substantially matching the thermodynamic envelope of the power cycle to the thermodynamic envelope of the heat source. Adjustments are made by sensing the inlet temperature of the heat source fluid and then setting a superheated vapor temperature and pressure to achieve a minimum temperature difference between the heat source fluid and the working fluid.

VERA Core Simulator Methodology for PWR Cycle Depletion

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kochunas, Brendan; Collins, Benjamin S; Jabaay, Daniel

2015-01-01

This paper describes the methodology developed and implemented in MPACT for performing high-fidelity pressurized water reactor (PWR) multi-cycle core physics calculations. MPACT is being developed primarily for application within the Consortium for the Advanced Simulation of Light Water Reactors (CASL) as one of the main components of the VERA Core Simulator, the others being COBRA-TF and ORIGEN. The methods summarized in this paper include a methodology for performing resonance self-shielding and computing macroscopic cross sections, 2-D/1-D transport, nuclide depletion, thermal-hydraulic feedback, and other supporting methods. These methods represent a minimal set needed to simulate high-fidelity models of a realistic nuclearmore » reactor. Results demonstrating this are presented from the simulation of a realistic model of the first cycle of Watts Bar Unit 1. The simulation, which approximates the cycle operation, is observed to be within 50 ppm boron (ppmB) reactivity for all simulated points in the cycle and approximately 15 ppmB for a consistent statepoint. The verification and validation of the PWR cycle depletion capability in MPACT is the focus of two companion papers.« less

Nuclear reactor with makeup water assist from residual heat removal system

DOEpatents

Corletti, Michael M.; Schulz, Terry L.

1993-01-01

A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path.

Nuclear reactor with makeup water assist from residual heat removal system

DOEpatents

Corletti, M.M.; Schulz, T.L.

1993-12-07

A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path. 2 figures.

Lessons Learned about Liquid Metal Reactors from FFTF Experience

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wootan, David W.; Casella, Andrew M.; Omberg, Ronald P.

2016-09-20

The Fast Flux Test Facility (FFTF) is the most recent liquid-metal reactor (LMR) to operate in the United States, from 1982 to 1992. FFTF is located on the DOE Hanford Site near Richland, Washington. The 400-MWt sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission test reactor was designed specifically to irradiate Liquid Metal Fast Breeder Reactor (LMFBR) fuel and components in prototypical temperature and flux conditions. FFTF played a key role in LMFBR development and testing activities. The reactor provided extensive capability for in-core irradiation testing, including eight core positions that could be used with independent instrumentation for the test specimens.more » In addition to irradiation testing capabilities, FFTF provided long-term testing and evaluation of plant components and systems for LMFBRs. The FFTF was highly successful and demonstrated outstanding performance during its nearly 10 years of operation. The technology employed in designing and constructing this reactor, as well as information obtained from tests conducted during its operation, can significantly influence the development of new advanced reactor designs in the areas of plant system and component design, component fabrication, fuel design and performance, prototype testing, site construction, and reactor operations. The FFTF complex included the reactor, as well as equipment and structures for heat removal, containment, core component handling and examination, instrumentation and control, and for supplying utilities and other essential services. The FFTF Plant was designed using a “system” concept. All drawings, specifications and other engineering documentation were organized by these systems. Efforts have been made to preserve important lessons learned during the nearly 10 years of reactor operation. A brief summary of Lessons Learned in the following areas will be discussed: Acceptance and Startup Testing of FFTF FFTF Cycle Reports« less

Advantages of liquid fluoride thorium reactor in comparison with light water reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bahri, Che Nor Aniza Che Zainul, E-mail: anizazainul@gmail.com; Majid, Amran Ab.; Al-Areqi, Wadeeah M.

2015-04-29

Liquid Fluoride Thorium Reactor (LFTR) is an innovative design for the thermal breeder reactor that has important potential benefits over the traditional reactor design. LFTR is fluoride based liquid fuel, that use the thorium dissolved in salt mixture of lithium fluoride and beryllium fluoride. Therefore, LFTR technology is fundamentally different from the solid fuel technology currently in use. Although the traditional nuclear reactor technology has been proven, it has perceptual problems with safety and nuclear waste products. The aim of this paper is to discuss the potential advantages of LFTR in three aspects such as safety, fuel efficiency and nuclearmore » waste as an alternative energy generator in the future. Comparisons between LFTR and Light Water Reactor (LWR), on general principles of fuel cycle, resource availability, radiotoxicity and nuclear weapon proliferation shall be elaborated.« less

An Expert Elicitation of the Proliferation Resistance of Using Small Modular Reactors (SMR) for the Expansion of Civilian Nuclear Systems.

PubMed

Siegel, Jonas; Gilmore, Elisabeth A; Gallagher, Nancy; Fetter, Steve

2018-02-01

To facilitate the use of nuclear energy globally, small modular reactors (SMRs) may represent a viable alternative or complement to large reactor designs. One potential benefit is that SMRs could allow for more proliferation resistant designs, manufacturing arrangements, and fuel-cycle practices at widespread deployment. However, there is limited work evaluating the proliferation resistance of SMRs, and existing proliferation assessment approaches are not well suited for these novel arrangements. Here, we conduct an expert elicitation of the relative proliferation resistance of scenarios for future nuclear energy deployment driven by Generation III+ light-water reactors, fast reactors, or SMRs. Specifically, we construct the scenarios to investigate relevant technical and institutional features that are postulated to enhance the proliferation resistance of SMRs. The experts do not consistently judge the scenario with SMRs to have greater overall proliferation resistance than scenarios that rely on conventional nuclear energy generation options. Further, the experts disagreed on whether incorporating a long-lifetime sealed core into an SMR design would strengthen or weaken proliferation resistance. However, regardless of the type of reactor, the experts judged that proliferation resistance would be enhanced by improving international safeguards and operating several multinational fuel-cycle facilities rather than supporting many more national facilities. © 2017 Society for Risk Analysis.

Calculation of Purification Systems of Hydrocarbonmoderated Reactors; CALCULO DE SISTEMAS DE PURIFICATION DE REACTORES MODERADOS FOR HIDROCARBUROS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Santos, A.A.

1958-01-01

culation of Purification Systems of Hydrocarbonmoderated Reactors). Agustin Alonso Santos. 1958. 23p. As as introduction to the calculation of the purification systems of bydrocarbon-moderated reactors, the effects of heat and radiation on the polyphenols are considered. The chemical, physical, and nuclear properties are tabulated. The formation velocity of the polymers and gases, pyrolysis, effects of heat on the polymer, and the activity accumulated in the moderator ars discussed. The calculation is based on the hypetheses that the radiation catalyzes the formation of polymers, the velocity of the polymerization reaction is constant, the polymer concentration is maintained at a limit whichmore » does not adversely affect the heat transfer properties, the velocity of the separation of polymers in the distillation column is in proportion to their concentration in the hydrocarbon and the pyrolysis causes gaseous products. Formulas are derived expressing the purified flow and the activities accumulated in the distillation residues. The results are applied to the parification system of the Organic Moderated Reactor Experiment (J.S.R.)« less

Analysis of space reactor system components: Investigation through simulation and non-nuclear testing

NASA Astrophysics Data System (ADS)

Bragg-Sitton, Shannon M.

The use of fission energy in space power and propulsion systems offers considerable advantages over chemical propulsion. Fission provides over six orders of magnitude higher energy density, which translates to higher vehicle specific impulse and lower specific mass. These characteristics enable ambitious space exploration missions. The natural space radiation environment provides an external source of protons and high energy, high Z particles that can result in the production of secondary neutrons through interactions in reactor structures. Applying the approximate proton source in geosynchronous orbit during a solar particle event, investigation using MCNPX 2.5.b for proton transport through the SAFE-400 heat pipe cooled reactor indicates an incoming secondary neutron current of (1.16 +/- 0.03) x 107 n/s at the core-reflector interface. This neutron current may affect reactor operation during low power maneuvers (e.g., start-up) and may provide a sufficient reactor start-up source. It is important that a reactor control system be designed to automatically adjust to changes in reactor power levels, maintaining nominal operation without user intervention. A robust, autonomous control system is developed and analyzed for application during reactor start-up, accounting for fluctuations in the radiation environment that result from changes in vehicle location or to temporal variations in the radiation field. Development of a nuclear reactor for space applications requires a significant amount of testing prior to deployment of a flight unit. High confidence in fission system performance can be obtained through relatively inexpensive non-nuclear tests performed in relevant environments, with the heat from nuclear fission simulated using electric resistance heaters. A series of non-nuclear experiments was performed to characterize various aspects of reactor operation. This work includes measurement of reactor core deformation due to material thermal expansion and

Measurement and Analysis of Structural Integrity of Reactor Core Support Structure in Pressurized Water Reactor (PWR) Plant

NASA Astrophysics Data System (ADS)

Ansari, Saleem A.; Haroon, Muhammad; Rashid, Atif; Kazmi, Zafar

2017-02-01

Extensive calculation and measurements of flow-induced vibrations (FIV) of reactor internals were made in a PWR plant to assess the structural integrity of reactor core support structure against coolant flow. The work was done to meet the requirements of the Fukushima Response Action Plan (FRAP) for enhancement of reactor safety, and the regulatory guide RG-1.20. For the core surveillance measurements the Reactor Internals Vibration Monitoring System (IVMS) has been developed based on detailed neutron noise analysis of the flux signals from the four ex-core neutron detectors. The natural frequencies, displacement and mode shapes of the reactor core barrel (CB) motion were determined with the help of IVMS. The random pressure fluctuations in reactor coolant flow due to turbulence force have been identified as the predominant cause of beam-mode deflection of CB. The dynamic FIV calculations were also made to supplement the core surveillance measurements. The calculational package employed the computational fluid dynamics, mode shape analysis, calculation of power spectral densities of flow & pressure fields and the structural response to random flow excitation forces. The dynamic loads and stiffness of the Hold-Down Spring that keeps the core structure in position against upward coolant thrust were also determined by noise measurements. Also, the boron concentration in primary coolant at any time of the core cycle has been determined with the IVMS.

Analysis of an open cycle gas core nuclear propulsion system using MHD driven vortices for fuel containment

NASA Astrophysics Data System (ADS)

Sedwick, Raymond John

1998-12-01

A novel method for containing gaseous uranium vapor in an open cycle nuclear space propulsion system is developed. In an attempt to increase the operating temperature of the nuclear reactor beyond the melting point of solid fuel rods (thus increasing specific impulse), the fuel is instead suspended as a vapor in the propellant using the pressure forces developed in a confined vortex flow. The introduction of the fuel as uranium hexafluoride is found to be effective in maintaining its vapor phase in the feed passages from the tank, but not in the main vortex. A mechanism by which the resulting condensation of the uranium may be tolerated is identified, and the electro- optical properties of the resulting mixture are investigated. Containment is modeled using a 1D- axisymmetric geometry, and radiative heat transfer is found to restrict the maximum specific impulse of the system to 1500 seconds using pumping pressures of 500 atm. The specific impulse is related to this pressure as pm1/4, allowing only marginal increases in Isp at increased pressure levels. Additional 2D- axisymmetric issues, such as non-uniform current distribution and bypass flows through the boundary layers, are investigated, with possible methods of solution cited. A two-group, two-region reactor analysis is performed, estimating the mass of the reactor to be about 10 metric tonnes, and establishing the thrust to weight ratio achievable by the system at about 50. To reduce the mass of the power system, a scheme for using cross-flow heat exchange with the propellant flow to minimize (and possibly eliminate) the need for radiators to reject waste heat is presented. (Copies available exclusively from MIT Libraries, Rm. 14-0551, Cambridge, MA 02139-4307. Ph. 617-253-5668; Fax 617-253-1690.)

Rankine cycle system and method

DOEpatents

Ernst, Timothy C.; Nelson, Christopher R.

2014-09-09

A Rankine cycle waste heat recovery system uses a receiver with a maximum liquid working fluid level lower than the minimum liquid working fluid level of a sub-cooler of the waste heat recovery system. The receiver may have a position that is physically lower than the sub-cooler's position. A valve controls transfer of fluid between several of the components in the waste heat recovery system, especially from the receiver to the sub-cooler. The system may also have an associated control module.

An Analysis of Methanol and Hydrogen Production via High-Temperature Electrolysis Using the Sodium Cooled Advanced Fast Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shannon M. Bragg-Sitton; Richard D. Boardman; Robert S. Cherry

2014-03-01

Integration of an advanced, sodium-cooled fast spectrum reactor into nuclear hybrid energy system (NHES) architectures is the focus of the present study. A techno-economic evaluation of several conceptual system designs was performed for the integration of a sodium-cooled Advanced Fast Reactor (AFR) with the electric grid in conjunction with wind-generated electricity. Cases in which excess thermal and electrical energy would be reapportioned within an integrated energy system to a chemical plant are presented. The process applications evaluated include hydrogen production via high temperature steam electrolysis and methanol production via steam methane reforming to produce carbon monoxide and hydrogen which feedmore » a methanol synthesis reactor. Three power cycles were considered for integration with the AFR, including subcritical and supercritical Rankine cycles and a modified supercritical carbon dioxide modified Brayton cycle. The thermal efficiencies of all of the modeled power conversions units were greater than 40%. A thermal efficiency of 42% was adopted in economic studies because two of the cycles either performed at that level or could potentially do so (subcritical Rankine and S-CO2 Brayton). Each of the evaluated hybrid architectures would be technically feasible but would demonstrate a different internal rate of return (IRR) as a function of multiple parameters; all evaluated configurations showed a positive IRR. As expected, integration of an AFR with a chemical plant increases the IRR when “must-take” wind-generated electricity is added to the energy system. Additional dynamic system analyses are recommended to draw detailed conclusions on the feasibility and economic benefits associated with AFR-hybrid energy system operation.« less

Mechanical strength of an ITER coil insulation system under static and dynamic load after reactor irradiation

NASA Astrophysics Data System (ADS)

Bittner-Rohrhofer, K.; Humer, K.; Weber, H. W.; Hamada, K.; Sugimoto, M.; Okuno, K.

2002-12-01

The insulation system proposed by the Japanese Home Team for the ITER Toroidal Field coil (TF coil) is a T-glass-fiber/Kapton reinforced epoxy prepreg system. In order to assess the material performance under the actual operating conditions of the coils, the insulation system was irradiated in the TRIGA reactor (Vienna) to a fast neutron fluence of 2×10 22 m -2 ( E>0.1 MeV). After measurements of swelling, all mechanical tests were carried out at 77 K. Tensile and short-beam-shear (SBS) tests were performed under static loading conditions. In addition, tension-tension fatigue experiments up to about 10 6 cycles were made. The laminate swells in the through-thickness direction by 0.86% at the highest dose level. The fatigue tests as well as the static tests do not show significant influences of the irradiation on the mechanical behavior of this composite.

Method and apparatus for enhancing reactor air-cooling system performance

DOEpatents

Hunsbedt, Anstein

1996-01-01

An enhanced decay heat removal system for removing heat from the inert gas-filled gap space between the reactor vessel and the containment vessel of a liquid metal-cooled nuclear reactor. Multiple cooling ducts in flow communication with the inert gas-filled gap space are incorporated to provide multiple flow paths for the inert gas to circulate to heat exchangers which remove heat from the inert gas, thereby introducing natural convection flows in the inert gas. The inert gas in turn absorbs heat directly from the reactor vessel by natural convection heat transfer.

Transport Reactor Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Berry, D.A.; Shoemaker, S.A.

1996-12-31

The Morgantown Energy Technology Center (METC) is currently evaluating hot gas desulfurization (HGD)in its on-site transport reactor facility (TRF). This facility was originally constructed in the early 1980s to explore advanced gasification processes with an entrained reactor, and has recently been modified to incorporate a transport riser reactor. The TRF supports Integrated Gasification Combined Cycle (IGCC) power systems, one of METC`s advanced power generation systems. The HGD subsystem is a key developmental item in reducing the cost and increasing the efficiency of the IGCC concept. The TRF is a unique facility with high-temperature, high-pressure, and multiple reactant gas composition capability.more » The TRF can be configured for reacting a single flow pass of gas and solids using a variety of gases. The gas input system allows six different gas inputs to be mixed and heated before entering the reaction zones. Current configurations allow the use of air, carbon dioxide, carbon monoxide, hydrogen, hydrogen sulfide, methane, nitrogen, oxygen, steam, or any mixture of these gases. Construction plans include the addition of a coal gas input line. This line will bring hot coal gas from the existing Fluidized-Bed Gasifier (FBG) via the Modular Gas Cleanup Rig (MGCR) after filtering out particulates with ceramic candle filters. Solids can be fed either by a rotary pocket feeder or a screw feeder. Particle sizes may range from 70 to 150 micrometers. Both feeders have a hopper that can hold enough solid for fairly lengthy tests at the higher feed rates, thus eliminating the need for lockhopper transfers during operation.« less

Thermally Simulated 32kW Direct-Drive Gas-Cooled Reactor: Design, Assembly, and Test

NASA Astrophysics Data System (ADS)

Godfroy, Thomas J.; Kapernick, Richard J.; Bragg-Sitton, Shannon M.

2004-02-01

One of the power systems under consideration for nuclear electric propulsion is a direct-drive gas-cooled reactor coupled to a Brayton cycle. In this system, power is transferred from the reactor to the Brayton system via a circulated closed loop gas. To allow early utilization, system designs must be relatively simple, easy to fabricate, and easy to test using non-nuclear heaters to closely mimic heat from fission. This combination of attributes will allow pre-prototypic systems to be designed, fabricated, and tested quickly and affordably. The ability to build and test units is key to the success of a nuclear program, especially if an early flight is desired. The ability to perform very realistic non-nuclear testing increases the success probability of the system. In addition, the technologies required by a concept will substantially impact the cost, time, and resources required to develop a successful space reactor power system. This paper describes design features, assembly, and test matrix for the testing of a thermally simulated 32kW direct-drive gas-cooled reactor in the Early Flight Fission - Test Facility (EFF-TF) at Marshall Space Flight Center. The reactor design and test matrix are provided by Los Alamos National Laboratories.

Advanced Nuclear Fuel Cycle Transitions: Optimization, Modeling Choices, and Disruptions

NASA Astrophysics Data System (ADS)

Carlsen, Robert W.

Many nuclear fuel cycle simulators have evolved over time to help understan the nuclear industry/ecosystem at a macroscopic level. Cyclus is one of th first fuel cycle simulators to accommodate larger-scale analysis with it liberal open-source licensing and first-class Linux support. Cyclus also ha features that uniquely enable investigating the effects of modeling choices o fuel cycle simulators and scenarios. This work is divided into thre experiments focusing on optimization, effects of modeling choices, and fue cycle uncertainty. Effective optimization techniques are developed for automatically determinin desirable facility deployment schedules with Cyclus. A novel method fo mapping optimization variables to deployment schedules is developed. Thi allows relationships between reactor types and scenario constraints to b represented implicitly in the variable definitions enabling the usage o optimizers lacking constraint support. It also prevents wasting computationa resources evaluating infeasible deployment schedules. Deployed power capacit over time and deployment of non-reactor facilities are also included a optimization variables There are many fuel cycle simulators built with different combinations o modeling choices. Comparing results between them is often difficult. Cyclus flexibility allows comparing effects of many such modeling choices. Reacto refueling cycle synchronization and inter-facility competition among othe effects are compared in four cases each using combinations of fleet of individually modeled reactors with 1-month or 3-month time steps. There are noticeable differences in results for the different cases. The larges differences occur during periods of constrained reactor fuel availability This and similar work can help improve the quality of fuel cycle analysi generally There is significant uncertainty associated deploying new nuclear technologie such as time-frames for technology availability and the cost of buildin advanced reactors

Simplified pulse reactor for real-time long-term in vitro testing of biological heart valves.

PubMed

Schleicher, Martina; Sammler, Günther; Schmauder, Michael; Fritze, Olaf; Huber, Agnes J; Schenke-Layland, Katja; Ditze, Günter; Stock, Ulrich A

2010-05-01

Long-term function of biological heart valve prostheses (BHV) is limited by structural deterioration leading to failure with associated arterial hypertension. The objective of this work was development of an easy to handle real-time pulse reactor for evaluation of biological and tissue engineered heart valves under different pressures and long-term conditions. The pulse reactor was made of medical grade materials for placement in a 37 degrees C incubator. Heart valves were mounted in a housing disc moving horizontally in culture medium within a cylindrical culture reservoir. The microprocessor-controlled system was driven by pressure resulting in a cardiac-like cycle enabling competent opening and closing of the leaflets with adjustable pulse rates and pressures between 0.25 to 2 Hz and up to 180/80 mmHg, respectively. A custom-made imaging system with an integrated high-speed camera and image processing software allow calculation of effective orifice areas during cardiac cycle. This simple pulse reactor design allows reproducible generation of patient-like pressure conditions and data collection during long-term experiments.

Army gas-cooled reactor systems program. Preliminary design report off-normal scram system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bushnell, W.H.; Malmstrom, S.A.

1965-06-01

The maximum allowable ML-1 fuel element cladding (hot spot) temperature is established by ANTS 201 at 1750/sup 0/F. The existing ML-1 design makes no provision for automatic scram when this limit is reached. Operating experience has indicated a requirement for such an automatic system during plant startup and a revised hot spot envelope (generated during conceptual design of the scram system) established the desirability of extending this protection to operation at full power conditions. It was also determined that the scram system should include circuitry to initiate an automatic scram if reactor ..delta..T exceeded 450/sup 0/F (the limit established inmore » ANTS 201) and if reactor power exceeded 6 kw(t) without coolant flow in the main loop. The preliminary design of the scram system (designated off-normal scram system) which will provide the required protection is described.« less

Nuclear reactor cooling system decontamination reagent regeneration

DOEpatents

Anstine, Larry D.; James, Dean B.; Melaika, Edward A.; Peterson, Jr., John P.

1985-01-01

An improved method for decontaminating the coolant system of water-cooled nuclear power reactors and for regenerating the decontamination solution. A small amount of one or more weak-acid organic complexing agents is added to the reactor coolant, and the pH is adjusted to form a decontamination solution which is circulated throughout the coolant system to dissolve metal oxides from the interior surfaces and complex the resulting metal ions and radionuclide ions. The coolant containing the complexed metal ions and radionuclide ions is passed through a strong-base anion exchange resin bed which has been presaturated with a solution containing the complexing agents in the same ratio and having the same pH as the decontamination solution. As the decontamination solution passes through the resin bed, metal-complexed anions are exchanged for the metal-ion-free anions on the bed, while metal-ion-free anions in the solution pass through the bed, thus removing the metal ions and regenerating the decontamination solution.

Thermal margin protection system for a nuclear reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Musick, C.R.

1974-02-12

A thermal margin protection system for a nuclear reactor is described where the coolant flow flow trip point and the calculated thermal margin trip point are switched simultaneously and the thermal limit locus is made more restrictive as the allowable flow rate is decreased. The invention is characterized by calculation of the thermal limit Locus in response to applied signals which accurately represent reactor cold leg temperature and core power; cold leg temperature being corrected for stratification before being utilized and reactor power signals commensurate with power as a function of measured neutron flux and thermal energy added to themore » coolant being auctioneered to select the more conservative measure of power. The invention further comprises the compensation of the selected core power signal for the effects of core radial peaking factor under maximum coolant flow conditions. (Official Oazette)« less

VISION User Guide - VISION (Verifiable Fuel Cycle Simulation) Model

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jacob J. Jacobson; Robert F. Jeffers; Gretchen E. Matthern

2009-08-01

The purpose of this document is to provide a guide for using the current version of the Verifiable Fuel Cycle Simulation (VISION) model. This is a complex model with many parameters; the user is strongly encouraged to read this user guide before attempting to run the model. This model is an R&D work in progress and may contain errors and omissions. It is based upon numerous assumptions. This model is intended to assist in evaluating “what if” scenarios and in comparing fuel, reactor, and fuel processing alternatives at a systems level for U.S. nuclear power. The model is not intendedmore » as a tool for process flow and design modeling of specific facilities nor for tracking individual units of fuel or other material through the system. The model is intended to examine the interactions among the components of a fuel system as a function of time varying system parameters; this model represents a dynamic rather than steady-state approximation of the nuclear fuel system. VISION models the nuclear cycle at the system level, not individual facilities, e.g., “reactor types” not individual reactors and “separation types” not individual separation plants. Natural uranium can be enriched, which produces enriched uranium, which goes into fuel fabrication, and depleted uranium (DU), which goes into storage. Fuel is transformed (transmuted) in reactors and then goes into a storage buffer. Used fuel can be pulled from storage into either separation of disposal. If sent to separations, fuel is transformed (partitioned) into fuel products, recovered uranium, and various categories of waste. Recycled material is stored until used by its assigned reactor type. Note that recovered uranium is itself often partitioned: some RU flows with recycled transuranic elements, some flows with wastes, and the rest is designated RU. RU comes out of storage if needed to correct the U/TRU ratio in new recycled fuel. Neither RU nor DU are designated as wastes. VISION is comprised of

Ongoing Development of a Series Bosch Reactor System

NASA Technical Reports Server (NTRS)

Abney, Morgan; Mansell, Matt; DuMez, Sam; Thomas, John; Cooper, Charlie; Long, David

2013-01-01

Future manned missions to deep space or planetary surfaces will undoubtedly require highly robust, efficient, and regenerable life support systems that require minimal consumables. To meet this requirement, NASA continues to explore a Bosch-based carbon dioxide reduction system to recover oxygen from CO2. In order to improve the equivalent system mass of Bosch systems, we seek to design and test a "Series Bosch" system in which two reactors in series are optimized for the two steps of the reaction, as well as to explore the use of in situ materials as carbon deposition catalysts. Here we report recent developments in this effort including assembly and initial testing of a Reverse Water-Gas Shift reactor (RWGSr) and initial testing of two gas separation membranes. The RWGSr was sized to reduce CO2 produced by a crew of four to carbon monoxide as the first stage in a Series Bosch system. The gas separation membranes, necessary to recycle unreacted hydrogen and CO2, were similarly sized. Additionally, we report results of preliminary experiments designed to determine the catalytic properties of Martian and Lunar regolith simulant for the carbon deposition step.

Ongoing Development of a Series Bosch Reactor System

NASA Technical Reports Server (NTRS)

Abney, Morgan B; Mansell, J. Matthew; Stanley, Christine; Edmunson, Jennifer; DuMez, Samuel J.; Chen, Kevin

2013-01-01

Future manned missions to deep space or planetary surfaces will undoubtedly incorporate highly robust, efficient, and regenerable life support systems that require minimal consumables. To meet this requirement, NASA continues to explore a Bosch-based carbon dioxide reduction system to recover oxygen from CO2. In order to improve the equivalent system mass of Bosch systems, we seek to design and test a "Series Bosch" system in which two reactors in series are optimized for the two steps of the reaction, as well as to explore the use of in situ materials as carbon deposition catalysts. Here we report recent developments in this effort including assembly and initial testing of a Reverse Water-Gas Shift reactor (RWGSr) and initial testing of two gas separation membranes. The RWGSr was sized to reduce CO2 produced by a crew of four to carbon monoxide as the first stage in a Series Bosch system. The gas separation membranes, necessary to recycle unreacted hydrogen and CO2, were similarly sized. Additionally, we report results of preliminary experiments designed to determine the catalytic properties of Martian regolith simulant for the carbon formation step.

Testimony of Fred R. Mynatt before the Energy Research and Development Subcommittee of the Committee on Science, Space, and Technology, US House of Representatives. [Advanced fuel technology, gas-cooled reactor technology, and liquid metal-cooled reactor technology programs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mynatt, F.R.

1987-03-18

This report provides a description of the statements submitted for the record to the committee on Science, Space, and Technology of the United States House of Representatives. These statements describe three principal areas of activity of the Advanced Reactor Technology Program of the Department of Energy (DOE). These areas are advanced fuel cycle technology, modular high-temperature gas-cooled reactor technology, and liquid metal-cooled reactor. The areas of automated reactor control systems, robotics, materials and structural design shielding and international cooperation were included in these statements describing the Oak Ridge National Laboratory's efforts in these areas. (FI)

Status of French reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ballagny, A.

1997-08-01

The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm{sup 3}. The OSIRIS reactor has already been converted to LEU. It will use U{sub 3}Si{sub 2} as soon as its present stock of UO{sub 2} fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (exceptmore » if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU.« less

Oxygen transport membrane reactor based method and system for generating electric power

DOEpatents

Kelly, Sean M.; Chakravarti, Shrikar; Li, Juan

2017-02-07

A carbon capture enabled system and method for generating electric power and/or fuel from methane containing sources using oxygen transport membranes by first converting the methane containing feed gas into a high pressure synthesis gas. Then, in one configuration the synthesis gas is combusted in oxy-combustion mode in oxygen transport membranes based boiler reactor operating at a pressure at least twice that of ambient pressure and the heat generated heats steam in thermally coupled steam generation tubes within the boiler reactor; the steam is expanded in steam turbine to generate power; and the carbon dioxide rich effluent leaving the boiler reactor is processed to isolate carbon. In another configuration the synthesis gas is further treated in a gas conditioning system configured for carbon capture in a pre-combustion mode using water gas shift reactors and acid gas removal units to produce hydrogen or hydrogen-rich fuel gas that fuels an integrated gas turbine and steam turbine system to generate power. The disclosed method and system can also be adapted to integrate with coal gasification systems to produce power from both coal and methane containing sources with greater than 90% carbon isolation.

Progress in space nuclear reactor power systems technology development - The SP-100 program

NASA Technical Reports Server (NTRS)

Davis, H. S.

1984-01-01

Activities related to the development of high-temperature compact nuclear reactors for space applications had reached a comparatively high level in the U.S. during the mid-1950s and 1960s, although only one U.S. nuclear reactor-powered spacecraft was actually launched. After 1973, very little effort was devoted to space nuclear reactor and propulsion systems. In February 1983, significant activities toward the development of the technology for space nuclear reactor power systems were resumed with the SP-100 Program. Specific SP-100 Program objectives are partly related to the determination of the potential performance limits for space nuclear power systems in 100-kWe and 1- to 100-MW electrical classes. Attention is given to potential missions and applications, regimes of possible space power applicability, safety considerations, conceptual system designs, the establishment of technical feasibility, nuclear technology, materials technology, and prospects for the future.

Evaluation of Erosion of the Dummy “EE” Plate 19 in YA Type ATR Fuel Element During Reactor PALM Cycles

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brower, Jeffrey O.; Glazoff, Michael V.; Eiden, Thomas J.

Advanced Test Reactor (ATR) Cycle 153B-1 was a 14-day, high-power, powered axial locator mechanism (PALM) operating cycle that completed on April 12, 2013. Cycle 153B-1 was a typical operating cycle for the ATR, and did not result in any unusual plant transients. ATR was started up and shut down as scheduled. The PALM drive physically moves the selected experiments into and out of the core to simulate reactor startup and heat up, and shutdown and cooldown transients, while the reactor remains in steady-state conditions. However, after the cycle was over, when the fuel elements were removed from the core andmore » inspected, several thousand flow-assisted erosion pits and “horseshoeing” defects were readily observed on the surface of the several YA-type fuel elements (these are aluminum “dummy” plates that contain no fuel). In order to understand these erosion phenomena, a thermal-hydraulic model of coolant channel 20 on a YA-M fuel element was generated. The boundaries of the model were the aluminum EE plate of a YA-M fuel element and a beryllium reflector block with 13 horizontal saw cuts which represented regions of zero flow. The heat generated in fuel plates 1 through 18 was modeled to be passing through the aluminum EE plate. The coolant channel 20 width was set at 0.058 in. (58 mils). It was established that the horizontal saw cuts had a significant effect on the temperature of the coolant. The flow, which was expected to vary linearly with gradual heating of the coolant as it passed through the channel, was extremely turbulent. The temperature rise, which was expected to be a smooth “S” curve, was represented by a series temperature rise “humps,” which occurred at each horizontal saw cut in the beryllium reflector block. Each of the 13 saw cuts had a chamfered edge which resulted in the coolant flow being re-directed as a jet across the coolant channel into the surface of the EE plate, which explained the temperature rise and the observed

Update on Small Modular Reactors Dynamic System Modeling Tool: Web Application

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hale, Richard Edward; Cetiner, Sacit M.; Fugate, David L.

Previous reports focused on the development of component and system models as well as end-to-end system models using Modelica and Dymola for two advanced reactor architectures: (1) Advanced Liquid Metal Reactor and (2) fluoride high-temperature reactor (FHR). The focus of this report is the release of the first beta version of the web-based application for model use and collaboration, as well as an update on the FHR model. The web-based application allows novice users to configure end-to-end system models from preconfigured choices to investigate the instrumentation and controls implications of these designs and allows for the collaborative development of individualmore » component models that can be benchmarked against test systems for potential inclusion in the model library. A description of this application is provided along with examples of its use and a listing and discussion of all the models that currently exist in the library.« less

Combustion flame-plasma hybrid reactor systems, and chemical reactant sources

DOEpatents

Kong, Peter C

2013-11-26

Combustion flame-plasma hybrid reactor systems, chemical reactant sources, and related methods are disclosed. In one embodiment, a combustion flame-plasma hybrid reactor system comprising a reaction chamber, a combustion torch positioned to direct a flame into the reaction chamber, and one or more reactant feed assemblies configured to electrically energize at least one electrically conductive solid reactant structure to form a plasma and feed each electrically conductive solid reactant structure into the plasma to form at least one product is disclosed. In an additional embodiment, a chemical reactant source for a combustion flame-plasma hybrid reactor comprising an elongated electrically conductive reactant structure consisting essentially of at least one chemical reactant is disclosed. In further embodiments, methods of forming a chemical reactant source and methods of chemically converting at least one reactant into at least one product are disclosed.

Enterprise and system of systems capability development life-cycle processes.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Beck, David Franklin

2014-08-01

This report and set of appendices are a collection of memoranda originally drafted circa 2007-2009 for the purpose of describing and detailing a models-based systems engineering approach for satisfying enterprise and system-of-systems life cycle process requirements. At the time there was interest and support to move from Capability Maturity Model Integration (CMMI) Level One (ad hoc processes) to Level Three. The main thrust of the material presents a rational exposâe of a structured enterprise development life cycle that uses the scientific method as a framework, with further rigor added from adapting relevant portions of standard systems engineering processes. While themore » approach described invokes application of the Department of Defense Architectural Framework (DoDAF), it is suitable for use with other architectural description frameworks.« less

Power Peaking Effect of OTTO Fuel Scheme Pebble Bed Reactor

NASA Astrophysics Data System (ADS)

Setiadipura, T.; Suwoto; Zuhair; Bakhri, S.; Sunaryo, G. R.

2018-02-01

Pebble Bed Reactor (PBR) type of Hight Temperature Gas-cooled Reactor (HTGR) is a very interesting nuclear reactor design to fulfill the growing electricity and heat demand with a superior passive safety features. Effort to introduce the PBR design to the market can be strengthen by simplifying its system with the Once-through-then-out (OTTO) cycle PBR in which the pebble fuel only pass the core once. Important challenge in the OTTO fuel scheme is the power peaking effect which limit the maximum nominal power or burnup of the design. Parametric survey is perform in this study to investigate the contribution of different design parameters to power peaking effect of OTTO cycle PBR. PEBBED code is utilized in this study to perform the equilibrium PBR core analysis for different design parameter and fuel scheme. The parameters include its core diameter, height-per-diameter (H/D), power density, and core nominal power. Results of this study show that diameter and H/D effectsare stronger compare to the power density and nominal core power. Results of this study might become an importance guidance for design optimization of OTTO fuel scheme PBR.

Method and apparatus for enhancing reactor air-cooling system performance

DOEpatents

Hunsbedt, A.

1996-03-12

An enhanced decay heat removal system is disclosed for removing heat from the inert gas-filled gap space between the reactor vessel and the containment vessel of a liquid metal-cooled nuclear reactor. Multiple cooling ducts in flow communication with the inert gas-filled gap space are incorporated to provide multiple flow paths for the inert gas to circulate to heat exchangers which remove heat from the inert gas, thereby introducing natural convection flows in the inert gas. The inert gas in turn absorbs heat directly from the reactor vessel by natural convection heat transfer. 6 figs.

Passive load follow analysis of the STAR-LM and STAR-H2 systems

NASA Astrophysics Data System (ADS)

Moisseytsev, Anton

A steady-state model for the calculation of temperature and pressure distributions, and heat and work balance for the STAR-LM and the STAR-H2 systems was developed. The STAR-LM system is designed for electricity production and consists of the lead cooled reactor on natural circulation and the supercritical carbon dioxide Brayton cycle. The STAR-H2 system uses the same reactor which is coupled to the hydrogen production plant, the Brayton cycle, and the water desalination plant. The Brayton cycle produces electricity for the on-site needs. Realistic modules for each system component were developed. The model also performs design calculations for the turbine and compressors for the CO2 Brayton cycle. The model was used to optimize the performance of the entire system as well as every system component. The size of each component was calculated. For the 400 MWt reactor power the STAR-LM produces 174.4 MWe (44% efficiency) and the STAR-H2 system produces 7450 kg H2/hr. The steady state model was used to conduct quasi-static passive load follow analysis. The control strategy was developed for each system; no control action on the reactor is required. As a main safety criterion, the peak cladding temperature is used. It was demonstrated that this temperature remains below the safety limit during both normal operation and load follow.

SYSTEM FOR UNLOADING REACTORS

DOEpatents

Rand, A.C. Jr.

1961-05-01

An unloading device for individual vertical fuel channels in a nuclear reactor is shown. The channels are arranged in parallel rows and underneath each is a separate supporting block on which the fuel in the channel rests. The blocks are raounted in contiguous rows on an array of parallel pairs of tracks over the bottom of the reactor. Oblong hollows in the blocks form a continuous passageway through the middle of the row of blocks on each pair of tracks. At the end of each passageway is a horizontal grappling rod with a T- or L extension at the end next to the reactor of a length to permit it to pass through the oblong passageway in one position, but when rotated ninety degrees the head will strike one of the longer sides of the oblong hollow of one of the blocks. The grappling rod is actuated by a controllable reciprocating and rotating device which extends it beyond any individual block desired, rotates it and retracts it far enough to permit the fuel in the vertical channel above the block to fall into a handling tank below the reactor.

Coupled reactor kinetics and heat transfer model for heat pipe cooled reactors

NASA Astrophysics Data System (ADS)

Wright, Steven A.; Houts, Michael

2001-02-01

Heat pipes are often proposed as cooling system components for small fission reactors. SAFE-300 and STAR-C are two reactor concepts that use heat pipes as an integral part of the cooling system. Heat pipes have been used in reactors to cool components within radiation tests (Deverall, 1973); however, no reactor has been built or tested that uses heat pipes solely as the primary cooling system. Heat pipe cooled reactors will likely require the development of a test reactor to determine the main differences in operational behavior from forced cooled reactors. The purpose of this paper is to describe the results of a systems code capable of modeling the coupling between the reactor kinetics and heat pipe controlled heat transport. Heat transport in heat pipe reactors is complex and highly system dependent. Nevertheless, in general terms it relies on heat flowing from the fuel pins through the heat pipe, to the heat exchanger, and then ultimately into the power conversion system and heat sink. A system model is described that is capable of modeling coupled reactor kinetics phenomena, heat transfer dynamics within the fuel pins, and the transient behavior of heat pipes (including the melting of the working fluid). This paper focuses primarily on the coupling effects caused by reactor feedback and compares the observations with forced cooled reactors. A number of reactor startup transients have been modeled, and issues such as power peaking, and power-to-flow mismatches, and loading transients were examined, including the possibility of heat flow from the heat exchanger back into the reactor. This system model is envisioned as a tool to be used for screening various heat pipe cooled reactor concepts, for designing and developing test facility requirements, for use in safety evaluations, and for developing test criteria for in-pile and out-of-pile test facilities. .

A small, 1400 deg Kelvin, reactor for Brayton space power systems

NASA Technical Reports Server (NTRS)

Lantz, E.; Mayo, W.

1972-01-01

A preliminary cost estimate for a small reactor in Brayton space power systems with (u-233)n or (pu-239)n as the fuel in the T-111 fuel elements totaled to about four million dollars; considered is a 22.8 in. diameter reactor with 247 fuel elements.

Safety approach to the selection of design criteria for the CRBRP reactor refueling system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Meisl, C J; Berg, G E; Sharkey, N F

1979-01-01

The selection of safety design criteria for Liquid Metal Fast Breeder Reactor (LMFBR) refueling systems required the extrapolation of regulations and guidelines intended for Light Water Reactor refueling systems and was encumbered by the lack of benefit from a commercially licensed predecessor other than Fermi. The overall approach and underlying logic are described for developing safety design criteria for the reactor refueling system (RRS) of the Clinch River Breeder Reactor Plant (CRBRP). The complete selection process used to establish the criteria is presented, from the definition of safety functions to the finalization of safety design criteria in the appropriate documents.more » The process steps are illustrated by examples.« less

NEUTRONIC REACTOR SYSTEM

DOEpatents

Daniels, F.

1957-10-15

Gas-cooled solid-moderator type reactors wherein the fissionable fuel and moderator materials are each in the form of solid pebbles, or discrete particles, and are substantially homogeneously mixed in the proper proportion and placed within the core of the reactor are described. The shape of these discrete particles must be such that voids are present between them when mixed together. Helium enters the bottom of the core and passes through the voids between the fuel and moderator particles to absorb the heat generated by the chain reaction. The hot helium gas is drawn off the top of the core and may be passed through a heat exchanger to produce steam.

Post impact behavior of mobile reactor core containment systems

NASA Technical Reports Server (NTRS)

Puthoff, R. L.; Parker, W. G.; Vanbibber, L. E.

1972-01-01

The reactor core containment vessel temperatures after impact, and the design variables that affect the post impact survival of the system are analyzed. The heat transfer analysis includes conduction, radiation, and convection in addition to the core material heats of fusion and vaporization under partially burial conditions. Also, included is the fact that fission products vaporize and transport radially outward and condense outward and condense on cooler surfaces, resulting in a moving heat source. A computer program entitled Executive Subroutines for Afterheat Temperature Analysis (ESATA) was written to consider this complex heat transfer analysis. Seven cases were calculated of a reactor power system capable of delivering up to 300 MW of thermal power to a nuclear airplane.

Burning high-level TRU waste in fusion fission reactors

NASA Astrophysics Data System (ADS)

Shen, Yaosong

2016-09-01

Recently, the concept of actinide burning instead of a once-through fuel cycle for disposing spent nuclear fuel seems to get much more attention. A new method of burning high-level transuranic (TRU) waste combined with Thorium-Uranium (Th-U) fuel in the subcritical reactors driven by external fusion neutron sources is proposed in this paper. The thorium-based TRU fuel burns all of the long-lived actinides via a hard neutron spectrum while outputting power. A one-dimensional model of the reactor concept was built by means of the ONESN_BURN code with new data libraries. The numerical results included actinide radioactivity, biological hazard potential, and much higher burnup rate of high-level transuranic waste. The comparison of the fusion-fission reactor with the thermal reactor shows that the harder neutron spectrum is more efficient than the soft. The Th-U cycle produces less TRU, less radiotoxicity and fewer long-lived actinides. The Th-U cycle provides breeding of 233U with a long operation time (>20 years), hence significantly reducing the reactivity swing while improving safety and burnup.

Fossil fuel combined cycle power system

DOEpatents

Labinov, Solomon Davidovich; Armstrong, Timothy Robert; Judkins, Roddie Reagan

2006-10-10

A system for converting fuel energy to electricity includes a reformer for converting a higher molecular weight gas into at least one lower molecular weight gas, at least one turbine to produce electricity from expansion of at least one of the lower molecular weight gases, and at least one fuel cell. The system can further include at least one separation device for substantially dividing the lower molecular weight gases into at least two gas streams prior to the electrochemical oxidization step. A nuclear reactor can be used to supply at least a portion of the heat the required for the chemical conversion process.

Systems and methods for dismantling a nuclear reactor

DOEpatents

Heim, Robert R; Adams, Scott Ryan; Cole, Matthew Denver; Kirby, William E; Linnebur, Paul Damon

2014-10-28

Systems and methods for dismantling a nuclear reactor are described. In one aspect the system includes a remotely controlled heavy manipulator ("manipulator") operatively coupled to a support structure, and a control station in a non-contaminated portion of a workspace. The support structure provides the manipulator with top down access into a bioshield of a nuclear reactor. At least one computing device in the control station provides remote control to perform operations including: (a) dismantling, using the manipulator, a graphite moderator, concrete walls, and a ceiling of the bioshield, the manipulator being provided with automated access to all internal portions of the bioshield; (b) loading, using the manipulator, contaminated graphite blocks from the graphite core and other components from the bioshield into one or more waste containers; and (c) dispersing, using the manipulator, dust suppression and contamination fixing spray to contaminated matter.

Autonomous Control Capabilities for Space Reactor Power Systems

NASA Astrophysics Data System (ADS)

Wood, Richard T.; Neal, John S.; Brittain, C. Ray; Mullens, James A.

2004-02-01

The National Aeronautics and Space Administration's (NASA's) Project Prometheus, the Nuclear Systems Program, is investigating a possible Jupiter Icy Moons Orbiter (JIMO) mission, which would conduct in-depth studies of three of the moons of Jupiter by using a space reactor power system (SRPS) to provide energy for propulsion and spacecraft power for more than a decade. Terrestrial nuclear power plants rely upon varying degrees of direct human control and interaction for operations and maintenance over a forty to sixty year lifetime. In contrast, an SRPS is intended to provide continuous, remote, unattended operation for up to fifteen years with no maintenance. Uncertainties, rare events, degradation, and communications delays with Earth are challenges that SRPS control must accommodate. Autonomous control is needed to address these challenges and optimize the reactor control design. In this paper, we describe an autonomous control concept for generic SRPS designs. The formulation of an autonomous control concept, which includes identification of high-level functional requirements and generation of a research and development plan for enabling technologies, is among the technical activities that are being conducted under the U.S. Department of Energy's Space Reactor Technology Program in support of the NASA's Project Prometheus. The findings from this program are intended to contribute to the successful realization of the JIMO mission.

An numerical analysis of high-temperature helium reactor power plant for co-production of hydrogen and electricity

NASA Astrophysics Data System (ADS)

Dudek, M.; Podsadna, J.; Jaszczur, M.

2016-09-01

In the present work, the feasibility of using a high temperature gas cooled nuclear reactor (HTR) for electricity generation and hydrogen production are analysed. The HTR is combined with a steam and a gas turbine, as well as with the system for heat delivery for medium temperature hydrogen production. Industrial-scale hydrogen production using copper-chlorine (Cu-Cl) thermochemical cycle is considered and compared with high temperature electrolysis. Presented cycle shows a very promising route for continuous, efficient, large-scale and environmentally benign hydrogen production without CO2 emissions. The results show that the integration of a high temperature helium reactor, with a combined cycle for electric power generation and hydrogen production, may reach very high efficiency and could possibly lead to a significant decrease of hydrogen production costs.

Proceedings of a Symposium on Advanced Compact Reactor Systems

NASA Technical Reports Server (NTRS)

1983-01-01

Reactor system technologies suitable for a variety of aerospace and terrestrial applications are considered. Technologies, safety and regulatory considerations, potential applications, and research and development opportunities are covered.

Technology for Bayton-cycle powerplants using solar and nuclear energy

NASA Technical Reports Server (NTRS)

English, R. E.

1986-01-01

Brayton cycle gas turbines have the potential to use either solar heat or nuclear reactors for generating from tens of kilowatts to tens of megawatts of power in space, all this from a single technology for the power generating system. Their development for solar energy dynamic power generation for the space station could be the first step in an evolution of such powerplants for a very wide range of applications. At the low power level of only 10 kWe, a power generating system has already demonstrated overall efficiency of 0.29 and operated 38 000 hr. Tests of improved components show that these components would raise that efficiency to 0.32, a value twice that demonstrated by any alternate concept. Because of this high efficiency, solar Brayton cycle power generators offer the potential to increase power per unit of solar collector area to levels exceeding four times that from photovoltaic powerplants using present technology for silicon solar cells. The technologies for solar mirrors and heat receivers are reviewed and assessed. This Brayton technology for solar powerplants is equally suitable for use with the nuclear reactors. The available long time creep data on the tantalum alloy ASTAR-811C show that such Brayton cycles can evolve to cycle peak temperatures of 1500 K (2240 F). And this same technology can be extended to generate 10 to 100 MW in space by exploiting existing technology for terrestrial gas turbines in the fields of both aircraft propulsion and stationary power generation.

Continuous enzymatic hydrolysis of lignocellulosic biomass in a membrane-reactor system: Continuous enzymatic hydrolysis of lignocellulosic biomass in a membrane-reactor system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stickel, Jonathan J.; Adhikari, Birendra; Sievers, David A.

Converting abundant lignocellulosic biomass to sugars as fungible precursors to fuels and chemicals has the potential to diversify the supply chain for those products, but further process improvements are needed to achieve economic viability. In the current work, process intensification of the key enzymatic hydrolysis unit operation is demonstrated by means of a membrane reactor system that was operated continuously. Lignocellulosic biomass (pretreated corn stover) and buffered enzyme solution were fed to a continuously stirred-tank reactor, and clarified sugar solution was withdrawn via a commercial tubular ultrafiltration membrane. The membrane permeance decline and membrane cleaning efficacy were studied and didmore » not vary significantly when increasing fraction insoluble solids (FIS) from 2.5% to 5%. Continuous enzymatic hydrolysis was successfully operated for more than 80 h. A model for the reactor system was able to predict dynamic behavior that was in reasonable agreement with experimental results. The modeled technical performance of anticipated commercial batch and continuous enzymatic hydrolysis processes were compared and showed that continuous operation would provide at least twice the volumetric productivity for the conditions studied. Further improvements are anticipated by better membrane selection and by increasing FIS.« less

Continuous enzymatic hydrolysis of lignocellulosic biomass in a membrane-reactor system: Continuous enzymatic hydrolysis of lignocellulosic biomass in a membrane-reactor system

DOE PAGES

Stickel, Jonathan J.; Adhikari, Birendra; Sievers, David A.; ...

2018-02-21

Converting abundant lignocellulosic biomass to sugars as fungible precursors to fuels and chemicals has the potential to diversify the supply chain for those products, but further process improvements are needed to achieve economic viability. In the current work, process intensification of the key enzymatic hydrolysis unit operation is demonstrated by means of a membrane reactor system that was operated continuously. Lignocellulosic biomass (pretreated corn stover) and buffered enzyme solution were fed to a continuously stirred-tank reactor, and clarified sugar solution was withdrawn via a commercial tubular ultrafiltration membrane. The membrane permeance decline and membrane cleaning efficacy were studied and didmore » not vary significantly when increasing fraction insoluble solids (FIS) from 2.5% to 5%. Continuous enzymatic hydrolysis was successfully operated for more than 80 h. A model for the reactor system was able to predict dynamic behavior that was in reasonable agreement with experimental results. The modeled technical performance of anticipated commercial batch and continuous enzymatic hydrolysis processes were compared and showed that continuous operation would provide at least twice the volumetric productivity for the conditions studied. Further improvements are anticipated by better membrane selection and by increasing FIS.« less

Cycle analysis of MCFC/gas turbine system

NASA Astrophysics Data System (ADS)

Musa, Abdullatif; Alaktiwi, Abdulsalam; Talbi, Mosbah

2017-11-01

High temperature fuel cells such as the solid oxide fuel cell (SOFC) and the molten carbonate fuel cell (MCFC) are considered extremely suitable for electrical power plant application. The molten carbonate fuel cell (MCFC) performances is evaluated using validated model for the internally reformed (IR) fuel cell. This model is integrated in Aspen Plus™. Therefore, several MCFC/Gas Turbine systems are introduced and investigated. One of this a new cycle is called a heat recovery (HR) cycle. In the HR cycle, a regenerator is used to preheat water by outlet air compressor. So the waste heat of the outlet air compressor and the exhaust gases of turbine are recovered and used to produce steam. This steam is injected in the gas turbine, resulting in a high specific power and a high thermal efficiency. The cycles are simulated in order to evaluate and compare their performances. Moreover, the effects of an important parameters such as the ambient air temperature on the cycle performance are evaluated. The simulation results show that the HR cycle has high efficiency.

Automated power control system for reactor TRIGA PUSPATI

NASA Astrophysics Data System (ADS)

Ghazali, Anith Khairunnisa; Minhat, Mohd Sabri; Hassan, Mohd Khair

2017-01-01

Reactor TRIGA PUSPATI (RTP) Mark II type undergoes safe operation for more than 30 years and the only research reactor exists in Malaysia. The main safety feature of Instrumentation and Control (I&C) system design is such that any failure in the electronic, or its associated components, does not lead to an uncontrolled rate of reactivity. The existed controller using feedback approach to control the reactor power. This paper introduces proposed controllers such as Model Reference Adaptive Control (MRAC) and Proportional Integral Derivatives (PID) controller for the RTP simulation. In RTP, the most important considered parameter is the reactor power and act as nervous system. To design a controller for complex plant like RTP is quite difficult due to high cost and safety factors cause by the failure of the controller. Furthermore, to overcome these problems, a simulator can be used to replace functions the hardware and test could then be simulated using this simulator. In order to find the best controller, several controllers were proposed and the result will be analysed for study the performances of the controller. The output result will be used to find out the best RTP power controller using MATLAB/Simulink and gives result as close as the real RTP performances. Currently, the structures of RTP was design using MATLAB/Simulink tool that consist of fission chamber, controller, control rod position, height-to-worth of control rods and a RTP model. The controller will control the control rod position to make sure that the reactivity still under the limitation parameter. The results given from each controller will be analysed and validated through experiment data collected from RTP.

Impact of Reactor Operating Parameters on Cask Reactivity in BWR Burnup Credit

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ilas, Germina; Betzler, Benjamin R; Ade, Brian J

This paper discusses the effect of reactor operating parameters used in fuel depletion calculations on spent fuel cask reactivity, with relevance for boiling-water reactor (BWR) burnup credit (BUC) applications. Assessments that used generic BWR fuel assembly and spent fuel cask configurations are presented. The considered operating parameters, which were independently varied in the depletion simulations for the assembly, included fuel temperature, bypass water density, specific power, and operating history. Different operating history scenarios were considered for the assembly depletion to determine the effect of relative power distribution during the irradiation cycles, as well as the downtime between cycles. Depletion, decay,more » and criticality simulations were performed using computer codes and associated nuclear data within the SCALE code system. Results quantifying the dependence of cask reactivity on the assembly depletion parameters are presented herein.« less

Life-Cycle Evaluation of Domestic Energy Systems

NASA Astrophysics Data System (ADS)

Bando, Shigeru; Hihara, Eiji

Among the growing number of environmental issues, the global warming due to the increasing emission of greenhouse gases, such as carbon dioxide CO2, is the most serious one. In order to reduce CO2 emissions in energy use, it is necessary to reduce primary energy consumption, and to replace energy sources with alternatives that emit less CO2.One option of such ideas is to replace fossil gas for water heating with electricity generated by nuclear power, hydraulic power, and other methods with low CO2 emission. It is also important to use energy efficiently and to reduce waste heat. Co-generation system is one of the applications to be able to use waste heat from a generator as much as possible. The CO2 heat pump water heaters, the polymer electrolyte fuel cells, and the micro gas turbines have high potential for domestic energy systems. In the present study, the life-cycle cost, the life-cycle consumption of primary energy and the life-cycle emission of CO2 of these domestic energy systems are compare. The result shows that the CO2 heat pump water heaters have an ability to reduce CO2 emission by 10%, and the co-generation systems also have another ability to reduce primary energy consumption by 20%.

User Guide for VISION 3.4.7 (Verifiable Fuel Cycle Simulation) Model

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jacob J. Jacobson; Robert F. Jeffers; Gretchen E. Matthern

2011-07-01

The purpose of this document is to provide a guide for using the current version of the Verifiable Fuel Cycle Simulation (VISION) model. This is a complex model with many parameters and options; the user is strongly encouraged to read this user guide before attempting to run the model. This model is an R&D work in progress and may contain errors and omissions. It is based upon numerous assumptions. This model is intended to assist in evaluating 'what if' scenarios and in comparing fuel, reactor, and fuel processing alternatives at a systems level. The model is not intended as amore » tool for process flow and design modeling of specific facilities nor for tracking individual units of fuel or other material through the system. The model is intended to examine the interactions among the components of a fuel system as a function of time varying system parameters; this model represents a dynamic rather than steady-state approximation of the nuclear fuel system. VISION models the nuclear cycle at the system level, not individual facilities, e.g., 'reactor types' not individual reactors and 'separation types' not individual separation plants. Natural uranium can be enriched, which produces enriched uranium, which goes into fuel fabrication, and depleted uranium (DU), which goes into storage. Fuel is transformed (transmuted) in reactors and then goes into a storage buffer. Used fuel can be pulled from storage into either separation or disposal. If sent to separations, fuel is transformed (partitioned) into fuel products, recovered uranium, and various categories of waste. Recycled material is stored until used by its assigned reactor type. VISION is comprised of several Microsoft Excel input files, a Powersim Studio core, and several Microsoft Excel output files. All must be co-located in the same folder on a PC to function. You must use Powersim Studio 8 or better. We have tested VISION with the Studio 8 Expert, Executive, and Education versions. The Expert and

Life Cycle Cost of C-130E Weapon System.

DTIC Science & Technology

1977-07-01

AFHRL-TR-7746 A-,I K lR FORCE g•r ___ FORCELIFE CYCLE COST OF C-130E WEAPON SYSTEM H By U Frank D. Brown "Gary A. WalkerU David H. WilsonBooing Awsp...I REPORT DOCUMENTO ON PAGE BEFORE COMPLETING FORM I?~ AFHR R.77-46 ( ~ LIFE CYCLE COST OF:C413A~ WEAPON SYSTEMteim.e M- 29 June 116- 3 June...human resources cost data life cycle cost historical weapon system analysis (I-WSA) inateuial resources 20). ABSTRACT (Con tinue on reverse siLa Mf

Fuel cycle cost uncertainty from nuclear fuel cycle comparison

DOE Office of Scientific and Technical Information (OSTI.GOV)

Li, J.; McNelis, D.; Yim, M.S.

2013-07-01

This paper examined the uncertainty in fuel cycle cost (FCC) calculation by considering both model and parameter uncertainty. Four different fuel cycle options were compared in the analysis including the once-through cycle (OT), the DUPIC cycle, the MOX cycle and a closed fuel cycle with fast reactors (FR). The model uncertainty was addressed by using three different FCC modeling approaches with and without the time value of money consideration. The relative ratios of FCC in comparison to OT did not change much by using different modeling approaches. This observation was consistent with the results of the sensitivity study for themore » discount rate. Two different sets of data with uncertainty range of unit costs were used to address the parameter uncertainty of the FCC calculation. The sensitivity study showed that the dominating contributor to the total variance of FCC is the uranium price. In general, the FCC of OT was found to be the lowest followed by FR, MOX, and DUPIC. But depending on the uranium price, the FR cycle was found to have lower FCC over OT. The reprocessing cost was also found to have a major impact on FCC.« less

Training courses on neutron detection systems on the ISIS research reactor: on-site and through internet training

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lescop, B.; Badeau, G.; Ivanovic, S.

Today, ISIS research reactor is an essential tool for Education and Training programs organized by the National Institute for Nuclear Science and Technology (INSTN) from CEA. In the field of nuclear instrumentation, the INSTN offers both, theoretical courses and training courses on the use of neutron detection systems taking advantage of the ISIS research reactor for the supply of a wide range of neutron fluxes. This paper describes the content of the training carried out on the use of neutron detectors and detection systems, on-site or remote. The ISIS reactor is a 700 kW open core pool type reactor. Themore » facility is very flexible since neutron detectors can be inserted into the core or its vicinity, and be used at different levels of power according to the needs of the course. Neutron fluxes, typically ranging from 1 to 10{sup 12} n/cm{sup 2}.s, can be obtained for the characterisation of the neutron detectors and detection systems. For the monitoring of the neutron density at low level of power, the Instrumentation and Control (I and C) system of the reactor is equipped with two detection systems, named BN1 and BN2. Each way contains a fission chamber, type CFUL01, connected to an electronic system type SIREX.The system works in pulse mode and exhibits two outputs: the counting rate and the doubling time. For the high level of power, the I and C is equipped with two detection systems HN1 and HN2.Each way contain a boron ionization chamber (type CC52) connected to an electronics system type SIREX. The system works in current mode and has two outputs: the current and the doubling time. For each mode, the trainees can observe and measure the signal at the different stages of the electronic system, with an oscilloscope. They can understand the role of each component of the detection system: detector, cable and each electronic block. The limitation of the detection modes and their operating range can be established from the measured signal. The trainees can

High-speed thermal cycling system and method of use

DOEpatents

Hansen, A.D.A.; Jaklevic, J.M.

1996-04-16

A thermal cycling system and method of use are described. The thermal cycling system is based on the circulation of temperature-controlled water directly to the underside of thin-walled polycarbonate plates. The water flow is selected from a manifold fed by pumps from heated reservoirs. The plate wells are loaded with typically 15-20 microliters of reagent mix for the PCR process. Heat transfer through the thin polycarbonate is sufficiently rapid that the contents reach thermal equilibrium with the water in less than 15 seconds. Complete PCR amplification runs of 40 three-step cycles have been performed in as little as 14.5 minutes, with the results showing substantially enhanced specificity compared to conventional technology requiring run times in excess of 100 minutes. The plate clamping station is designed to be amenable to robotic loading and unloading of the system. It includes a heated lid, thus eliminating the need for mineral oil overlay of the reactants. The present system includes three or more plate holder stations, fed from common reservoirs but operating with independent switching cycles. The system can be modularly expanded. 13 figs.

High-speed thermal cycling system and method of use

DOEpatents

Hansen, Anthony D. A.; Jaklevic, Joseph M.

1996-01-01

A thermal cycling system and method of use are described. The thermal cycling system is based on the-circulation of temperature-controlled water directly to the underside of thin-walled polycarbonate microtiter plates. The water flow is selected from a manifold fed by pumps from heated reservoirs. The plate wells are loaded with typically 15-20 .mu.l of reagent mix for the PCR process. Heat transfer through the thin polycarbonate is sufficiently rapid that the contents reach thermal equilibrium with the water in less than 15 seconds. Complete PCR amplification runs of 40 three-step cycles have been performed in as little as 14.5 minutes, with the results showing substantially enhanced specificity compared to conventional technology requiring run times in excess of 100 minutes. The plate clamping station is designed to be amenable to robotic loading and unloading of the system. It includes a heated lid, thus eliminating the need for mineral oil overlay of the reactants. The present system includes three or more plate holder stations, fed from common reservoirs but operating with independent switching cycles. The system can be modularly expanded.

Design study of long-life PWR using thorium cycle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Subkhi, Moh. Nurul; Su'ud, Zaki; Waris, Abdul

2012-06-06

Design study of long-life Pressurized Water Reactor (PWR) using thorium cycle has been performed. Thorium cycle in general has higher conversion ratio in the thermal spectrum domain than uranium cycle. Cell calculation, Burn-up and multigroup diffusion calculation was performed by PIJ-CITATION-SRAC code using libraries based on JENDL 3.2. The neutronic analysis result of infinite cell calculation shows that {sup 231}Pa better than {sup 237}Np as burnable poisons in thorium fuel system. Thorium oxide system with 8%{sup 233}U enrichment and 7.6{approx} 8%{sup 231}Pa is the most suitable fuel for small-long life PWR core because it gives reactivity swing less than 1%{Delta}k/kmore » and longer burn up period (more than 20 year). By using this result, small long-life PWR core can be designed for long time operation with reduced excess reactivity as low as 0.53%{Delta}k/k and reduced power peaking during its operation.« less

System-Level Heat Transfer Analysis, Thermal- Mechanical Cyclic Stress Analysis, and Environmental Fatigue Modeling of a Two-Loop Pressurized Water Reactor. A Preliminary Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mohanty, Subhasish; Soppet, William; Majumdar, Saurin

This report provides an update on an assessment of environmentally assisted fatigue for light water reactor components under extended service conditions. This report is a deliverable in April 2015 under the work package for environmentally assisted fatigue under DOE's Light Water Reactor Sustainability program. In this report, updates are discussed related to a system level preliminary finite element model of a two-loop pressurized water reactor (PWR). Based on this model, system-level heat transfer analysis and subsequent thermal-mechanical stress analysis were performed for typical design-basis thermal-mechanical fatigue cycles. The in-air fatigue lives of components, such as the hot and cold legs,more » were estimated on the basis of stress analysis results, ASME in-air fatigue life estimation criteria, and fatigue design curves. Furthermore, environmental correction factors and associated PWR environment fatigue lives for the hot and cold legs were estimated by using estimated stress and strain histories and the approach described in NUREG-6909. The discussed models and results are very preliminary. Further advancement of the discussed model is required for more accurate life prediction of reactor components. This report only presents the work related to finite element modelling activities. However, in between multiple tensile and fatigue tests were conducted. The related experimental results will be presented in the year-end report.« less

Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jennifer Lyons; Wade R. Marcum; Mark D. DeHart

2014-01-01

The Advanced Test Reactor (ATR), under the Reduced Enrichment for Research and Test Reactors (RERTR) Program and the Global Threat Reduction Initiative (GTRI), is conducting feasibility studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low enriched uranium (LEU) composition. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary analysis of TRIGA® fuel within the current ATR fuel envelopes and compares it to the functional requirements delineated by themore » Naval Reactors Program, which includes: greater than 4.8E+14 fissions/s/g of 235U, a fast to thermal neutron flux ratio that is less than 5% deviation of its current value, a constant cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other parameters outside those put forth by the Naval Reactors Program which are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time.« less

Evaluation of Corrosion of the Dummy “EE” Plate 19 in YA Type ATR Fuel Element During Reactor PALM Cycles

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brower, Jeffrey Owen; Glazoff, Michael Vasily; Eiden, Thomas John

Advanced Test Reactor (ATR) Cycle 153B-1 was a 14-day, high-power, powered axial locator mechanism (PALM) operating cycle that completed on April 12, 2013. Cycle 153B-1 was a typical operating cycle for the ATR and did not result in any unusual plant transients. ATR was started up and shut down as scheduled. The PALM drive physically moves the selected experiments into and out of the core to simulate reactor startup and heat up, and shutdown and cooldown transients, while the reactor remains in steady state conditions. However, after the cycle was over, several thousand of the flow-assisted corrosion pits and “horseshoeing”more » defects were readily observable on the surface of the several YA-type fuel elements (these are “dummy” plates that contain no fuel). In order understand these corrosion phenomena a thermal-hydraulic model of coolant channel 20 on a YA-M fuel element was generated. The boundaries of the model were the aluminum EE plate of a YA-M fuel element and a beryllium reflector block with 13 horizontal saw cuts which represented regions of zero flow. The heat generated in fuel plates 1 through 18 was modeled to be passing through the aluminum EE plate. The coolant channel 20 width was set at 0.058 in. (58 mils). It was established that the horizontal saw cuts had a significant effect on the temperature of the coolant. The flow, which was expected to vary linearly with gradual heating of the coolant as it passed through the channel, was extremely turbulent. The temperature rise, which was expected to be a smooth “S” curve, was represented by a series temperature rise “humps,” which occurred at each horizontal saw cut in the beryllium reflector block. Each of the 13 saw cuts had a chamfered edge which resulted in the coolant flow being re-directed as a jet across the coolant channel into the surface of the EE plate, which explained the temperature rise and the observed sscalloping and possibly pitting degradation on the YA-M fuel

System analyses on advanced nuclear fuel cycle and waste management

NASA Astrophysics Data System (ADS)

Cheon, Myeongguk

To evaluate the impacts of accelerator-driven transmutation of waste (ATW) fuel cycle on a geological repository, two mathematical models are developed: a reactor system analysis model and a high-level waste (HLW) conditioning model. With the former, fission products and residual trans-uranium (TRU) contained in HLW generated from a reference ATW plant operations are quantified and the reduction of TRU inventory included in commercial spent-nuclear fuel (CSNF) is evaluated. With the latter, an optimized waste loading and composition in solidification of HLW are determined and the volume reduction of waste packages associated with CSNF is evaluated. WACOM, a reactor system analysis code developed in this study for burnup calculation, is validated by ORIGEN2.1 and MCNP. WACOM is used to perform multicycle analysis for the reference lead-bismuth eutectic (LBE) cooled transmuter. By applying the results of this analysis to the reference ATW deployment scenario considered in the ATW roadmap, the HLW generated from the ATW fuel cycle is quantified and the reduction of TRU inventory contained in CSNF is evaluated. A linear programming (LP) model has been developed for determination of an optimized waste loading and composition in solidification of HLW. The model has been applied to a US-defense HLW. The optimum waste loading evaluated by the LP model was compared with that estimated by the Defense Waste Processing Facility (DWPF) in the US and a good agreement was observed. The LP model was then applied to the volume reduction of waste packages associated with CSNF. Based on the obtained reduction factors, the expansion of Yucca Mountain Repository (YMR) capacity is evaluated. It is found that with the reference ATW system, the TRU contained in CSNF could be reduced by a factor of ˜170 in terms of inventory and by a factor of ˜40 in terms of toxicity under the assumed scenario. The number of waste packages related to CSNF could be reduced by a factor of ˜8 in terms of

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

DOEpatents

McDermott, D.J.; Schrader, K.J.; Schulz, T.L.

1994-05-03

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

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

DOEpatents

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

1994-01-01

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

Control of Advanced Reactor-Coupled Heat Exchanger System: Incorporation of Reactor Dynamics in System Response to Load Disturbances

DOE PAGES

Skavdahl, Isaac; Utgikar, Vivek; Christensen, Richard; ...

2016-05-24

We present an alternative control schemes for an Advanced High Temperature Reactor system consisting of a reactor, an intermediate heat exchanger, and a secondary heat exchanger (SHX) in this paper. One scheme is designed to control the cold outlet temperature of the SHX (T co) and the hot outlet temperature of the intermediate heat exchanger (T ho2) by manipulating the hot-side flow rates of the heat exchangers (F h/F h2) responding to the flow rate and temperature disturbances. The flow rate disturbances typically require a larger manipulation of the flow rates than temperature disturbances. An alternate strategy examines the controlmore » of the cold outlet temperature of the SHX (T co) only, since this temperature provides the driving force for energy production in the power conversion unit or the process application. The control can be achieved by three options: (1) flow rate manipulation; (2) reactor power manipulation; or (3) a combination of the two. The first option has a quicker response but requires a large flow rate change. The second option is the slowest but does not involve any change in the flow rates of streams. The final option appears preferable as it has an intermediate response time and requires only a minimal flow rate change.« less

The Liquid Annular Reactor System (LARS) propulsion

NASA Technical Reports Server (NTRS)

Powell, James; Ludewig, Hans; Horn, Frederick; Lenard, Roger

1990-01-01

A concept for very high specific impulse (greater than 2000 seconds) direct nuclear propulsion is described. The concept, termed the liquid annular reactor system (LARS), uses liquid nuclear fuel elements to heat hydrogen propellant to very high temperatures (approximately 6000 K). Operating pressure is moderate (approximately 10 atm), with the result that the outlet hydrogen is virtually 100 percent dissociated to monatomic H. The molten fuel is contained in a solid container of its own material, which is rotated to stabilize the liquid layer by centripetal force. LARS reactor designs are described, together with neutronic and thermal-hydraulic analyses. Power levels are on the order of 200 megawatts. Typically, LARS designs use seven rotating fuel elements, are beryllium moderated, and have critical radii of approximately 100 cm (core L/D approximately equal to 1.5).

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

NASA Astrophysics Data System (ADS)

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

Auxiliary reactor for a hydrocarbon reforming system

DOEpatents

Clawson, Lawrence G.; Dorson, Matthew H.; Mitchell, William L.; Nowicki, Brian J.; Bentley, Jeffrey M.; Davis, Robert; Rumsey, Jennifer W.

2006-01-17

An auxiliary reactor for use with a reformer reactor having at least one reaction zone, and including a burner for burning fuel and creating a heated auxiliary reactor gas stream, and heat exchanger for transferring heat from auxiliary reactor gas stream and heat transfer medium, preferably two-phase water, to reformer reaction zone. Auxiliary reactor may include first cylindrical wall defining a chamber for burning fuel and creating a heated auxiliary reactor gas stream, the chamber having an inlet end, an outlet end, a second cylindrical wall surrounding first wall and a second annular chamber there between. The reactor being configured so heated auxiliary reactor gas flows out the outlet end and into and through second annular chamber and conduit which is disposed in second annular chamber, the conduit adapted to carry heat transfer medium and being connectable to reformer reaction zone for additional heat exchange.

Thermal swing reactor including a multi-flight auger

DOEpatents

Ermanoski, Ivan

2017-03-07

A thermal swing reactor including a multi-flight auger and methods for solar thermochemical reactions are disclosed. The reactor includes a multi-flight auger having different helix portions having different pitch. Embodiments of reactors include at least two distinct reactor portions between which there is at least a pressure differential. In embodiments, reactive particles are exchanged between portions during a reaction cycle to thermally reduce the particles at first conditions and oxidize the particles at second conditions to produce chemical work from heat.

Systems and methods for managing shared-path instrumentation and irradiation targets in a nuclear reactor

DOEpatents

Heinold, Mark R.; Berger, John F.; Loper, Milton H.; Runkle, Gary A.

2015-12-29

Systems and methods permit discriminate access to nuclear reactors. Systems provide penetration pathways to irradiation target loading and offloading systems, instrumentation systems, and other external systems at desired times, while limiting such access during undesired times. Systems use selection mechanisms that can be strategically positioned for space sharing to connect only desired systems to a reactor. Selection mechanisms include distinct paths, forks, diverters, turntables, and other types of selectors. Management methods with such systems permits use of the nuclear reactor and penetration pathways between different systems and functions, simultaneously and at only distinct desired times. Existing TIP drives and other known instrumentation and plant systems are useable with access management systems and methods, which can be used in any nuclear plant with access restrictions.

77 FR 55877 - Initial Test Program of Condensate and Feedwater Systems for Light-Water Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2012-09-11

...-492- 3668; email: [email protected] . NRC's Agencywide Documents Access and Management System... Systems for Light-Water Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Regulatory guide; issuance... Systems for Boiling Water Reactor Power Plants.'' This regulatory guide is being revised to: (1) Expand...

Next generation fuel irradiation capability in the High Flux Reactor Petten

NASA Astrophysics Data System (ADS)

Fütterer, Michael A.; D'Agata, Elio; Laurie, Mathias; Marmier, Alain; Scaffidi-Argentina, Francesco; Raison, Philippe; Bakker, Klaas; de Groot, Sander; Klaassen, Frodo

2009-07-01

This paper describes selected equipment and expertise on fuel irradiation testing at the High Flux Reactor (HFR) in Petten, The Netherlands. The reactor went critical in 1961 and holds an operating license up to at least 2015. While HFR has initially focused on Light Water Reactor fuel and materials, it also played a decisive role since the 1970s in the German High Temperature Reactor (HTR) development program. A variety of tests related to fast reactor development in Europe were carried out for next generation fuel and materials, in particular for Very High Temperature Reactor (V/HTR) fuel, fuel for closed fuel cycles (U-Pu and Th-U fuel cycle) and transmutation, as well as for other innovative fuel types. The HFR constitutes a significant European infrastructure tool for the development of next generation reactors. Experimental facilities addressed include V/HTR fuel tests, a coated particle irradiation rig, and tests on fast reactor, transmutation and thorium fuel. The rationales for these tests are given, results are provided and further work is outlined.