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

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

Boldon, Lauren M.; Sabharwall, Piyush

2014-08-01

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

KERENA safety concept in the context of the Fukushima accident

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

Zacharias, T.; Novotny, C.; Bielor, E.

Within the last three years AREVA NP and E.On KK finalized the basic design of KERENA which is a medium sized innovative boiling water reactor, based on the operational experience of German BWR nuclear power plants (NPPs). It is a generation III reactor design with a net electrical output of about 1250 MW. It combines active safety equipment of service-proven designs with new passive safety components, both safety classified. The passive systems utilize basic laws of physics, such as gravity and natural convection, enabling them to function without electric power. Even actuation of these systems is performed thanks to basicmore » physic laws. The degree of diversity in component and system design, achieved by combining active and passive equipment, results in a very low core damage frequency. The Fukushima accident enhanced the world wide discussion about the safety of operating nuclear power plants. World wide stress tests for operating nuclear power plants are being performed embracing both natural and man made hazards. Beside the assessment of existing power plants, also new designs are analyzed regarding the system response to beyond design base accidents. KERENA's optimal combination of diversified cooling systems (active and passive) allows passing efficiently such tests, with a high level of confidence. This paper describes the passive safety components and the KERENA reactor behavior after a Fukushima like accident. (authors)« less

Ultraviolet light-emitting diodes in water disinfection.

PubMed

Vilhunen, Sari; Särkkä, Heikki; Sillanpää, Mika

2009-06-01

The novel system of ultraviolet light-emitting diodes (UV LEDs) was studied in water disinfection. Conventional UV lamps, like mercury vapor lamp, consume much energy and are considered to be problem waste after use. UV LEDs are energy efficient and free of toxicants. This study showed the suitability of LEDs in disinfection and provided information of the effect of two emitted wavelengths and different test mediums to Escherichia coli destruction. Common laboratory strain of E. coli (K12) was used and the effects of two emitted wavelengths (269 and 276 nm) were investigated with two photolytic batch reactors both including ten LEDs. The effects of test medium were examined with ultrapure water, nutrient and water, and nutrient and water with humic acids. Efficiency of reactors was almost the same even though the one emitting higher wavelength had doubled optical power compared to the other. Therefore, the effect of wavelength was evident and the radiation emitted at 269 nm was more powerful. Also, the impact of background was studied and noticed to have only slight deteriorating effect. In the 5-min experiment, the bacterial reduction of three to four log colony-forming units (CFU) per cubic centimeter was achieved, in all cases. When turbidity of the test medium was greater, part of the UV radiation was spent on the absorption and reactions with extra substances on liquid. Humic acids can also coat the bacteria reducing the sensitivity of the cells to UV light. The lower wavelength was distinctly more efficient when the optical power is considered, even though the difference of wavelengths was small. The reason presumably is the greater absorption of DNA causing more efficient bacterial breakage. UV LEDs were efficient in E. coli destruction, even if LEDs were considered to have rather low optical power. The effect of wavelengths was noticeable but the test medium did not have much impact. This study found UV LEDs to be an optimal method for bacterial disinfection. The emitted wavelength was found to be an essential factor when using LEDs; thus, care should be taken in selecting the proper LED for maximum disinfection.

Applications of the Aqueous Self-Cooled Blanket concept

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

Steiner, D.; Embrechts, M.J.; Varsamis, G.

1986-11-01

In this paper a novel water-cooled blanket concept is examined. This concept, designated the Aqueous Self-Cooled Blanket (ASCB), employs water with small amounts of dissolved fertile compounds as both the coolant and the breeding medium. The ASCB concept is reviewed and its application in three different contexts is examined: (1) power reactors; (2) near-term devices such as NET; and (3) fusion-fission hybrids.

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.

The role of accelerators in the nuclear fuel cycle

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

Takahashi, Hiroshi.

1990-01-01

The use of neutrons produced by the medium energy proton accelerator (1 GeV--3 GeV) has considerable potential in reconstructing the nuclear fuel cycle. About 1.5 {approximately} 2.5 ton of fissile material can be produced annually by injecting a 450 MW proton beam directly into fertile materials. A source of neutrons, produced by a proton beam, to supply subcritical reactors could alleviate many of the safety problems associated with critical assemblies, such as positive reactivity coefficients due to coolant voiding. The transient power of the target can be swiftly controlled by controlling the power of the proton beam. Also, the usemore » of a proton beam would allow more flexibility in the choice of fuel and structural materials which otherwise might reduce the reactivity of reactors. This paper discusses the rate of accelerators in the transmutation of radioactive wastes of the nuclear fuel cycles. 34 refs., 17 figs., 9 tabs.« less

Wave propagation simulation in the upper core of sodium-cooled fast reactors using a spectral-element method for heterogeneous media

NASA Astrophysics Data System (ADS)

Nagaso, Masaru; Komatitsch, Dimitri; Moysan, Joseph; Lhuillier, Christian

2018-01-01

ASTRID project, French sodium cooled nuclear reactor of 4th generation, is under development at the moment by Alternative Energies and Atomic Energy Commission (CEA). In this project, development of monitoring techniques for a nuclear reactor during operation are identified as a measure issue for enlarging the plant safety. Use of ultrasonic measurement techniques (e.g. thermometry, visualization of internal objects) are regarded as powerful inspection tools of sodium cooled fast reactors (SFR) including ASTRID due to opacity of liquid sodium. In side of a sodium cooling circuit, heterogeneity of medium occurs because of complex flow state especially in its operation and then the effects of this heterogeneity on an acoustic propagation is not negligible. Thus, it is necessary to carry out verification experiments for developments of component technologies, while such kind of experiments using liquid sodium may be relatively large-scale experiments. This is why numerical simulation methods are essential for preceding real experiments or filling up the limited number of experimental results. Though various numerical methods have been applied for a wave propagation in liquid sodium, we still do not have a method for verifying on three-dimensional heterogeneity. Moreover, in side of a reactor core being a complex acousto-elastic coupled region, it has also been difficult to simulate such problems with conventional methods. The objective of this study is to solve these 2 points by applying three-dimensional spectral element method. In this paper, our initial results on three-dimensional simulation study on heterogeneous medium (the first point) are shown. For heterogeneity of liquid sodium to be considered, four-dimensional temperature field (three spatial and one temporal dimension) calculated by computational fluid dynamics (CFD) with Large-Eddy Simulation was applied instead of using conventional method (i.e. Gaussian Random field). This three-dimensional numerical experiment yields that we could verify the effects of heterogeneity of propagation medium on waves in Liquid sodium.

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.

Thermal analysis of heat and power plant with high temperature reactor and intermediate steam cycle

NASA Astrophysics Data System (ADS)

Fic, Adam; Składzień, Jan; Gabriel, Michał

2015-03-01

Thermal analysis of a heat and power plant with a high temperature gas cooled nuclear reactor is presented. The main aim of the considered system is to supply a technological process with the heat at suitably high temperature level. The considered unit is also used to produce electricity. The high temperature helium cooled nuclear reactor is the primary heat source in the system, which consists of: the reactor cooling cycle, the steam cycle and the gas heat pump cycle. Helium used as a carrier in the first cycle (classic Brayton cycle), which includes the reactor, delivers heat in a steam generator to produce superheated steam with required parameters of the intermediate cycle. The intermediate cycle is provided to transport energy from the reactor installation to the process installation requiring a high temperature heat. The distance between reactor and the process installation is assumed short and negligable, or alternatively equal to 1 km in the analysis. The system is also equipped with a high temperature argon heat pump to obtain the temperature level of a heat carrier required by a high temperature process. Thus, the steam of the intermediate cycle supplies a lower heat exchanger of the heat pump, a process heat exchanger at the medium temperature level and a classical steam turbine system (Rankine cycle). The main purpose of the research was to evaluate the effectiveness of the system considered and to assess whether such a three cycle cogeneration system is reasonable. Multivariant calculations have been carried out employing the developed mathematical model. The results have been presented in a form of the energy efficiency and exergy efficiency of the system as a function of the temperature drop in the high temperature process heat exchanger and the reactor pressure.

Fuel development for gas-cooled fast reactors

NASA Astrophysics Data System (ADS)

Meyer, M. K.; Fielding, R.; Gan, J.

2007-09-01

The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High-Temperature Reactor (VHTR), as well as actinide burning concepts [A Technology Roadmap for Generation IV Nuclear Energy Systems, US DOE Nuclear Energy Research Advisory Committee and the Generation IV International Forum, December 2002]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the US and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic 'honeycomb' structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

Controlled-Turbulence Bioreactors

NASA Technical Reports Server (NTRS)

Wolf, David A.; Schwartz, Ray; Trinh, Tinh

1989-01-01

Two versions of bioreactor vessel provide steady supplies of oxygen and nutrients with little turbulence. Suspends cells in environment needed for sustenance and growth, while inflicting less damage from agitation and bubbling than do propeller-stirred reactors. Gentle environments in new reactors well suited to delicate mammalian cells. One reactor kept human kidney cells alive for as long as 11 days. Cells grow on carrier beads suspended in liquid culture medium that fills cylindrical housing. Rotating vanes - inside vessel but outside filter - gently circulates nutrient medium. Vessel stationary; magnetic clutch drives filter cylinder and vanes. Another reactor creates even less turbulence. Oxygen-permeable tubing wrapped around rod extending along central axis. Small external pump feeds oxygen to tubing through rotary coupling, and oxygen diffuses into liquid medium.

Design requirements for innovative homogeneous reactor, lesson learned from Fukushima accident

NASA Astrophysics Data System (ADS)

Arbie, Bakri; Pinem, Suryan; Sembiring, Tagor; Subki, Iyos

2012-06-01

The Fukushima disaster is the largest nuclear accident since the 1986 Chernobyl disaster, but it is more complex as multiple reactors and spent fuel pools are involved. The severity of the nuclear accident is rated 7 in the International Nuclear Events Scale. Expert said that "Fukushima is the biggest industrial catastrophe in the history of mankind". According to Mitsuru Obe, in The Wall Street Journal, May 16th of 2011, TEPCO estimates the nuclear fuel was exposed to the air less than five hours after the earthquake struck. Fuel rods melted away rapidly as the temperatures inside the core reached 2800 C within six hours. In less than 16 hours, the reactor core melted and dropped to the bottom of the pressure vessel. The information should be evaluated in detail. In Germany several nuclear power plant were shutdown, Italy postponed it's nuclear power program and China reviewed their nuclear power program. Different news come from Britain, in October 11, 2011, the Safety Committee said all clear for nuclear power in Britain, because there are no risk of strong earthquake and tsunami in the region. Due to this severe fact, many nuclear scientists and engineer from all over the world are looking for a new approach, such as homogeneous reactor which was developed in Oak Ridge National Laboratory in 1960-ies, during Dr. Alvin Weinberg tenure as the Director of ORNL. The paper will describe the design requirement that will be used as the basis for innovative homogeneous reactor. Innovative Homogeneous Reactor is expected to reduce core melt by two decades (4), since the fuel is intermix homogeneously with coolant and secondly we eliminate the used fuel rod which need to be cooled for a long period of time. In order to be successful for its implementation of the innovative system, testing and validation, three phases of development will be introduced. The first phase is Low Level Goals is really the proof of concept;the Medium Level Goal is Technical Goalsand the High Level Goals which is Business Goals.

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.

Production of polygalacturonases by Aspergillus oryzae in stirred tank and internal- and external-loop airlift reactors.

PubMed

Fontana, Roselei Claudete; da Silveira, Maurício Moura

2012-11-01

The production of endo- and exo-polygalacturonase (PG) by Aspergillus oryzae was assessed in stirred tank reactors (STRs), internal-loop airlift reactors (ILARs) and external-loop airlift reactors (ELARs). For STR production, we compared culture media formulated with either pectin (WBE) or partially hydrolyzed pectin. The highest enzyme activities were obtained in medium that contained 50% pectin in hydrolyzed form (WBE5). PG production in the three reactor types was compared for WBE5 and low salt WBE medium, with additional salts added at 48, 60 and 72h (WBES). The ELARs performed better than the ILARs in WBES medium where the exo-PG was the same concentration as for STRs and the endo-PG was 20% lower. These results indicate that PG production is higher under experimental conditions that result in higher cell growth with minimum pH values less than 3.0. Copyright © 2012 Elsevier Ltd. All rights reserved.

Use of liquid metals in nuclear and thermonuclear engineering, and in other innovative technologies

NASA Astrophysics Data System (ADS)

Rachkov, V. I.; Arnol'dov, M. N.; Efanov, A. D.; Kalyakin, S. G.; Kozlov, F. A.; Loginov, N. I.; Orlov, Yu. I.; Sorokin, A. P.

2014-05-01

By now, a good deal of experience has been gained with using liquid metals as coolants in nuclear power installations; extensive knowledge has been gained about the physical, thermophysical, and physicochemical properties of these coolants; and the scientific principles and a set of methods and means for handling liquid metals as coolants for nuclear power installations have been elaborated. Prototype and commercialgrade sodium-cooled NPP power units have been developed, including the BOR-60, BN-350, and BN-600 power units (the Soviet Union); the Rapsodie, Phenix, and Superphenix power units (France), the EBR-II power unit (the United States); and the PFR power unit (the United Kingdom). In Russia, dedicated nuclear power installations have been constructed, including those with a lead-bismuth coolant for nuclear submarines and with sodium-potassium alloy for spacecraft (the Buk and Topol installations), which have no analogs around the world. Liquid metals (primarily lithium and its alloy with lead) hold promise for use in thermonuclear power engineering, where they can serve not only as a coolant, but also as tritium-producing medium. In this article, the physicochemical properties of liquid metal coolants, as well as practical experience gained from using them in nuclear and thermonuclear power engineering and in innovative technologies are considered, and the lines of further research works are formulated. New results obtained from investigations carried out on the Pb-Bi and Pb for the SVBR and BREST fast-neutron reactors (referred to henceforth as fast reactors) and for controlled accelerator systems are described.

The role of inertial fusion energy in the energy marketplace of the 21st century and beyond

NASA Astrophysics Data System (ADS)

John Perkins, L.

The viability of inertial fusion in the 21st century and beyond will be determined by its ultimate cost, complexity, and development path relative to other competing, long term, primary energy sources. We examine this potential marketplace in terms of projections for population growth, energy demands, competing fuel sources and environmental constraints (CO 2), and show that the two competitors for inertial fusion energy (IFE) in the medium and long term are methane gas hydrates and advanced, breeder fission; both have potential fuel reserves that will last for thousands of years. Relative to other classes of fusion concepts, we argue that the single largest advantage of the inertial route is the perception by future customers that the IFE fusion power core could achieve credible capacity factors, a result of its relative simplicity, the decoupling of the driver and reactor chamber, and the potential to employ thick liquid walls. In particular, we show that the size, cost and complexity of the IFE reactor chamber is little different to a fission reactor vessel of the same thermal power. Therefore, relative to fission, because of IFE's tangible advantages in safety, environment, waste disposal, fuel supply and proliferation, our research in advanced targets and innovative drivers can lead to a certain, reduced-size driver at which future utility executives will be indifferent to the choice of an advanced fission plant or an advanced IFE power plant; from this point on, we have a competitive commercial product. Finally, given that the major potential customer for energy in the next century is the present developing world, we put the case for future IFE "reservations" which could be viable propositions providing sufficient reliability and redundancy can be realized for each modular reactor unit.

Using real options to evaluate the flexibility in the deployment of SMR

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

Locatelli, G.; Mancini, M.; Ruiz, F.

2012-07-01

According to recent estimations the financial gap between Large Reactors (LR) and Small Medium Reactors (SMRs) seems not as huge as the economy of scale would suggest, so the SMRs are going to be important players of the worldwide nuclear renaissance. POLIMIs INCAS model has been developed to compare the investment in SMR with respect to LR. It provides the value of IRR (Internal Rate of Return), NPV (Net Present Value), LUEC (Levelized Unitary Electricity Cost), up-front investment, etc. The aim of this research is to integrate the actual INCAS model, based on discounted cash flows, with the real optionmore » theory to measure flexibility of the investor to expand, defer or abandon a nuclear project, under future uncertainties. The work compares the investment in a large nuclear power plant with a series of smaller, modular nuclear power plants on the same site. As a consequence it compares the benefits of the large power plant, coming from the economy of scale, to the benefit of the modular project (flexibility) concluding that managerial flexibility can be measured and used by an investor to face the investment risks. (authors)« less

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.

10 CFR 52.167 - Issuance of manufacturing license.

Code of Federal Regulations, 2010 CFR

2010-01-01

... proposed reactor(s) can be incorporated into a nuclear power plant and operated at sites having... design and manufacture the proposed nuclear power reactor(s); (5) The proposed inspections, tests... the construction of a nuclear power facility using the manufactured reactor(s). (2) A holder of a...

10 CFR 50.60 - Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for...

Code of Federal Regulations, 2010 CFR

2010-01-01

... lightwater nuclear power reactors for normal operation. 50.60 Section 50.60 Energy NUCLEAR REGULATORY... lightwater nuclear power reactors for normal operation. (a) Except as provided in paragraph (b) of this section, all light-water nuclear power reactors, other than reactor facilities for which the...

10 CFR 50.60 - Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for...

Code of Federal Regulations, 2011 CFR

2011-01-01

... lightwater nuclear power reactors for normal operation. 50.60 Section 50.60 Energy NUCLEAR REGULATORY... lightwater nuclear power reactors for normal operation. (a) Except as provided in paragraph (b) of this section, all light-water nuclear power reactors, other than reactor facilities for which the...

The component content of active particles in a plasma-chemical reactor based on volume barrier discharge

NASA Astrophysics Data System (ADS)

Soloshenko, I. A.; Tsiolko, V. V.; Pogulay, S. S.; Terent'yeva, A. G.; Bazhenov, V. Yu; Shchedrin, A. I.; Ryabtsev, A. V.; Kuzmichev, A. I.

2007-02-01

In this paper the results of theoretical and experimental studies of the component content of active particles formed in a plasma-chemical reactor composed of a multiple-cell generator of active particles, based on volume barrier discharge, and a working chamber are presented. For calculation of the content of uncharged plasma components an approach is proposed which is based on averaging of the power introduced over the entire volume. Advantages of such an approach lie in an absence of fitting parameters, such as the dimensions of microdischarges, their surface density and rate of breakdown. The calculation and the experiment were accomplished with the use of dry air (20% relative humidity) as the plasma generating medium. Concentrations of O3, HNO3, HNO2, N2 O5 and NO3 were measured experimentally in the discharge volume and working chamber for the residence time of particles on a discharge of 0.3 s and more and discharge specific power of 1.5 W cm-3. It has been determined that the best agreement between the calculation and the experiment occurs at calculated gas medium temperatures in the discharge plasma of about 400-425 K, which correspond to the experimentally measured rotational temperature of nitrogen. In most cases the calculated concentrations of O3, HNO3, HNO2, N2O5 and NO3 for the barrier discharge and the working chamber are in fairly good agreement with the respective measured values.

Biological production of ethanol from coal. Task 4 report, Continuous reactor studies

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

Not Available

The production of ethanol from synthesis gas by the anaerobic bacterium C. ljungdahlii has been demonstrated in continuous stirred tank reactors (CSTRs), CSTRs with cell recycle and trickle bed reactors. Various liquid media were utilized in these studies including basal medium, basal media with 1/2 B-vitamins and no yeast extract and a medium specifically designed for the growth of C. ljungdahlii in the CSTR. Ethanol production was successful in each of the three reactor types, although trickle bed operation with C. ljungdahlii was not as good as with the stirred tank reactors. Operation in the CSTR with cell recycle wasmore » particularly promising, producing 47 g/L ethanol with only minor concentrations of the by-product acetate.« less

Reactor engineering support of operations at the Davis-Besse nuclear power station

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

Kelley, D.B.

1995-12-31

Reactor engineering functions differ greatly from unit to unit; however, direct support of the reactor operators during reactor startups and operational transients is common to all units. This paper summarizes the support the reactor engineers provide the reactor operators during reactor startups and power changes through the use of automated computer programs at the Davis-Besse nuclear power station.

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which the...

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which the...

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which the...

Acetone-butanol-ethanol (ABE) fermentation in an immobilized cell trickle bed reactor.

PubMed

Park, C H; Okos, M R; Wankat, P C

1989-06-05

Acetone-butanol-ethanol (ABE) fermentation was successfully carried out in an immobilized cell trickle bed reactor. The reactor was composed of two serial columns packed with Clostridium acetobutylicum ATCC 824 entrapped on the surface of natural sponge segments at a cell loading in the range of 2.03-5.56 g dry cells/g sponge. The average cell loading was 3.58 g dry cells/g sponge. Batch experiments indicated that a critical pH above 4.2 is necessary for the initiation of cell growth. One of the media used during continuous experiments consisted of a salt mixture alone and the other a nutrient medium containing a salt mixture with yeast extract and peptone. Effluent pH was controlled by supplying various fractions of the two different types of media. A nutrient medium fraction above 0.6 was crucial for successful fermentation in a trickle bed reactor. The nutrient medium fraction is the ratio of the volume of the nutrient medium to the total volume of nutrient plus salt medium. Supplying nutrient medium to both columns continuously was an effective way to meet both pH and nutrient requirement. A 257-mL reactor could ferment 45 g/L glucose from an initial concentration of 60 g/L glucose at a rate of 70 mL/h. Butanol, acetone, and ethanol concentrations were 8.82, 5.22, and 1.45 g/L, respectively, with a butanol and total solvent yield of 19.4 and 34.1 wt %. Solvent productivity in an immobilized cell trickle bed reactor was 4.2 g/L h, which was 10 times higher than that obtained in a batch fermentation using free cells and 2.76 times higher than that of an immobilized CSTR. If the nutrient medium fraction was below 0.6 and the pH was below 4.2, the system degenerated. Oxygen also contributed to the system degeneration. Upon degeneration, glucose consumption and solvent yield decreased to 30.9 g/L and 23.0 wt %, respectively. The yield of total liquid product (40.0 wt %) and butanol selectivity (60.0 wt %) remained almost constant. Once the cells were degenerated, they could not be recovered.

Characterization of an LED based photoreactor to degrade 4-chlorophenol in an aqueous medium using coumarin (C-343) sensitized TiO2.

PubMed

Ghosh, Jyoti P; Langford, Cooper H; Achari, Gopal

2008-10-16

A detailed performance evaluation of a simple high intensity LED based photoreactor exploiting a narrow wavelength range of the LED to match the spectrum of a dye in a photocatalysis system is reported. A dye sensitized (coumarin-343, lambda max = 446 nm) TiO 2 photocatalyst was used for the degradation of 4-chlorophenol (4-CP) in an aqueous medium using the 436 nm LED based photoreactor. The LED reactor performed competitively with a conventional multilamp reactor and sunlight in the degradation of 4-CP. Light intensities entering the reaction vessel were measured by conventional ferrioxalate actinometry. The results can be fitted by approximate first order kinetic behavior in this system. Hydroxyl radicals were detected by spin trapping EPR, and effects of OH radical quenchers on kinetics suggest that the reaction is initiated by these radicals or their equivalents. LEDs operating at competitive intensities offer a number of advantages to the photochemist or the environmental engineer via long life, efficient current to light conversion, narrow bandwidth, forward directed output, and direct current power for remote operation. Matching light source spectrum to chromophore is a key.

Thermally integrated staged methanol reformer and method

DOEpatents

Skala, Glenn William; Hart-Predmore, David James; Pettit, William Henry; Borup, Rodney Lynn

2001-01-01

A thermally integrated two-stage methanol reformer including a heat exchanger and first and second reactors colocated in a common housing in which a gaseous heat transfer medium circulates to carry heat from the heat exchanger into the reactors. The heat transfer medium comprises principally hydrogen, carbon dioxide, methanol vapor and water vapor formed in a first stage reforming reaction. A small portion of the circulating heat transfer medium is drawn off and reacted in a second stage reforming reaction which substantially completes the reaction of the methanol and water remaining in the drawn-off portion. Preferably, a PrOx reactor will be included in the housing upstream of the heat exchanger to supplement the heat provided by the heat exchanger.

10 CFR 72.1 - Purpose.

Code of Federal Regulations, 2011 CFR

2011-01-01

... RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Provisions § 72.1 Purpose. The... receive, transfer, and possess power reactor spent fuel, power reactor-related Greater than Class C (GTCC... reactor spent fuel, high-level radioactive waste, power reactor-related GTCC waste, and other radioactive...

10 CFR 72.1 - Purpose.

Code of Federal Regulations, 2010 CFR

2010-01-01

... RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Provisions § 72.1 Purpose. The... receive, transfer, and possess power reactor spent fuel, power reactor-related Greater than Class C (GTCC... reactor spent fuel, high-level radioactive waste, power reactor-related GTCC waste, and other radioactive...

Linear free energy correlations for fission product release from the Fukushima-Daiichi nuclear accident.

PubMed

Abrecht, David G; Schwantes, Jon M

2015-03-03

This paper extends the preliminary linear free energy correlations for radionuclide release performed by Schwantes et al., following the Fukushima-Daiichi Nuclear Power Plant accident. Through evaluations of the molar fractionations of radionuclides deposited in the soil relative to modeled radionuclide inventories, we confirm the initial source of the radionuclides to the environment to be from active reactors rather than the spent fuel pool. Linear correlations of the form In χ = −α ((ΔGrxn°(TC))/(RTC)) + β were obtained between the deposited concentrations, and the reduction potentials of the fission product oxide species using multiple reduction schemes to calculate ΔG°rxn (TC). These models allowed an estimate of the upper bound for the reactor temperatures of TC between 2015 and 2060 K, providing insight into the limiting factors to vaporization and release of fission products during the reactor accident. Estimates of the release of medium-lived fission products 90Sr, 121mSn, 147Pm, 144Ce, 152Eu, 154Eu, 155Eu, and 151Sm through atmospheric venting during the first month following the accident were obtained, indicating that large quantities of 90Sr and radioactive lanthanides were likely to remain in the damaged reactor cores.

77 FR 8902 - Draft Regulatory Guide: Issuance, Availability Decommissioning of Nuclear Power Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2012-02-15

... Decommissioning of Nuclear Power Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Draft regulatory guide... draft regulatory guide (DG) DG-1271 ``Decommissioning of Nuclear Power Reactors.'' This guide describes... Regulatory Guide 1.184, ``Decommissioning of Nuclear Power Reactors,'' dated July 2000. This proposed...

10 CFR 50.72 - Immediate notification requirements for operating nuclear power reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

... power reactors. 50.72 Section 50.72 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... notification requirements for operating nuclear power reactors. (a) General requirements. 1 (1) Each nuclear... requirements for immediate notification of the NRC by licensed operating nuclear power reactors are contained...

10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license...

10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 2 2013-01-01 2013-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license...

10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 2 2014-01-01 2014-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license...

10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license...

10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 2 2012-01-01 2012-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license...

Zirconium Hydride Space Power Reactor design.

NASA Technical Reports Server (NTRS)

Asquith, J. G.; Mason, D. G.; Stamp, S.

1972-01-01

The Zirconium Hydride Space Power Reactor being designed and fabricated at Atomics International is intended for a wide range of potential applications. Throughout the program a series of reactor designs have been evaluated to establish the unique requirements imposed by coupling with various power conversion systems and for specific applications. Current design and development emphasis is upon a 100 kilowatt thermal reactor for application in a 5 kwe thermoelectric space power generating system, which is scheduled to be fabricated and ground tested in the mid 70s. The reactor design considerations reviewed in this paper will be discussed in the context of this 100 kwt reactor and a 300 kwt reactor previously designed for larger power demand applications.

Heat exchanger for reactor core and the like

DOEpatents

Kaufman, Jay S.; Kissinger, John A.

1986-01-01

A compact bayonet tube type heat exchanger which finds particular application as an auxiliary heat exchanger for transfer of heat from a reactor gas coolant to a secondary fluid medium. The heat exchanger is supported within a vertical cavity in a reactor vessel intersected by a reactor coolant passage at its upper end and having a reactor coolant return duct spaced below the inlet passage. The heat exchanger includes a plurality of relatively short length bayonet type heat exchange tube assemblies adapted to pass a secondary fluid medium therethrough and supported by primary and secondary tube sheets which are releasibly supported in a manner to facilitate removal and inspection of the bayonet tube assemblies from an access area below the heat exchanger. Inner and outer shrouds extend circumferentially of the tube assemblies and cause the reactor coolant to flow downwardly internally of the shrouds over the tube bundle and exit through the lower end of the inner shroud for passage to the return duct in the reactor vessel.

A small, 1400 K, reactor for Brayton space power systems.

NASA Technical Reports Server (NTRS)

Lantz, E.; Mayo, W.

1972-01-01

An investigation was conducted to determine minimum dimensions and minimum weight obtainable in a design for a reactor using uranium-233 nitride or plutonium-239 nitride as fuel. Such a reactor had been considered by Krasner et al. (1971). Present space power status is discussed, together with questions of reactor design and power distribution in the reactor. The characteristics of various reactor types are compared, giving attention also to a zirconium hydride reactor.

10 CFR 50.72 - Immediate notification requirements for operating nuclear power reactors.

Code of Federal Regulations, 2010 CFR

2010-01-01

... power reactors. 50.72 Section 50.72 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... notification requirements for operating nuclear power reactors. (a) General requirements. 1 (1) Each nuclear power reactor licensee licensed under §§ 50.21(b) or 50.22 holding an operating license under this part...

10 CFR 50.72 - Immediate notification requirements for operating nuclear power reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... power reactors. 50.72 Section 50.72 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... notification requirements for operating nuclear power reactors. (a) General requirements. 1 (1) Each nuclear power reactor licensee licensed under §§ 50.21(b) or 50.22 holding an operating license under this part...

Decommissioning of German Nuclear Research Facilities under the Governance of the Federal Ministry of Education and Research

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

Weigl, M.

2008-07-01

Since the announcement of the first nuclear program in 1956, nuclear R and D in Germany has been supported by the Federal Government under four nuclear programs and later on under more general energy R and D programs. The original goal was to help German industry to achieve safe, low-cost generation of energy and self-sufficiency in the various branches of nuclear technology, including the fast breeder reactor and the fuel cycle. Several national research centers were established to host or operate experimental and demonstration plants. These are mainly located at the sites of the national research centers at Juelich andmore » Karlsruhe. In the meantime, all these facilities were shut down and most of them are now in a state of decommissioning and dismantling (D and D). Meanwhile, Germany is one of the leading countries in the world in the field of D and D. Two big demonstration plants, the Niederaichbach Nuclear Power Plant (KKN) a heavy-water cooled pressure tube reactor with carbon-dioxide cooling and the Karlstein Superheated Steam Reactor (HDR) a boiling light water reactor with a thermal power of 100 MW, are totally dismantled and 'green field' is reached. For two other projects the return to 'green field' sites will be reached by the end of this decade. These are the dismantling of the Multi-Purpose Research Reactor (MZFR) and the Compact Sodium Cooled Reactor (KNK) both located at the Forschungszentrum Karlsruhe. Within these projects a lot of new solutions und innovative techniques were tested, which were developed at German universities and in small and medium sized companies mostly funded by the Federal Ministry of Education and Research (BMBF). For example, high performance underwater cutting technologies like plasma arc cutting and contact arc metal cutting. (authors)« less

Power monitoring in space nuclear reactors using silicon carbide radiation detectors

NASA Technical Reports Server (NTRS)

Ruddy, Frank H.; Patel, Jagdish U.; Williams, John G.

2005-01-01

Space reactor power monitors based on silicon carbide (SiC) semiconductor neutron detectors are proposed. Detection of fast leakage neutrons using SiC detectors in ex-core locations could be used to determine reactor power: Neutron fluxes, gamma-ray dose rates and ambient temperatures have been calculated as a function of distance from the reactor core, and the feasibility of power monitoring with SiC detectors has been evaluated at several ex-core locations. Arrays of SiC diodes can be configured to provide the required count rates to monitor reactor power from startup to full power Due to their resistance to temperature and the effects of neutron and gamma-ray exposure, SiC detectors can be expected to provide power monitoring information for the fill mission of a space reactor.

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.

Five Lectures on Nuclear Reactors Presented at Cal Tech

DOE R&D Accomplishments Database

Weinberg, Alvin M.

1956-02-10

The basic issues involved in the physics and engineering of nuclear reactors are summarized. Topics discussed include theory of reactor design, technical problems in power reactors, physical problems in nuclear power production, and future developments in nuclear power. (C.H.)

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... behavior of the reactor system during a loss-of-coolant accident. Comparisons to applicable experimental...

Digital computer study of nuclear reactor thermal transients during startup of 60-kWe Brayton power conversion system

NASA Technical Reports Server (NTRS)

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

1974-01-01

A digital computer study was made of reactor thermal transients during startup of the Brayton power conversion loop of a 60-kWe reactor Brayton power system. A startup procedure requiring the least Brayton system complication was tried first; this procedure caused violations of design limits on key reactor variables. Several modifications of this procedure were then found which caused no design limit violations. These modifications involved: (1) using a slower rate of increase in gas flow; (2) increasing the initial reactor power level to make the reactor respond faster; and (3) appropriate reactor control drum manipulation during the startup transient.

Energy-technological complex with reactor for torrefaction

NASA Astrophysics Data System (ADS)

Kuzmina, J. S.; Director, L. B.; Zaichenko, V. M.

2016-11-01

To eliminate shortcomings of raw plant materials pelletizing process with thermal treatment (low-temperature pyrolysis or torrefaction) can be applied. This paper presents a mathematical model of energy-technological complex (ETC) for combined production of heat, electricity and solid biofuels torrefied pellets. According to the structure the mathematical model consists of mathematical models of main units of ETC and the relationships between them and equations of energy and material balances. The equations describe exhaust gas straining action through a porous medium formed by pellets. Decomposition rate of biomass was calculated by using the gross-reaction diagram, which is responsible for the disintegration of raw material. A mathematical model has been tested according to bench experiments on one reactor module. From nomographs, designed for a particular configuration of ETC it is possible to determine the basic characteristics of torrefied pellets (rate of weight loss, heating value and heat content) specifying only two parameters (temperature and torrefaction time). It is shown that the addition of reactor for torrefaction to gas piston engine can improve the energy efficiency of power plant.

Applicability of 100kWe-class of space reactor power systems to NASA manned space station missions

NASA Technical Reports Server (NTRS)

Silverman, S. W.; Willenberg, H. J.; Robertson, C.

1985-01-01

An assessment is made of a manned space station operating with sufficiently high power demands to require a multihundred kilowatt range electrical power system. The nuclear reactor is a competitor for supplying this power level. Load levels were selected at 150kWe and 300kWe. Interactions among the reactor electrical power system, the manned space station, the space transportation system, and the mission were evaluated. The reactor shield and the conversion equipment were assumed to be in different positions with respect to the station; on board, tethered, and on a free flyer platform. Mission analyses showed that the free flyer concept resulted in unacceptable costs and technical problems. The tethered reactor providing power to an electrolyzer for regenerative fuel cells on the space station, results in a minimum weight shield and can be designed to release the reactor power section so that it moves to a high altitude orbit where the decay period is at least 300 years. Placing the reactor on the station, on a structural boom is an attractive design, but heavier than the long tethered reactor design because of the shield weight for manned activity near the reactor.

Potential of Electric Power Production from Microbial Fuel Cell (MFC) in Evapotranspiration Reactor for Leachate Treatment Using Alocasia macrorrhiza Plant and Eleusine indica Grass

NASA Astrophysics Data System (ADS)

Zaman, Badrus; Wardhana, Irawan Wisnu

2018-02-01

Microbial fuel cell is one of attractive electric power generator from nature bacterial activity. While, Evapotranspiration is one of the waste water treatment system which developed to eliminate biological weakness that utilize the natural evaporation process and bacterial activity on plant roots and plant media. This study aims to determine the potential of electrical energy from leachate treatment using evapotranspiration reactor. The study was conducted using local plant, namely Alocasia macrorrhiza and local grass, namely Eleusine Indica. The system was using horizontal MFC by placing the cathodes and anodes at different chamber (i.e. in the leachate reactor and reactor with plant media). Carbon plates was used for chatode-anodes material with size of 40 cm x 10 cm x1 cm. Electrical power production was measure by a digital multimeter for 30 days reactor operation. The result shows electric power production was fluctuated during reactor operation from all reactors. The electric power generated from each reactor was fluctuated, but from the reactor using Alocasia macrorrhiza plant reach to 70 μwatt average. From the reactor using Eleusine Indica grass was reached 60 μwatt average. Electric power production fluctuation is related to the bacterial growth pattern in the soil media and on the plant roots which undergo the adaptation process until the middle of the operational period and then in stable growth condition until the end of the reactor operation. The results indicate that the evapotranspiration reactor using Alocasia macrorrhiza plant was 60-95% higher electric power potential than using Eleusine Indica grass in short-term (30-day) operation. Although, MFC system in evapotranspiration reactor system was one of potential system for renewable electric power generation.

77 FR 60039 - Non-Power Reactor License Renewal

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-02

... NUCLEAR REGULATORY COMMISSION 10 CFR Part 50 [NRC-2011-0087] RIN 3150-AI96 Non-Power Reactor... the final regulatory basis for rulemaking to streamline non-power reactor license renewal. This final... Reactor (RTR) License Renewal Process. This contemplated rulemaking also recommends conforming changes to...

77 FR 38742 - Non-Power Reactor License Renewal

Federal Register 2010, 2011, 2012, 2013, 2014

2012-06-29

...-0087] RIN 3150-AI96 Non-Power Reactor License Renewal AGENCY: Nuclear Regulatory Commission. ACTION... reactors. This contemplated rulemaking would also make conforming changes to address technical issues in existing non-power reactor regulations. The NRC is seeking input from the public, licensees, certificate...

10 CFR 50.64 - Limitations on the use of highly enriched uranium (HEU) in domestic non-power reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... domestic non-power reactors. 50.64 Section 50.64 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... Permits § 50.64 Limitations on the use of highly enriched uranium (HEU) in domestic non-power reactors. (a) Applicability. The requirements of this section apply to all non-power reactors. (b) Requirements. (1) The...

10 CFR 50.83 - Release of part of a power reactor facility or site for unrestricted use.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 1 2011-01-01 2011-01-01 false Release of part of a power reactor facility or site for... of a power reactor facility or site for unrestricted use. (a) Prior written NRC approval is required... release. Nuclear power reactor licensees seeking NRC approval shall— (1) Evaluate the effect of releasing...

10 CFR 50.64 - Limitations on the use of highly enriched uranium (HEU) in domestic non-power reactors.

Code of Federal Regulations, 2010 CFR

2010-01-01

... domestic non-power reactors. 50.64 Section 50.64 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... Permits § 50.64 Limitations on the use of highly enriched uranium (HEU) in domestic non-power reactors. (a) Applicability. The requirements of this section apply to all non-power reactors. (b) Requirements. (1) The...

10 CFR 50.83 - Release of part of a power reactor facility or site for unrestricted use.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Release of part of a power reactor facility or site for... of a power reactor facility or site for unrestricted use. (a) Prior written NRC approval is required... release. Nuclear power reactor licensees seeking NRC approval shall— (1) Evaluate the effect of releasing...

Inherently Safe Fission Power System for Lunar Outposts

NASA Astrophysics Data System (ADS)

Schriener, Timothy M.; El-Genk, Mohamed S.

2013-09-01

This paper presents the Solid Core-Sectored Compact Reactor (SC-SCoRe) and power system for future lunar outposts. The power system nominally provides 38 kWe continuously for 21 years, employs static components and has no single point failures in reactor cooling or power generation. The reactor core has six sectors, each has a separate pair of primary and secondary loops with liquid NaK-56 working fluid, thermoelectric (TE) power conversion and heat-pipes radiator panels. The electromagnetic (EM) pumps in the primary and secondary loops, powered with separate TE power units, ensure operation reliability and passive decay heat removal from the reactor after shutdown. The reactor poses no radiological concerns during launch, and remains sufficiently subcritical, with the radial reflector dissembled, when submerged in wet sand and the core flooded with seawater, following a launch abort accident. After 300 years of storage below grade on the Moon, the total radioactivity in the post-operation reactor drops below 164 Ci, a low enough radioactivity for a recovery and safe handling of the reactor.

Brookhaven highlights, October 1978-September 1979. [October 1978 to September 1979

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

Not Available

1979-01-01

These highlights present an overview of the major research and development achievements at Brookhaven National Laboratory from October 1978 to September 1979. Specific areas covered include: accelerator and high energy physics programs; high energy physics research; the AGS and improvements to the AGS; neutral beam development; heavy ion fusion; superconducting power cables; ISABELLE storage rings; the BNL Tandem accelerator; heavy ion experiments at the Tandem; the High Flux Beam Reactor; medium energy physics; nuclear theory; atomic and applied physics; solid state physics; neutron scattering studies; x-ray scattering studies; solid state theory; defects and disorder in solids; surface physics; the Nationalmore » Synchrotron Light Source ; Chemistry Department; Biology Department; Medical Department; energy sciences; environmental sciences; energy technology programs; National Center for Analysis of Energy Systems; advanced reactor systems; nuclear safety; National Nuclear Data Center; nuclear materials safeguards; Applied Mathematics Department; and support activities. (GHT)« less

Immobilization of Trametes versicolor cultures for improving laccase production in bubble column reactor intensified by sonication.

PubMed

Wang, Feng; Guo, Chen; Liu, Chun-Zhao

2013-01-01

The mycelia of Trametes versicolor immobilized in alginate beads provided higher laccase production than that in pelleted form. An efficient ultrasonic treatment enhanced laccase production from the immobilized T. versicolor cultures. The optimized treatment process consisted of exposing 36-h-old bead cultures to 7-min ultrasonic treatments twice with a 12-h interval using a fixed ultrasonic power and frequency (120 W, 40 kHz). Using the intensification strategy with sonication, laccase production increased by more than 2.1-fold greater than the untreated control in both flasks and bubble column reactors. The enhancement of laccase production by ultrasonic treatment is related to the improved mass transfer of nutrients and product between the liquid medium and the gel matrix. These results provide a basis for the large-scale and highly-efficient production of laccase using sonobioreactors.

Thermionic fast spectrum reactor-converter on the basis of multi-cell TFE

NASA Astrophysics Data System (ADS)

Ponomarev-Stepnoi, N. N.; Kompaniets, G. V.; Poliakov, D. N.; Stepennov, B. S.; Andreev, P. V.; Zhabotinsky, E. E.; Nikolaev, Yu. V.; Lapochkin, N. V.

2001-02-01

Today Russian experts have technological experience in development of in-core thermionic converters for reactors of space nuclear power plants. Such a converter contains nuclear fuel inside and really represents a fuel element of a reactor. Two types of reactors can be considered on the basis of these thermionic fuel elements: with thermal or intermediate neutron spectrum, and with fast neutron spectrum. The first type is characterized by the presence of moderator in core that ensures most economical usage of nuclear fuel. The estimation shows that moderated system is the most effective in the power range of about 5 ... 100 kWe. The power systems of higher level are characterized by larger dimensions due to the presence of moderator. The second type of reactor is considered for higher power levels. This power range is about hundreds kWe. Dimensions of the fast reactor and core configuration are determined by the necessity to ensure the required net output power, on the one hand, and the necessity to ensure critical state on the other hand. In the case of using in-core thermionic fuel elements of the specified design, minimal reactor output power is determined by reactor criticality condition, and maximum reactor power output is determined by specifications and launcher capabilities. In the present paper the effective multiplication factor of a fast spectrum reactor on the basis of a multi-cell TFE developed by ``Lutch'' is considered a function of the total number of TFEs in the reactor. The MCU Monte-Carlo code, developed in Russia (Alekseev, et al., 1991), was used for computations. TFE computational models are placed in the nodes of a uniform triangular lattice and surrounded with pressure vessel and a side reflector. Ordinary fuel pins without thermionic converters were used instead of some TFEs to optimize criticality parameters, dimensions and output power of the reactor. General weight parameters of the reactor are presented in the paper. .

10 CFR 2.1115 - Designation of issues for adjudicatory hearing.

Code of Federal Regulations, 2010 CFR

2010-01-01

... at Civilian Nuclear Power Reactors § 2.1115 Designation of issues for adjudicatory hearing. (a) After... reactor already licensed to operate at the site, or any civilian nuclear power reactor for which a... the issuance of a construction permit or operating license for a civilian nuclear power reactor at...

Effect of impeller type and mechanical agitation on the mass transfer and power consumption aspects of ASBR operation treating synthetic wastewater.

PubMed

Michelan, Rogério; Zimmer, Thiago R; Rodrigues, José A D; Ratusznei, Suzana M; de Moraes, Deovaldo; Zaiat, Marcelo; Foresti, Eugenio

2009-03-01

The effect of flow type and rotor speed was investigated in a round-bottom reactor with 5 L useful volume containing 2.0 L of granular biomass. The reactor treated 2.0 L of synthetic wastewater with a concentration of 800 mgCOD/L in 8-h cycles at 30 degrees C. Five impellers, commonly used in biological processes, have been employed to this end, namely: a turbine and a paddle impeller with six-vertical-flat-blades, a turbine and a paddle impeller with six-45 degrees -inclined-flat-blades and a three-blade-helix impeller. Results showed that altering impeller type and rotor speed did not significantly affect system stability and performance. Average organic matter removal efficiency was about 84% for filtered samples, total volatile acids concentration was below 20 mgHAc/L and bicarbonate alkalinity a little less than 400 mgCaCO3/L for most of the investigated conditions. However, analysis of the first-order kinetic model constants showed that alteration in rotor speed resulted in an increase in the values of the kinetic constants (for instance, from 0.57 h(-1) at 50 rpm to 0.84 h(-1) at 75 rpm when the paddle impeller with six-45 degrees -inclined-flat-blades was used) and that axial flow in mechanically stirred reactors is preferable over radial-flow when the vertical-flat-blade impeller is compared to the inclined-flat-blade impeller (for instance at 75 rpm, from 0.52 h(-1) with the six-flat-blade-paddle impeller to 0.84 h(-1) with the six-45 degrees -inclined-flat-blade-paddle impeller), demonstrating that there is a rotor speed and an impeller type that maximize solid-liquid mass transfer in the reaction medium. Furthermore, power consumption studies in this reduced reactor volume showed that no high power transfer is required to improve mass transfer (less than 0.6 kW/10(3)m3).

78 FR 64028 - Decommissioning of Nuclear Power Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-25

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0035] Decommissioning of Nuclear Power Reactors AGENCY... Commission (NRC) is issuing Revision 1 of regulatory guide (RG) 1.184 ``Decommissioning of Nuclear Power... the NRC's regulations relating to the decommissioning process for nuclear power reactors. The revision...

A Basic LEGO Reactor Design for the Provision of Lunar Surface Power

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

John Darrell Bess

2008-06-01

A final design has been established for a basic Lunar Evolutionary Growth-Optimized (LEGO) Reactor using current and near-term technologies. The LEGO Reactor is a modular, fast-fission, heatpipe-cooled, clustered-reactor system for lunar-surface power generation. The reactor is divided into subcritical units that can be safely launched with lunar shipments from Earth, and then emplaced directly into holes drilled into the lunar regolith to form a critical reactor assembly. The regolith would not just provide radiation shielding, but serve as neutron-reflector material as well. The reactor subunits are to be manufactured using proven and tested materials for use in radiation environments, suchmore » as uranium-dioxide fuel, stainless-steel cladding and structural support, and liquid-sodium heatpipes. The LEGO Reactor system promotes reliability, safety, and ease of manufacture and testing at the cost of an increase in launch mass per overall rated power level and a reduction in neutron economy when compared to a single-reactor system. A single unshielded LEGO Reactor subunit has an estimated mass of approximately 448 kg and provides approximately 5 kWe. The overall envelope for a single subunit with fully extended radiator panels has a height of 8.77 m and a diameter of 0.50 m. Six subunits could provide sufficient power generation throughout the initial stages of establishing a lunar outpost. Portions of the reactor may be neutronically decoupled to allow for reduced power production during unmanned periods of base operations. During later stages of lunar-base development, additional subunits may be emplaced and coupled into the existing LEGO Reactor network, subject to lunar base power demand. Improvements in reactor control methods, fuel form and matrix, shielding, as well as power conversion and heat rejection techniques can help generate an even more competitive LEGO Reactor design. Further modifications in the design could provide power generative opportunities for use on other extraterrestrial surfaces.« less

A new safety channel based on ¹⁷N detection in research reactors.

PubMed

Seyfi, Somayye; Gharib, Morteza

2015-10-01

Tehran research reactor (TRR) is a representative of pool type research reactors using light water, as coolant and moderator. This reactor is chosen as a prototype to demonstrate and prove the feasibility of (17)N detection as a new redundant channel for reactor power measurement. In TRR, similar to other pool type reactors, neutron detectors are immersed in the pool around the core as the main power measuring devices. In the present article, a different approach, using out of water neutron detector, is employed to measure reactor power. This new method is based on (17)O (n,p) (17)N reaction taking place inside the core and subsequent measurement of delayed neutrons emitted due to (17)N disintegration. Count and measurement of neutrons around outlet water pipe provides a reliable redundant safety channel to measure reactor power. Results compared with other established channels indicate a good agreement and shows a linear interdependency with true thermal power. Safety of reactor operation is improved with installation & use of this new power measuring channel. The new approach may equally serve well as a redundant channel in all other types of reactors having coolant comprised of oxygen in its molecular constituents. Contrary to existing channels, this one is totally out of water and thus is an advantage over current instrumentations. It is proposed to employ the same idea on other reactors (nuclear power plants too) to improve safety criteria. Copyright © 2015 Elsevier Ltd. All rights reserved.

Reactor design and integration into a nuclear electric spacecraft

NASA Technical Reports Server (NTRS)

Phillips, W. M.; Koenig, D. R.

1978-01-01

One of the well-defined applications for nuclear power in space is nuclear electric propulsion (NEP). Mission studies have identified the optimum power level (400 kWe). A single Shuttle launch requirement and science-package integration have added additional constraints to the design. A reactor design which will meet these constraints has been studied. The reactor employs 90 fuel elements, each heat pipe cooled. Reactor control is obtained with BeO/B4C drums in a BeO reflector. The balance of the spacecraft is shielded from the reactor with LiH. Power conditioning and reactor control drum drives are located behind the LiH with the power conditioning. Launch safety, mechanical design and integration with the power conversion subsystem are discussed.

Linear Free Energy Correlations for Fission Product Release from the Fukushima-Daiichi Nuclear Accident

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

Abrecht, David G.; Schwantes, Jon M.

This paper extends the preliminary linear free energy correlations for radionuclide release performed by Schwantes, et al., following the Fukushima-Daiichi Nuclear Power Plant accident. Through evaluations of the molar fractionations of radionuclides deposited in the soil relative to modeled radionuclide inventories, we confirm the source of the radionuclides to be from active reactors rather than the spent fuel pool. Linear correlations of the form ln χ = -α (ΔG rxn°(T C))/(RT C)+β were obtained between the deposited concentration and the reduction potential of the fission product oxide species using multiple reduction schemes to calculate ΔG° rxn(T C). These models allowedmore » an estimate of the upper bound for the reactor temperatures of T C between 2130 K and 2220 K, providing insight into the limiting factors to vaporization and release of fission products during the reactor accident. Estimates of the release of medium-lived fission products 90Sr, 121mSn, 147Pm, 144Ce, 152Eu, 154Eu, 155Eu, 151Sm through atmospheric venting and releases during the first month following the accident were performed, and indicate large quantities of 90Sr and radioactive lanthanides were likely to remain in the damaged reactor cores.« less

Reactor-Scale Cultivation of the Hyperthermophilic Methanarchaeon Methanococcus jannaschii to High Cell Densities

PubMed Central

Mukhopadhyay, Biswarup; Johnson, Eric F.; Wolfe, Ralph S.

1999-01-01

For the hyperthermophilic and barophilic methanarchaeon Methanococcus jannaschii, we have developed a medium and protocols for reactor-scale cultivation that improved the final cell yield per liter from ∼0.5 to ∼7.5 g of packed wet cells (∼1.8 g dry cell mass) under autotrophic growth conditions and to ∼8.5 g of packed wet cells (∼2 g dry cell mass) with yeast extract (2 g liter−1) and tryptone (2 g liter−1) as medium supplements. For growth in a sealed bottle it was necessary to add Se to the medium, and a level of 2 μM for added Se gave the highest final cell yield. In a reactor M. jannaschii grew without added Se in the medium; it is plausible that the cells received Se as a contaminant from the reactor vessel and the H2S supply. But, for the optimal performance of a reactor culture, an addition of Se to a final concentration of 50 to 100 μM was needed. Also, cell growth in a reactor culture was inhibited at much higher Se concentrations. These observations and the data from previous work with methanogen cell extracts (B. C. McBride and R. S. Wolfe, Biochemistry 10:4312–4317, 1971) suggested that from a continuously sparged reactor culture Se was lost in the exhaust gas as volatile selenides, and this loss raised the apparent required level of and tolerance for Se. In spite of having a proteinaceous cell wall, M. jannaschii withstood an impeller tip speed of 235.5 cms−1, which was optimal for achieving high cell density and also was the higher limit for the tolerated shear rate. The organism secreted one or more acidic compounds, which lowered pH in cultures without pH control; this secretion continued even after cessation of growth. PMID:10543823

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.

10 CFR 50.36a - Technical specifications on effluents from nuclear power reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... reactors. 50.36a Section 50.36a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND...; Ineligibility of Certain Applicants § 50.36a Technical specifications on effluents from nuclear power reactors..., including expected occurrences, as low as is reasonably achievable, each licensee of a nuclear power reactor...

10 CFR 50.36a - Technical specifications on effluents from nuclear power reactors.

Code of Federal Regulations, 2010 CFR

2010-01-01

... reactors. 50.36a Section 50.36a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND...; Ineligibility of Certain Applicants § 50.36a Technical specifications on effluents from nuclear power reactors..., including expected occurrences, as low as is reasonably achievable, each licensee of a nuclear power reactor...

10 CFR 50.36a - Technical specifications on effluents from nuclear power reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

... reactors. 50.36a Section 50.36a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND...; Ineligibility of Certain Applicants § 50.36a Technical specifications on effluents from nuclear power reactors..., including expected occurrences, as low as is reasonably achievable, each licensee of a nuclear power reactor...

Reactor for simulation and acceleration of solar ultraviolet damage

NASA Technical Reports Server (NTRS)

Laue, E.; Gupta, A.

1979-01-01

An environmental test chamber providing acceleration of UV radiation and precise temperature control (+ or -)1 C was designed, constructed and tested. This chamber allows acceleration of solar ultraviolet up to 30 suns while maintaining temperature of the absorbing surface at 30 C - 60 C. This test chamber utilizes a filtered medium pressure mercury arc as the source of radiation, and a combination of selenium radiometer and silicon radiometer to monitor solar ultraviolet (295-340 nm) and total radiant power output, respectively. Details of design and construction and operational procedures are presented along with typical test data.

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.

CHARACTERISTIC QUALITIES OF SOME ATOMIC POWER STATIONS (in Hungarian)

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

Ligeti, G.

1962-04-01

Mostly as the result of economic factors, the current rate of construction of public atomic power stations has slowed down. The use of atomic energy is considered economical only in a few special cases, such as ship propulsion or supplying power to remote regions. For this reason, many reactors were designed especially for the construction of such midget'' power stations, operating at power levels ranging from 10 to 70 Mw. Technical details are given of such already-built or proposed systems, including the following: pressurized- water reactors such as the Babcock and Wilcox 60-Mw reactor, using 2.4% U/sup 235/ fuel; themore » Humphrey-Glasow Company's 20 Mw reactor; the gascooled system of the de Havilland Company; the organicmoderated reactor of the English Electric Company; the organic-moderated system of the Hawker-Siddeley Nuclear Power Company; the boiling-water reactor of the Mitchell Engineering Company and the steam-cooled, heavy-water reactor of the Rolls-Royce & Vickers Company. (TTT)« less

Test Results from a Direct Drive Gas Reactor Simulator Coupled to a Brayton Power Conversion Unit

NASA Technical Reports Server (NTRS)

Hervol, David S.; Briggs, Maxwell H.; Owen, Albert K.; Bragg-Sitton, Shannon M.; Godfroy, Thomas J.

2010-01-01

Component level testing of power conversion units proposed for use in fission surface power systems has typically been done using relatively simple electric heaters for thermal input. These heaters do not adequately represent the geometry or response of proposed reactors. As testing of fission surface power systems transitions from the component level to the system level it becomes necessary to more accurately replicate these reactors using reactor simulators. The Direct Drive Gas-Brayton Power Conversion Unit test activity at the NASA Glenn Research Center integrates a reactor simulator with an existing Brayton test rig. The response of the reactor simulator to a change in Brayton shaft speed is shown as well as the response of the Brayton to an insertion of reactivity, corresponding to a drum reconfiguration. The lessons learned from these tests can be used to improve the design of future reactor simulators which can be used in system level fission surface power tests.

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

NASA Astrophysics Data System (ADS)

Ilham, Muhammad; Su'ud, Zaki

2017-01-01

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

10 CFR 171.3 - Scope.

Code of Federal Regulations, 2010 CFR

2010-01-01

... holding an operating license for a power reactor, test reactor or research reactor issued under part 50 of... authorizes operation of a power reactor. The regulations in this part also apply to any person holding a...

SL-1 Accident Briefing Report - 1961 Nuclear Reactor Meltdown Educational Documentary

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

None

2013-09-25

U.S. Atomic Energy Commission (Idaho Operations Office) briefing about the SL-1 Nuclear Reactor Meltdown. The SL-1, or Stationary Low-Power Reactor Number One, was a United States Army experimental nuclear power reactor which underwent a steam explosion and meltdown on January 3, 1961, killing its three operators. The direct cause was the improper withdrawal of the central control rod, responsible for absorbing neutrons in the reactor core. The event is the only known fatal reactor accident in the United States. The accident released about 80 curies (3.0 TBq) of Iodine-131, which was not considered significant due to its location in amore » remote desert of Idaho. About 1,100 curies (41 TBq) of fission products were released into the atmosphere. The facility, located at the National Reactor Testing Station approximately 40 miles (64 km) west of Idaho Falls, Idaho, was part of the Army Nuclear Power Program and was known as the Argonne Low Power Reactor (ALPR) during its design and build phase. It was intended to provide electrical power and heat for small, remote military facilities, such as radar sites near the Arctic Circle, and those in the DEW Line. The design power was 3 MW (thermal). Operating power was 200 kW electrical and 400 kW thermal for space heating. In the accident, the core power level reached nearly 20 GW in just four milliseconds, precipitating the reactor accident and steam explosion.« less

SL-1 Accident Briefing Report - 1961 Nuclear Reactor Meltdown Educational Documentary

ScienceCinema

None

2018-01-16

U.S. Atomic Energy Commission (Idaho Operations Office) briefing about the SL-1 Nuclear Reactor Meltdown. The SL-1, or Stationary Low-Power Reactor Number One, was a United States Army experimental nuclear power reactor which underwent a steam explosion and meltdown on January 3, 1961, killing its three operators. The direct cause was the improper withdrawal of the central control rod, responsible for absorbing neutrons in the reactor core. The event is the only known fatal reactor accident in the United States. The accident released about 80 curies (3.0 TBq) of Iodine-131, which was not considered significant due to its location in a remote desert of Idaho. About 1,100 curies (41 TBq) of fission products were released into the atmosphere. The facility, located at the National Reactor Testing Station approximately 40 miles (64 km) west of Idaho Falls, Idaho, was part of the Army Nuclear Power Program and was known as the Argonne Low Power Reactor (ALPR) during its design and build phase. It was intended to provide electrical power and heat for small, remote military facilities, such as radar sites near the Arctic Circle, and those in the DEW Line. The design power was 3 MW (thermal). Operating power was 200 kW electrical and 400 kW thermal for space heating. In the accident, the core power level reached nearly 20 GW in just four milliseconds, precipitating the reactor accident and steam explosion.

Neutron fluxes in test reactors

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

Youinou, Gilles Jean-Michel

Communicate the fact that high-power water-cooled test reactors such as the Advanced Test Reactor (ATR), the High Flux Isotope Reactor (HFIR) or the Jules Horowitz Reactor (JHR) cannot provide fast flux levels as high as sodium-cooled fast test reactors. The memo first presents some basics physics considerations about neutron fluxes in test reactors and then uses ATR, HFIR and JHR as an illustration of the performance of modern high-power water-cooled test reactors.

Bio-reactor chamber

NASA Technical Reports Server (NTRS)

Chandler, Joseph A. (Inventor)

1989-01-01

A bioreactor for cell culture is disclosed which provides for the introduction of fresh medium without excessive turbulent action. The fresh medium enters the bioreactor through a filter with a backwash action which prevents the cells from settling on the filter. The bioreactor is sealed and depleted medium is forced out of the container as fresh medium is added.

Influence of power supply on the generation of ozone and degradation of phenol in a surface discharge reactor

NASA Astrophysics Data System (ADS)

Zhao, Yan; Shang, Kefeng; Duan, Lijuan; Li, Yue; An, Jiutao; Zhang, Chunyang; Lu, Na; Li, Jie; Wu, Yan

2013-03-01

A surface Dielectric Barrier Discharge (DBD) reactor was utilized to degrade phenol in water. Different power supplies applied to the DBD reactor affect the discharge modes, the formation of chemically active species and thus the removal efficiency of pollutants. It is thus important to select an optimized power supply for the DBD reactor. In this paper, the influence of the types of power supplies including alternate current (AC) and bipolar pulsed power supply on the ozone generation in a surface discharge reactor was measured. It was found that compared with bipolar pulsed power supply, higher energy efficiency of O3 generation was obtained when DBD reactor was supplied with 50Hz AC power supply. The highest O3 generation was approximate 4 mg kJ-1 moreover, COD removal efficiency of phenol wastewater reached 52.3% after 3 h treatment under an AC peak voltage of 2.6 kV.

Thermionic reactors for space nuclear power

NASA Technical Reports Server (NTRS)

Homeyer, W. G.; Merrill, M. H.; Holland, J. W.; Fisher, C. R.; Allen, D. T.

1985-01-01

Thermionic reactor designs for a variety of space power applications spanning the range from 5 kWe to 3 MWe are described. In all of these reactors, nuclear heat is converted directly to electrical energy in thermionic fuel elements (TFEs). A circulating reactor coolant carries heat from the core of TFEs directly to a heat rejection radiator system. The recent design of a thermionic reactor to meet the SP-100 requirements is emphasized. Design studies of reactors at other power levels show that the same TFE can be used over a broad range in power, and that design modifications can extend the range to many megawatts. The design of the SP-100 TFE is similar to that of TFEs operated successfully in test reactors, but with design improvements to extend the operating lifetime to seven years.

78 FR 73898 - Operator Licensing Examination Standards for Power Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-09

... Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Draft NUREG; request for comment. SUMMARY: The U.S..., Revision 10, ``Operator Licensing Examination Standards for Power Reactors.'' DATES: Submit comments [email protected] . Both of the Office of New Reactors; or Timothy Kolb, Office of Nuclear Reactor Regulation, U...

MC 2 -3: Multigroup Cross Section Generation Code for Fast Reactor Analysis

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

Lee, Changho; Yang, Won Sik

This paper presents the methods and performance of the MC2 -3 code, which is a multigroup cross-section generation code for fast reactor analysis, developed to improve the resonance self-shielding and spectrum calculation methods of MC2 -2 and to simplify the current multistep schemes generating region-dependent broad-group cross sections. Using the basic neutron data from ENDF/B data files, MC2 -3 solves the consistent P1 multigroup transport equation to determine the fundamental mode spectra for use in generating multigroup neutron cross sections. A homogeneous medium or a heterogeneous slab or cylindrical unit cell problem is solved in ultrafine (2082) or hyperfine (~400more » 000) group levels. In the resolved resonance range, pointwise cross sections are reconstructed with Doppler broadening at specified temperatures. The pointwise cross sections are directly used in the hyperfine group calculation, whereas for the ultrafine group calculation, self-shielded cross sections are prepared by numerical integration of the pointwise cross sections based upon the narrow resonance approximation. For both the hyperfine and ultrafine group calculations, unresolved resonances are self-shielded using the analytic resonance integral method. The ultrafine group calculation can also be performed for a two-dimensional whole-core problem to generate region-dependent broad-group cross sections. Verification tests have been performed using the benchmark problems for various fast critical experiments including Los Alamos National Laboratory critical assemblies; Zero-Power Reactor, Zero-Power Physics Reactor, and Bundesamt für Strahlenschutz experiments; Monju start-up core; and Advanced Burner Test Reactor. Verification and validation results with ENDF/B-VII.0 data indicated that eigenvalues from MC2 -3/DIF3D agreed well with Monte Carlo N-Particle5 MCNP5 or VIM Monte Carlo solutions within 200 pcm and regionwise one-group fluxes were in good agreement with Monte Carlo solutions.« less

Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR)

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

A. Joseph Palmer; Gerry L. McCormick; Shannon J. Corrigan

2010-06-01

2010 International Congress on Advances in Nuclear Power Plants (ICAPP’10) ANS Annual Meeting Imbedded Topical San Diego, CA June 13 – 17, 2010 Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR) Author: A. Joseph Palmer, Mechanical Engineer, Irradiation Test Programs, 208-526-8700, Joe.Palmer@INL.gov Affiliation: Idaho National Laboratory P.O. Box 1625, MS-3840 Idaho Falls, ID 83415 INL/CON-10-17680 ABSTRACT Most test reactors are equipped with shuttle facilities (sometimes called rabbit tubes) whereby small capsules can be inserted into the reactor and retrieved during power operations. With the installation of Hydraulic Shuttle Irradiation System (HSIS) this capability has beenmore » restored to the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The general design and operating principles of this system were patterned after the hydraulic rabbit at Oak Ridge National Laboratory’s (ORNL) High Flux Isotope Reactor (HFIR), which has operated successfully for many years. Using primary coolant as the motive medium the HSIS system is designed to simultaneously transport fourteen shuttle capsules, each 16 mm OD x 57 mm long, to and from the B-7 position of the reactor. The B-7 position is one of the higher flux positions in the reactor with typical thermal and fast (>1 Mev) fluxes of 2.8E+14 n/cm2/sec and 1.9E+14 n/cm2/sec respectively. The available space inside each shuttle is approximately 14 mm diameter x 50 mm long. The shuttle containers are made from titanium which was selected for its low neutron activation properties and durability. Shuttles can be irradiated for time periods ranging from a few minutes to several months. The Send and Receive Station (SRS) for the HSIS is located 2.5 m deep in the ATR canal which allows irradiated shuttles to be easily moved from the SRS to a wet loaded cask, or transport pig. The HSIS system first irradiated (empty) shuttles in September 2009 and has since completed a Readiness Assessment in November 2009. The HSIS is a key component of the ATR National Scientific User Facility (NSUF) operated by Battelle Energy Alliance, LLC and is available to a wide variety of university researchers for nuclear fuels and materials experiments as well as medical isotope research and production.« less

77 FR 74697 - Meeting of the ACRS, Subcommittee on U.S. Evolutionary Power Reactor; Notice of Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2012-12-17

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS, Subcommittee on U.S. Evolutionary Power Reactor; Notice of Meeting The ACRS Subcommittee on U.S. Evolutionary Power Reactor (U.S. EPR) will hold a meeting on January 17, 2013, Room T-2B1, 11545 Rockville Pike...

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

Fitzpatrick, F.C.; Gray, D.D.; Hyndman, J.R.

The thermal, ecological, and social impacts of a 40-reactor NEC are compared to impacts from four 10-reactor NECs and ten 4-reactor power plants. The comparison was made for surrogate sites in western Tennessee. The surrogate site for the 40-reactor NEC is located on Kentucky Lake. A layout is postulated for ten clusters of four reactors each with 2.5-mile spacing between clusters. The plants use natural-draft cooling towers. A transmission system is proposed for delivering the power (48,000 MW) to five load centers. Comparable transmission systems are proposed for the 10-reactor NECs and the 4-reactor dispersed sites delivering power to themore » same load centers. (auth)« less

Space nuclear power systems; Proceedings of the 8th Symposium, Albuquerque, NM, Jan. 6-10, 1991. Pts. 1-3

NASA Technical Reports Server (NTRS)

El-Genk, Mohamed S. (Editor); Hoover, Mark D. (Editor)

1991-01-01

The present conference discusses NASA mission planning for space nuclear power, lunar mission design based on nuclear thermal rockets, inertial-electrostatic confinement fusion for space power, nuclear risk analysis of the Ulysses mission, the role of the interface in refractory metal alloy composites, an advanced thermionic reactor systems design code, and space high power nuclear-pumped lasers. Also discussed are exploration mission enhancements with power-beaming, power requirement estimates for a nuclear-powered manned Mars rover, SP-100 reactor design, safety, and testing, materials compatibility issues for fabric composite radiators, application of the enabler to nuclear electric propulsion, orbit-transfer with TOPAZ-type power sources, the thermoelectric properties of alloys, ruthenium silicide as a promising thermoelectric material, and innovative space-saving device for high-temperature piping systems. The second volume of this conference discusses engine concepts for nuclear electric propulsion, nuclear technologies for human exploration of the solar system, dynamic energy conversion, direct nuclear propulsion, thermionic conversion technology, reactor and power system control, thermal management, thermionic research, effects of radiation on electronics, heat-pipe technology, radioisotope power systems, and nuclear fuels for power reactors. The third volume discusses space power electronics, space nuclear fuels for propulsion reactors, power systems concepts, space power electronics systems, the use of artificial intelligence in space, flight qualifications and testing, microgravity two-phase flow, reactor manufacturing and processing, and space and environmental effects.

Geoantineutrino spectrum and slow nuclear burning on the boundary of the liquid and solid phases of the Earth's core

NASA Astrophysics Data System (ADS)

Rusov, V. D.; Pavlovich, V. N.; Vaschenko, V. N.; Tarasov, V. A.; Zelentsova, T. N.; Bolshakov, V. N.; Litvinov, D. A.; Kosenko, S. I.; Byegunova, O. A.

2007-09-01

We give an alternative description of the data produced in the KamLAND experiment. Assuming the existence of a natural nuclear reactor on the boundary of the liquid and solid phases of the Earth's core, a geoantineutrino spectrum is obtained. This assumption is based on the experimental results of V. Anisichkin and his collaborators on the interaction of uranium dioxide and uranium carbide with iron-nickel and silica-alumina melts at high pressure (5-10 GPa) and temperature (1600-2200°C), which led to the proposal of the existence of an actinide shell in the Earth's core. We describe the operating mechanism of this reactor as solitary waves of nuclear burning in 238U and/or 232Th medium, in particular, as neutron fission progressive waves of Feoktistov and/or Teller et al. type. Next, we propose a simplified model for the accumulation and burn-up kinetics in Feoktistov's U-Pu fuel cycle. We also apply this model for numerical simulations of neutron fission wave in a two-phase UO2/Fe medium on the surface of the Earth's solid core. The proposed georeactor model offers a mechanism for the generation of 3He. The 3He/4He distribution in the Earth's interior is calculated, which in turn can be used as a natural quantitative criterion of the georeactor thermal power. Finally, we give a tentative estimation of the geoantineutrino intensity and spectrum on the Earth's surface. For this purpose we use the O'Nions et al. geochemical model of mantle differentiation and crust growth complemented by a nuclear energy source (georeactor with power of 30 TW).

Formulation and experimental evaluation of closed-form control laws for the rapid maneuvering of reactor neutronic power

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

Bernard, J.A.

1989-09-01

This report describes both the theoretical development and the experimental evaluation of a novel, robust methodology for the time-optimal adjustment of a reactor's neutronic power under conditions of closed-loop digital control. Central to the approach are the MIT-SNL Period-Generated Minimum Time Control Laws' which determine the rate at which reactivity should be changed in order to cause a reactor's neutronic power to conform to a specified trajectory. Using these laws, reactor power can be safely raised by five to seven orders of magnitude in a few seconds. The MIT-SNL laws were developed to facilitate rapid increases of neutronic power onmore » spacecraft reactors operating in an SDI environment. However, these laws are generic and have other applications including the rapid recovery of research and test reactors subsequent to an unanticipated shutdown, power increases following the achievement of criticality on commercial reactors, power adjustments on commercial reactors so as to minimize thermal stress, and automated startups. The work reported here was performed by the Massachusetts Institute of Technology under contract to the Sandia National Laboratories. Support was also provided by the US Department of Energy's Division of University and Industry Programs. The work described in this report is significant in that a novel solution to the problem of time-optimal control of neutronic power was identified, in that a rigorous description of a reactor's dynamics was derived in that the rate of change of reactivity was recognized as the proper control signal, and in that extensive experimental trials were conducted of these newly developed concepts on actual nuclear reactors. 43 refs., 118 figs., 11 tabs.« less

Fission-powered in-core thermoacoustic sensor

DOE PAGES

Garrett, Steven L.; Smith, James A.; Smith, Robert W. M.; ...

2016-04-07

A thermoacoustic engine is operated within the core of a nuclear reactor to acoustically telemeter coolant temperature (frequency-encoded) and reactor power level (amplitude-encoded) outside the reactor, thus providing the values of these important parameters without external electrical power or wiring. We present data from two hydrophones in the coolant (far from the core) and an accelerometer attached to a structure outside the reactor. Furthermore, these signals have been detected even in the presence of substantial background noise generated by the reactor's fluid pumps.

Fission-powered in-core thermoacoustic sensor

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

Garrett, Steven L.; Smith, James A.; Smith, Robert W. M.

2016-04-04

A thermoacoustic engine is operated within the core of a nuclear reactor to acoustically telemeter coolant temperature (frequency-encoded) and reactor power level (amplitude-encoded) outside the reactor, thus providing the values of these important parameters without external electrical power or wiring. We present data from two hydrophones in the coolant (far from the core) and an accelerometer attached to a structure outside the reactor. These signals have been detected even in the presence of substantial background noise generated by the reactor's fluid pumps.

Irradiation Tests Supporting LEU Conversion of Very High Power Research Reactors in the US

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

Woolstenhulme, N. E.; Cole, J. I.; Glagolenko, I.

The US fuel development team is developing a high density uranium-molybdenum alloy monolithic fuel to enable conversion of five high-power research reactors. Previous irradiation tests have demonstrated promising behavior for this fuel design. A series of future irradiation tests will enable selection of final fuel fabrication process and provide data to qualify the fuel at moderately-high power conditions for use in three of these five reactors. The remaining two reactors, namely the Advanced Test Reactor and High Flux Isotope Reactor, require additional irradiation tests to develop and demonstrate the fuel’s performance with even higher power conditions, complex design features, andmore » other unique conditions. This paper reviews the program’s current irradiation testing plans for these moderately-high irradiation conditions and presents conceptual testing strategies to illustrate how subsequent irradiation tests will build upon this initial data package to enable conversion of these two very-high power research reactors.« less

Alternative approaches to fusion. [reactor design and reactor physics for Tokamak fusion reactors

NASA Technical Reports Server (NTRS)

Roth, R. J.

1976-01-01

The limitations of the Tokamak fusion reactor concept are discussed and various other fusion reactor concepts are considered that employ the containment of thermonuclear plasmas by magnetic fields (i.e., stellarators). Progress made in the containment of plasmas in toroidal devices is reported. Reactor design concepts are illustrated. The possibility of using fusion reactors as a power source in interplanetary space travel and electric power plants is briefly examined.

Mini-cavity plasma core reactors for dual-mode space nuclear power/propulsion systems. M.S. Thesis

NASA Technical Reports Server (NTRS)

Chow, S.

1976-01-01

A mini-cavity plasma core reactor is investigated for potential use in a dual-mode space power and propulsion system. In the propulsive mode, hydrogen propellant is injected radially inward through the reactor solid regions and into the cavity. The propellant is heated by both solid driver fuel elements surrounding the cavity and uranium plasma before it is exhausted out the nozzle. The propellant only removes a fraction of the driver power, the remainder is transferred by a coolant fluid to a power conversion system, which incorporates a radiator for heat rejection. Neutronic feasibility of dual mode operation and smaller reactor sizes than those previously investigated are shown to be possible. A heat transfer analysis of one such reactor shows that the dual-mode concept is applicable when power generation mode thermal power levels are within the same order of magnitude as direct thrust mode thermal power levels.

Thermionic reactor power conditioner design for nuclear electric propulsion.

NASA Technical Reports Server (NTRS)

Jacobsen, A. S.; Tasca, D. M.

1971-01-01

Consideration of the effects of various thermionic reactor parameters and requirements upon spacecraft power conditioning design. A basic spacecraft is defined using nuclear electric propulsion, requiring approximately 120 kWe. The interrelationships of reactor operating characteristics and power conditioning requirements are discussed and evaluated, and the effects on power conditioner design and performance are presented.

Preliminary plan for testing a thermionic reactor in the Plum Brook Space Power Facility

NASA Technical Reports Server (NTRS)

Haley, F. A.

1972-01-01

A preliminary plan is presented for testing a thermionic reactor in the Plum Brook Space Power Facility (SPF). A technical approach, cost estimate, manpower estimate, and schedule are presented to cover a 2 year full power reactor test.

10 CFR 2.1105 - Definitions.

Code of Federal Regulations, 2014 CFR

2014-01-01

... Nuclear Fuel Storage Capacity at Civilian Nuclear Power Reactors § 2.1105 Definitions. As used in this part: (a) Civilian nuclear power reactor means a civilian nuclear power plant required to be licensed... nuclear fuel means fuel that has been withdrawn from a nuclear reactor following irradiation, the...

10 CFR 2.1105 - Definitions.

Code of Federal Regulations, 2013 CFR

2013-01-01

... Nuclear Fuel Storage Capacity at Civilian Nuclear Power Reactors § 2.1105 Definitions. As used in this part: (a) Civilian nuclear power reactor means a civilian nuclear power plant required to be licensed... nuclear fuel means fuel that has been withdrawn from a nuclear reactor following irradiation, the...

Fuel supply of nuclear power industry with the introduction of fast reactors

NASA Astrophysics Data System (ADS)

Muraviev, E. V.

2014-12-01

The results of studies conducted for the validation of the updated development strategy for nuclear power industry in Russia in the 21st century are presented. Scenarios with different options for the reprocessing of spent fuel of thermal reactors and large-scale growth of nuclear power industry based on fast reactors of inherent safety with a breeding ratio of ˜1 in a closed nuclear fuel cycle are considered. The possibility of enhanced fuel breeding in fast reactors is also taken into account in the analysis. The potential to establish a large-scale nuclear power industry that covers 100% of the increase in electric power requirements in Russia is demonstrated. This power industry may be built by the end of the century through the introduction of fast reactors (replacing thermal ones) with a gross uranium consumption of up to ˜1 million t and the termination of uranium mining even if the reprocessing of spent fuel of thermal reactors is stopped or suffers a long-term delay.

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

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

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

1991-09-01

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

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

NASA Astrophysics Data System (ADS)

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

2018-01-01

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

An adaptive load-following control system for a space nuclear power system

NASA Astrophysics Data System (ADS)

Metzger, John D.; El-Genk, Mohamed S.

An adaptive load-following control system is proposed for a space nuclear power system. The conceptual design of the SP-100 space nuclear power system proposes operating the nuclear reactor at a base thermal power and accommodating changes in the electrical power demand with a shunt regulator. It is necessary to increase the reactor thermal power if the payload electrical demand exceeds the peak system electrical output for the associated reactor power. When it is necessary to change the nuclear reactor power to meet a change in the power demand, the power ascension or descension must be accomplished in a predetermined manner to avoid thermal stresses in the system and to achieve the desired reactor period. The load-following control system described has the ability to adapt to changes in the system and to changes in the satellite environment. The application is proposed of the model reference adaptive control (MRAC). The adaptive control system has the ability to control the dynamic response of nonlinear systems. Three basic subsets of adaptive control are: (1) gain scheduling, (2) self-tuning regulators, and (3) model reference adaptive control.

Evaluation of an Innovative Approach to Validation of ...

EPA Pesticide Factsheets

UV disinfection is an effective process for inactivating many microbial pathogens found in source waters with the potential as stand-alone treatment or in combination with other disinfectants. For surface and groundwater sourced drinking water applications, the U.S. Environmental Protection Agency (USEPA) provided guidance on the validation of UV reactors nearly a decade ago. The focus of the guidance was primarily for inactivation of Cryptosporidium and Giardia. Over the last ten years many lessons have been learned, validation practices have been modified, new science issues discovered, and changes in operation & monitoring of UV systems need to be addressed. Also, there remains no standard approach for validating UV reactors to meet a 4-log (99.99%) inactivation of viruses. USEPA in partnership with the Cadmus Group, Carollo Engineers, and other State & Industry collaborators, are evaluating new approaches for validating UV reactors to meet groundwater & surface water pathogen inactivation including viruses for low-pressure and medium-pressure UV systems. A particular challenge for medium-pressure UV is the monitoring of low-wavelength germicidal contributions for appropriate crediting of disinfection under varying reactor conditions of quartz sleeve fouling, lamp aging, and changes in UV absorbance of the water over time. In the current effort, bench and full-scale studies are being conducted on a low pressure (LP) UV reactor and a medium pressure (MP) UV re

The Angra Project: Monitoring Nuclear Reactors with Antineutrino Detectors

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

Anjos, J. C.; Barbosa, A. F.; Lima, H. P. Jr.

2010-03-30

We present the status of the Angra Neutrino project, describing the development of an antineutrino detector aimed at monitoring nuclear reactor activity. The experiment will take place at the Brazilian nuclear power plant located in Angra dos Reis. The Angra II reactor, with 4 GW of thermal power, will be used as a source of antineutrinos. A water Cherenkov detector will be placed above ground in a commercial container outside the reactor containment, about 30 m from the reactor core. With a detector of one ton scale a few thousand antineutrino interactions per day are expected. We intend, in amore » first step, to use the measured neutrino event rate to monitor the on--off status and the thermal power delivered by the reactor. In addition to the safeguards issues the project will provide an alternative tool to have an independent measurement of the reactor power.« less

The Angra Project: Monitoring Nuclear Reactors with Antineutrino Detectors

NASA Astrophysics Data System (ADS)

Anjos, J. C.; Barbosa, A. F.; Bezerra, T. J. C.; Chimenti, P.; Gonzalez, L. F. G.; Kemp, E.; de Oliveira, M. A. Leigui; Lima, H. P.; Lima, R. M.; Nunokawa, H.

2010-03-01

We present the status of the Angra Neutrino project, describing the development of an antineutrino detector aimed at monitoring nuclear reactor activity. The experiment will take place at the Brazilian nuclear power plant located in Angra dos Reis. The Angra II reactor, with 4 GW of thermal power, will be used as a source of antineutrinos. A water Cherenkov detector will be placed above ground in a commercial container outside the reactor containment, about 30 m from the reactor core. With a detector of one ton scale a few thousand antineutrino interactions per day are expected. We intend, in a first step, to use the measured neutrino event rate to monitor the on—off status and the thermal power delivered by the reactor. In addition to the safeguards issues the project will provide an alternative tool to have an independent measurement of the reactor power.

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) - as an advanced and promising reactor system that offers solutions to the above problems. The difference (not confrontation) between the approaches to nuclear power development based on the principles of “inherent safety” and “natural safety” is demonstrated.

5 CFR 5801.102 - Prohibited securities.

Code of Federal Regulations, 2014 CFR

2014-01-01

... licenses for facilities which generate electric energy by means of a nuclear reactor; (2) State or local... reactor or a low-level waste facility; (3) Entities manufacturing or selling nuclear power or test reactors; (4) Architectural-engineering companies providing services relating to a nuclear power reactor...

5 CFR 5801.102 - Prohibited securities.

Code of Federal Regulations, 2010 CFR

2010-01-01

... licenses for facilities which generate electric energy by means of a nuclear reactor; (2) State or local... reactor or a low-level waste facility; (3) Entities manufacturing or selling nuclear power or test reactors; (4) Architectural-engineering companies providing services relating to a nuclear power reactor...

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.

A facility for testing 10 to 100-kWe space power reactors

NASA Astrophysics Data System (ADS)

Carlson, William F.; Bitten, Ernest J.

1993-01-01

This paper describes an existing facility that could be used in a cost-effective manner to test space power reactors in the 10 to 100-kWe range before launch. The facility has been designed to conduct full power tests of 100-kWe SP-100 reactor systems and already has the structural features that would be required for lower power testing. The paper describes a reasonable scenario starting with the acceptance at the test site of the unfueled reactor assembly and the separately shipped nuclear fuel. After fueling the reactor and installing it in the facility, cold critical tests are performed, and the reactor is then shipped to the launch site. The availability of this facility represents a cost-effective means of performing the required prelaunch test program.

Space nuclear power systems; Proceedings of the 8th Symposium, Albuquerque, NM, Jan. 6-10, 1991. Pts. 1-3

NASA Astrophysics Data System (ADS)

El-Genk, Mohamed S.; Hoover, Mark D.

1991-07-01

The present conference discusses NASA mission planning for space nuclear power, lunar mission design based on nuclear thermal rockets, inertial-electrostatic confinement fusion for space power, nuclear risk analysis of the Ulysses mission, the role of the interface in refractory metal alloy composites, an advanced thermionic reactor systems design code, and space high power nuclear-pumped lasers. Also discussed are exploration mission enhancements with power-beaming, power requirement estimates for a nuclear-powered manned Mars rover, SP-100 reactor design, safety, and testing, materials compatibility issues for fabric composite radiators, application of the enabler to nuclear electric propulsion, orbit-transfer with TOPAZ-type power sources, the thermoelectric properties of alloys, ruthenium silicide as a promising thermoelectric material, and innovative space-saving device for high-temperature piping systems. The second volume of this conference discusses engine concepts for nuclear electric propulsion, nuclear technologies for human exploration of the solar system, dynamic energy conversion, direct nuclear propulsion, thermionic conversion technology, reactor and power system control, thermal management, thermionic research, effects of radiation on electronics, heat-pipe technology, radioisotope power systems, and nuclear fuels for power reactors. The third volume discusses space power electronics, space nuclear fuels for propulsion reactors, power systems concepts, space power electronics systems, the use of artificial intelligence in space, flight qualifications and testing, microgravity two-phase flow, reactor manufacturing and processing, and space and environmental effects. (For individual items see A93-13752 to A93-13937)

10 CFR Appendix N to Part 52 - Standardization of Nuclear Power Plant Designs: Combined Licenses To Construct and Operate...

Code of Federal Regulations, 2011 CFR

2011-01-01

... Licenses To Construct and Operate Nuclear Power Reactors of Identical Design at Multiple Sites N Appendix N... Designs: Combined Licenses To Construct and Operate Nuclear Power Reactors of Identical Design at Multiple... construct and operate nuclear power reactors of identical design (“common design”) to be located at multiple...

USSR Report, Energy, No. 147.

DTIC Science & Technology

1983-05-18

based on low-temperature reactors ; atomic heat and electric power stations (ATETs); The restructuring of the energy balance for the 1980-2000 period...ASPT) based on low-temperature reactors ; atomic heat and electric power stations (TETs); industrial atomic power stations (AETS) based on high-temper...ature reactors ) and high-efficiency long-distance heat transport (in conjunc- tion with high-temperature nuclear power sources: ASDT). The

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.

Space Nuclear Power Plant Pre-Conceptual Design Report, For Information

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

B. Levine

2006-01-27

This letter transmits, for information, the Project Prometheus Space Nuclear Power Plant (SNPP) Pre-Conceptual Design Report completed by the Naval Reactors Prime Contractor Team (NRPCT). This report documents the work pertaining to the Reactor Module, which includes integration of the space nuclear reactor with the reactor radiation shield, energy conversion, and instrumentation and control segments. This document also describes integration of the Reactor Module with the Heat Rejection segment, the Power Conditioning and Distribution subsystem (which comprise the SNPP), and the remainder of the Prometheus spaceship.

Comparative study between single core model and detail core model of CFD modelling on reactor core cooling behaviour

NASA Astrophysics Data System (ADS)

Darmawan, R.

2018-01-01

Nuclear power industry is facing uncertainties since the occurrence of the unfortunate accident at Fukushima Daiichi Nuclear Power Plant. The issue of nuclear power plant safety becomes the major hindrance in the planning of nuclear power program for new build countries. Thus, the understanding of the behaviour of reactor system is very important to ensure the continuous development and improvement on reactor safety. Throughout the development of nuclear reactor technology, investigation and analysis on reactor safety have gone through several phases. In the early days, analytical and experimental methods were employed. For the last four decades 1D system level codes were widely used. The continuous development of nuclear reactor technology has brought about more complex system and processes of nuclear reactor operation. More detailed dimensional simulation codes are needed to assess these new reactors. Recently, 2D and 3D system level codes such as CFD are being explored. This paper discusses a comparative study on two different approaches of CFD modelling on reactor core cooling behaviour.

76 FR 78173 - Options for Developing the Regulatory Basis for Streamlining Non-Power Reactor License Renewal...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-12-16

... Emergency Preparedness AGENCY: Nuclear Regulatory Commission. ACTION: Notice of public meeting. SUMMARY: The... non-power reactor license renewal and non-power reactor emergency preparedness. This meeting is a... potential enhancements to emergency preparedness requirements. This meeting is open to the public. DATES...

77 FR 58419 - Guidelines for Preparing and Reviewing Licensing Applications for Instrumentation and Control...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-09-20

... Applications for Instrumentation and Control Upgrades for Non-Power Reactors AGENCY: Nuclear Regulatory...-Power Reactors: Format and Content,'' for instrumentation and control upgrades and NUREG-1537, Part 2, ``Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors: Standard Review...

10 CFR 140.96 - Appendix F-Indemnity locations.

Code of Federal Regulations, 2010 CFR

2010-01-01

... construction area of the nuclear power reactor, as determined by the Commission. Such area will not necessarily... or combined license under 10 CFR part 52 is issued for such additional nuclear power reactors. (2) In... an existing nuclear power reactor, the geographical boundaries of the indemnity location shall...

76 FR 74630 - Making Changes to Emergency Plans for Nuclear Power Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2011-12-01

... NUCLEAR REGULATORY COMMISSION 10 CFR Parts 50 and 52 RIN 3150-AI10 [NRC-2008-0122] Making Changes to Emergency Plans for Nuclear Power Reactors AGENCY: Nuclear Regulatory Commission. ACTION... guide (RG) 1.219, ``Guidance on Making Changes to Emergency Plans for Nuclear Power Reactors.'' This...

10 CFR 50.72 - Immediate notification requirements for operating nuclear power reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 1 2012-01-01 2012-01-01 false Immediate notification requirements for operating nuclear power reactors. 50.72 Section 50.72 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... notification requirements for operating nuclear power reactors. (a) General requirements. 1 (1) Each nuclear...

10 CFR 50.44 - Combustible gas control for nuclear power reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 1 2013-01-01 2013-01-01 false Combustible gas control for nuclear power reactors. 50.44 Section 50.44 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION... for nuclear power reactors. (a) Definitions—(1) Inerted atmosphere means a containment atmosphere with...

10 CFR 50.44 - Combustible gas control for nuclear power reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 1 2011-01-01 2011-01-01 false Combustible gas control for nuclear power reactors. 50.44 Section 50.44 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION... for nuclear power reactors. (a) Definitions—(1) Inerted atmosphere means a containment atmosphere with...

10 CFR 50.72 - Immediate notification requirements for operating nuclear power reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 1 2013-01-01 2013-01-01 false Immediate notification requirements for operating nuclear power reactors. 50.72 Section 50.72 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... notification requirements for operating nuclear power reactors. (a) General requirements. 1 (1) Each nuclear...

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.

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.

Biomass Production of Hairy Roots of Artemisia annua and Arachis hypogaea in a Scaled-Up Mist Bioreactor

PubMed Central

Sivakumar, Ganapathy; Liu, Chunzhao; Towler, Melissa J.

2014-01-01

Hairy roots have the potential to produce a variety of valuable small and large molecules. The mist reactor is a gas phase bioreactor that has shown promise for low-cost culture of hairy roots. Using a newer, disposable culture bag, mist reactor performance was studied with two species, Artemisia annua L. and Arachis hypogaea (peanut), at scales from 1 to 20 L. Both species of hairy roots when grown at 1 L in the mist reactor showed growth rates that surpassed that in shake flasks. From the information gleaned at 1 L, Arachis was scaled further to 4 and then 20 L. Misting duty cycle, culture medium flow rate, and timing of when flow rate was increased were varied. In a mist reactor increasing the misting cycle or increasing the medium flow rate are the two alternatives for increased delivery of liquid nutrients to the root bed. Longer misting cycles beyond 2–3 min were generally deemed detrimental to growth. On the other hand, increasing the medium flow rate to the sonic nozzle especially during the exponential phase of root growth (weeks 2–3) was the most important factor for increasing growth rates and biomass yields in the 20 L reactors. A. hypogaea growth in 1 L reactors was μ = 0.173 day−1 with biomass yield of 12.75 g DWL−1. This exceeded that in shake flasks at μ = 0.166 day−1 and 11.10 g DWL−1. Best growth rate and biomass yield at 20 L was μ = 0.147 and 7.77 g DWL−1, which was mainly achieved when medium flow rate delivery was increased. The mist deposition model was further evaluated using this newer reactor design and when the apparent thickness of roots (+hairs) was taken into account, the empirical data correlated with model predictions. Together these results establish the most important conditions to explore for future optimization of the mist bioreactor for culture of hairy roots. PMID:20687140

Design of megawatt power level heat pipe reactors

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

Mcclure, Patrick Ray; Poston, David Irvin; Dasari, Venkateswara Rao

An important niche for nuclear energy is the need for power at remote locations removed from a reliable electrical grid. Nuclear energy has potential applications at strategic defense locations, theaters of battle, remote communities, and emergency locations. With proper safeguards, a 1 to 10-MWe (megawatt electric) mobile reactor system could provide robust, self-contained, and long-term power in any environment. Heat pipe-cooled fast-spectrum nuclear reactors have been identified as a candidate for these applications. Heat pipe reactors, using alkali metal heat pipes, are perfectly suited for mobile applications because their nature is inherently simpler, smaller, and more reliable than “traditional” reactors.more » The goal of this project was to develop a scalable conceptual design for a compact reactor and to identify scaling issues for compact heat pipe cooled reactors in general. Toward this goal two detailed concepts were developed, the first concept with more conventional materials and a power of about 2 MWe and a the second concept with less conventional materials and a power level of about 5 MWe. A series of more qualitative advanced designs were developed (with less detail) that show power levels can be pushed to approximately 30 MWe.« less

Assessment of nuclear reactor concepts for low power space applications

NASA Technical Reports Server (NTRS)

Klein, Andrew C.; Gedeon, Stephen R.; Morey, Dennis C.

1988-01-01

The results of a preliminary small reactor concepts feasibility and safety evaluation designed to provide a first order validation of the nuclear feasibility and safety of six small reactor concepts are given. These small reactor concepts have potential space applications for missions in the 1 to 20 kWe power output range. It was concluded that low power concepts are available from the U.S. nuclear industry that have the potential for meeting both the operational and launch safety space mission requirements. However, each design has its uncertainties, and further work is required. The reactor concepts must be mated to a power conversion technology that can offer safe and reliable operation.

78 FR 48501 - Agency Information Collection Activities: Proposed Collection; Comment Request

Federal Register 2010, 2011, 2012, 2013, 2014

2013-08-08

... storage installations, decommissioned power reactors, power reactors under construction, research and test reactors, agreement states, non-agreement states, as well as departments of health, medical centers, steel...

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.

Operators in the Plum Brook Reactor Facility Control Room

NASA Image and Video Library

1970-03-21

Donald Rhodes, left, and Clyde Greer, right, monitor the operation of the National Aeronautics and Space Administration’s (NASA) Plum Brook Reactor Facility from the control room. The 60-megawatt test reactor, NASA’s only reactor, was the eighth largest test reactor in the world. The facility was built by the Lewis Research Center in the late 1950s to study the effects of radiation on different materials that could be used to construct nuclear propulsion systems for aircraft or rockets. The reactor went critical for the first time in 1961. For the next two years, two operators were on duty 24 hours per day working on the fission process until the reactor reached its full-power level in 1963. Reactor Operators were responsible for monitoring and controlling the reactor systems. Once the reactor was running under normal operating conditions, the work was relatively uneventful. Normally the reactor was kept at a designated power level within certain limits. Occasionally the operators had to increase the power for a certain test. The shift supervisor and several different people would get together and discuss the change before boosting the power. All operators were required to maintain a Reactor Operator License from the Atomic Energy Commission. The license included six months of training, an eight-hour written exam, a four-hour walkaround, and testing on the reactor controls.

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.

High-intensity power-resolved radiation imaging of an operational nuclear reactor.

PubMed

Beaumont, Jonathan S; Mellor, Matthew P; Villa, Mario; Joyce, Malcolm J

2015-10-09

Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors.

High-intensity power-resolved radiation imaging of an operational nuclear reactor

PubMed Central

Beaumont, Jonathan S.; Mellor, Matthew P.; Villa, Mario; Joyce, Malcolm J.

2015-01-01

Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors. PMID:26450669

10 CFR Appendix N to Part 50 - Standardization of Nuclear Power Plant Designs: Permits To Construct and Licenses To Operate...

Code of Federal Regulations, 2011 CFR

2011-01-01

... Construct and Licenses To Operate Nuclear Power Reactors of Identical Design at Multiple Sites N Appendix N... Construct and Licenses To Operate Nuclear Power Reactors of Identical Design at Multiple Sites Section 101... nuclear power reactors of essentially the same design to be located at different sites. 1 1 If the design...

10 CFR Appendix N to Part 50 - Standardization of Nuclear Power Plant Designs: Permits To Construct and Licenses To Operate...

Code of Federal Regulations, 2012 CFR

2012-01-01

... Construct and Licenses To Operate Nuclear Power Reactors of Identical Design at Multiple Sites N Appendix N... Construct and Licenses To Operate Nuclear Power Reactors of Identical Design at Multiple Sites Section 101... nuclear power reactors of essentially the same design to be located at different sites. 1 1 If the design...

10 CFR Appendix N to Part 50 - Standardization of Nuclear Power Plant Designs: Permits To Construct and Licenses To Operate...

Code of Federal Regulations, 2013 CFR

2013-01-01

... Construct and Licenses To Operate Nuclear Power Reactors of Identical Design at Multiple Sites N Appendix N... Construct and Licenses To Operate Nuclear Power Reactors of Identical Design at Multiple Sites Section 101... nuclear power reactors of essentially the same design to be located at different sites. 1 1 If the design...

75 FR 70042 - In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-16

... NUCLEAR REGULATORY COMMISSION [Docket Nos (Redacted), License Nos (Redacted), EA (Redacted); NRC- 2010-0351] In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent Nuclear Fuel; Order Modifying License (Effective Immediately) I. The licensees identified in...

75 FR 79423 - In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-12-20

... NUCLEAR REGULATORY COMMISSION [Docket Nos. (Redacted), License Nos.: (Redacted), EA (Redacted); NRC- 2010-0351] In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent Nuclear Fuel; Order Modifying License (Effective Immediately) I The licensees identified in...

The fractalline properties of experimentally simulated PWR fuel crud

NASA Astrophysics Data System (ADS)

Dumnernchanvanit, I.; Mishra, V. K.; Zhang, N. Q.; Robertson, S.; Delmore, A.; Mota, G.; Hussey, D.; Wang, G.; Byers, W. A.; Short, M. P.

2018-02-01

The buildup of fouling deposits on nuclear fuel rods, known as crud, continues to challenge the worldwide fleet of light water reactors (LWRs). Crud may cause serious operational problems for LWRs, including axial power shifts, accelerated fuel clad corrosion, increased primary circuit radiation dose rates, and in some instances has led directly to fuel failure. Numerous studies continue to attempt to model and predict the effects of crud, but each makes critical assumptions regarding how to treat the complex, porous microstructure of crud and its resultant effects on temperature, pressure, and crud chemistry. In this study, we demonstrate that crud is indeed a fractalline porous medium using flowing loop experiments, validating the most recent models of its effects on LWR fuel cladding. This crud is shown to match that in other LWR-prototypical facilities through a porosity-fractal dimension scaling law. Implications of this result range from post-mortem analysis of the effects of crud on reactor fuel performance, to utilizing crud's fractalline dimensions to quantify the effectiveness of anti-fouling measures.

Application of lower aliphatic alcohols as reducing agents for increasing efficiency of the LCLD process

NASA Astrophysics Data System (ADS)

Semenok, Dmitrii

2014-05-01

A method is described that is promising for application metal conductors on ceramic substrates during printed-circuit boards (PCBs) production without masking plate. The main idea of laser-induced metal deposition from solution (LCLD) consists of implementation of chemical micro reactor by using a focused laser beam. In this reactor the red/ox reaction would be initiated due to heating of a reaction medium. We used a 532 nm DPSS laser (power: 2100 mW) and water solutions of organic alcohols with low molecular weight, ethanol and isopropanol as reductants. The results of deposition were studied using the SEM, EDX methods and impedance spectroscopy. The equivalent resistance-capacitance circuit of copper tracks was constructed. The experiments showed that increasing the rate of deposition of nanostructured copper tracks up to 50 μm/s with electrical resistivity 5 Ohm/cm is possible by replacing the well-known reductants such as formaldehyde and D-sorbitol with iso-propanol.

Analysis of UF6 breeder reactor power plants

NASA Technical Reports Server (NTRS)

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

1976-01-01

Gaseous UF6 fueled breeder reactor design and technical applications of such concepts are summarized. Special attention was given to application in nuclear power plants and to reactor efficiency and safety factors.

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 to absorb the decay heat of the reactor fuel while cooling the PAHTR after an emergency shutdown. The simulated reactivity insertion accident assessment determined the maximum allowable reactivity insertion to the PAHTR as a function of shutdown response times.

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.

75 FR 13142 - Florida Power and Light Company; Turkey Point, Units 3 and 4; Exemption

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-18

... Light Company; Turkey Point, Units 3 and 4; Exemption 1.0 Background Florida Power and Light Company... ferritic materials of pressure-retaining components of the reactor coolant pressure boundary of light water... reactor coolant pressure boundary of light water nuclear power reactors to provide adequate margins of...

78 FR 64029 - Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-25

... NUCLEAR REGULATORY COMMISSION [NRC-2013-0237] Cost-Benefit Analysis for Radwaste Systems for Light... (RG) 1.110, ``Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors... components for light water nuclear power reactors. ADDRESSES: Please refer to Docket ID NRC-2013-0237 when...

10 CFR 50.36a - Technical specifications on effluents from nuclear power reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 1 2012-01-01 2012-01-01 false Technical specifications on effluents from nuclear power reactors. 50.36a Section 50.36a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND...; Ineligibility of Certain Applicants § 50.36a Technical specifications on effluents from nuclear power reactors...

10 CFR 50.36a - Technical specifications on effluents from nuclear power reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 1 2013-01-01 2013-01-01 false Technical specifications on effluents from nuclear power reactors. 50.36a Section 50.36a Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND...; Ineligibility of Certain Applicants § 50.36a Technical specifications on effluents from nuclear power reactors...

Acclimatization process of tofu wastewater on hybrid upflow anaerobic sludge blanket reactor using polyvinyl chloride rings as a growth medium

NASA Astrophysics Data System (ADS)

Yanqoritha, Nyimas; Turmuzi, Muhammad; Derlini

2017-05-01

The appropriate process to resolve sewage contamination which have a high organic using anaerobic technology. Hybrid Upflow Anaerobic Sludge Blanket reactor is one of the anaerobic process which consists of a suspended growth media and attached growth media. The reactor has the ability to work at high load rate, sludge produced easily settles, high biomass and the separation of gas, solid and liquid excelent. The purpose of research is to study the acclimatization process in the reactor of Hybrid Upflow Anaerobic Sludge Blanket using a polyvinl chloride ring as the attached growth medium. Reactor of Hybrid Upflow Anaerobic Sludge Blanket use a working volume of 8.6 L. The operation consisting of 3 L suspended reactor and 5.6 L attached reactor. Acclimatization is conducted by providing the substrate from the smallest concentration of COD up to a concentration that will be processed. During the 50th day, acclimatization process assumed the bacteria begin to work, indicated by the dissolved COD and VSS decrease and biogas production. Due to the wastewater containing the high of protein in consequence operational parameters should be controlled and some precautions should be taken to prevent process partially or totally inhibited.

Application of the Enabler to nuclear electric propulsion

NASA Astrophysics Data System (ADS)

Pierce, Bill L.

This paper describes a power system concept that provides the electric power for a baseline electric propulsion system for a piloted mission to Mars. A 10-MWe space power system is formed by coupling an Enabler reactor with a simple non-recuperated closed Brayton cycle. The Enabler reactor is a gas-cooled reactor based on proven reactor technology developed under the NERVA/Rover programs. The selected power cycle, which uses a helium-xenon mixture at 1920 K at the turbine inlet, is diagramed and described. The specific mass of the power system over the power range from 5 to 70 MWe is given. The impact of operating life on the specific mass of a 10-MWe system is also shown.

Transmutation of actinides in power reactors.

PubMed

Bergelson, B R; Gerasimov, A S; Tikhomirov, G V

2005-01-01

Power reactors can be used for partial short-term transmutation of radwaste. This transmutation is beneficial in terms of subsequent storage conditions for spent fuel in long-term storage facilities. CANDU-type reactors can transmute the main minor actinides from two or three reactors of the VVER-1000 type. A VVER-1000-type reactor can operate in a self-service mode with transmutation of its own actinides.

NEUTRONIC REACTOR POWER PLANT

DOEpatents

Metcalf, H.E.

1962-12-25

This patent relates to a nuclear reactor power plant incorporating an air-cooled, beryllium oxide-moderated, pebble bed reactor. According to the invention means are provided for circulating a flow of air through tubes in the reactor to a turbine and for directing a sidestream of the circu1ating air through the pebble bed to remove fission products therefrom as well as assist in cooling the reactor. (AEC)

Generation of OH Radical by Ultrasonic Irradiation in Batch and Circulatory Reactor

NASA Astrophysics Data System (ADS)

Fang, Yu; Shimizu, Sayaka; Yamamoto, Takuya; Komarov, Sergey

2018-03-01

Ultrasonic technology has been widely investigated in the past as one of the advance oxidation processes to treat wastewater, in this process acoustic cavitation causes generation of OH radical, which play a vital role in improving the treatment efficiency. In this study, OH radical formation rate was measured in batch and circulatory reactor by using Weissler reaction at various ultrasound output power. It is found that the generation rate in batch reactor is higher than that in circulatory reactor at the same output power. The generation rate tended to be slower when output power exceeds 137W. The optimum condition for circulatory reactor was found to be 137W output and 4L/min flow rate. Results of aluminum foil erosion test revealed a strong dependence of cavitation zone length on the ultrasound output power. This is assumed to be one of the reasons why the generation rate of HO radicals becomes slower at higher output power in circulatory reactor.

REACTOR-FLASH BOILER-FLYWHEEL POWER PLANT

DOEpatents

Loeb, E.

1961-01-17

A power generator in the form of a flywheel with four reactors positioned about its rim is described. The reactors are so positioned that steam, produced in the reactor, exists tangentially to the flywheel, giving it a rotation. The reactors are incompletely moderated without water. The water enters the flywheel at its axis, under sufficient pressure to force it through the reactors, where it is converted to steam. The fuel consists of parallel twisted ribbons assembled to approximate a cylinder.

Navy Nuclear-Powered Surface Ships: Background, Issues, and Options for Congress

DTIC Science & Technology

2010-09-29

to design a smaller scale version of a naval pressurized water reactor , or to design a new reactor type potentially using a thorium liquid salt...integrated nuclear power system capable of use on destroyer- sized vessels either using a pressurized water reactor or a thorium liquid salt reactor ...nuclear reactors for Navy surface ships. The text of Section 246 is as follows: SEC. 246. STUDY ON THORIUM -LIQUID FUELED REACTORS FOR NAVAL FORCES

Method of locating a leaking fuel element in a fast breeder power reactor

DOEpatents

Honekamp, John R.; Fryer, Richard M.

1978-01-01

Leaking fuel elements in a fast reactor are identified by measuring the ratio of .sup.134 Xe to .sup.133 Xe in the reactor cover gas following detection of a fuel element leak, this ratio being indicative of the power and burnup of the failed fuel element. This procedure can be used to identify leaking fuel elements in a power breeder reactor while continuing operation of the reactor since the ratio measured is that of the gases stored in the plenum of the failed fuel element. Thus, use of a cleanup system for the cover gas makes it possible to identify sequentially a multiplicity of leaking fuel elements without shutting the reactor down.

Core follow calculation with the nTRACER numerical reactor and verification using power reactor measurement data

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

Jung, Y. S.; Joo, H. G.; Yoon, J. I.

The nTRACER direct whole core transport code employing the planar MOC solution based 3-D calculation method, the subgroup method for resonance treatment, the Krylov matrix exponential method for depletion, and a subchannel thermal/hydraulic calculation solver was developed for practical high-fidelity simulation of power reactors. Its accuracy and performance is verified by comparing with the measurement data obtained for three pressurized water reactor cores. It is demonstrated that accurate and detailed multi-physic simulation of power reactors is practically realizable without any prior calculations or adjustments. (authors)

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.

Project Longshot: An unmanned probe to Alpha Centauri

NASA Technical Reports Server (NTRS)

Beals, Keith A.; Beaulieu, Martin; Dembia, Frank J.; Kerstiens, Joseph; Kramer, Daniel L.; West, Jeffrey R.; Zito, James A.

1988-01-01

A preliminary design is presented for an unmanned probe to Alpha Centauri with a planned launch early in the 21st century. The probe would be assembled at the space station and take approx. 100 yrs to reach the nearest star. Several technologies must be developed in order for this mission to be possible. A pulsed fusion microexplosion drive with 1,000,000 secs of specific impulse is the primary enabling technology. A large, long life fission reactor with 300 kW power output is also required. Communications lasers would use a 0.532 micrometer wavelength since there is minimal power output by the stars in that frequency band. A laser with an input power of 250 kW would allow for a data rate of 1000 bits per second at maximum range. There are 3 types of information to be gathered by the probe: properties of the interstellar medium, characteristics of the three star Alpha Centauri system, and astrometry.

10 CFR 72.210 - General license issued.

Code of Federal Regulations, 2010 CFR

2010-01-01

... NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General License for Storage of Spent Fuel at Power Reactor Sites § 72.210 General license issued. A general license is... reactor sites to persons authorized to possess or operate nuclear power reactors under 10 CFR part 50 or...

10 CFR 72.210 - General license issued.

Code of Federal Regulations, 2011 CFR

2011-01-01

... NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General License for Storage of Spent Fuel at Power Reactor Sites § 72.210 General license issued. A general license is... reactor sites to persons authorized to possess or operate nuclear power reactors under 10 CFR part 50 or...

Application of Molten Salt Reactor Technology to Nuclear Electric Propulsion Mission

NASA Technical Reports Server (NTRS)

Patton, Bruce; Sorensen, Kirk; Rodgers, Stephen L. (Technical Monitor)

2002-01-01

Nuclear electric propulsion (NEP) and planetary surface power missions require reactors that are lightweight, operationally robust, and scalable in power for widely varying scientific mission objectives. Molten salt reactor technology meets all of these requirements and offers an interesting alternative to traditional gas cooled, liquid metal, and heat pipe space reactors.

Goals of thermionic program for space power

NASA Technical Reports Server (NTRS)

English, R. E.

1981-01-01

The thermionic and Brayton reactor concepts were compared for application to space power. For a turbine inlet temperature of 15000 K the Brayton powerplant weighted 5 to 40% less than the thermionic concept. The out of core concept separates the thermionic converters from their reactor. Technical risks are diminished by: (1) moving the insolator out of the reactor; (2) allowing a higher thermal flux for the thermionic converters than is required of the reactor fuel; and (3) eliminating fuel swelling's threat against lifetime of the thermionic converters. Overall performance can be improved by including power processing in system optimization for design and technology on more efficient, higher temperature power processors. The thermionic reactors will be larger than those for competitive systems with higher conversion efficiency and lower reactor operating temperatures. It is concluded that although the effect of reactor size on shield weight will be modest for unmanned spacecraft, the penalty in shield weight will be large for manned or man-tended spacecraft.

10 CFR 50.69 - Risk-informed categorization and treatment of structures, systems and components for nuclear...

Code of Federal Regulations, 2012 CFR

2012-01-01

..., systems and components for nuclear power reactors. 50.69 Section 50.69 Energy NUCLEAR REGULATORY..., systems and components for nuclear power reactors. (a) Definitions. Risk-Informed Safety Class (RISC)-1... holder of a license to operate a light water reactor (LWR) nuclear power plant under this part; a holder...

10 CFR 50.69 - Risk-informed categorization and treatment of structures, systems and components for nuclear...

Code of Federal Regulations, 2013 CFR

2013-01-01

..., systems and components for nuclear power reactors. 50.69 Section 50.69 Energy NUCLEAR REGULATORY..., systems and components for nuclear power reactors. (a) Definitions. Risk-Informed Safety Class (RISC)-1... holder of a license to operate a light water reactor (LWR) nuclear power plant under this part; a holder...

10 CFR 50.69 - Risk-informed categorization and treatment of structures, systems and components for nuclear...

Code of Federal Regulations, 2014 CFR

2014-01-01

..., systems and components for nuclear power reactors. 50.69 Section 50.69 Energy NUCLEAR REGULATORY..., systems and components for nuclear power reactors. (a) Definitions. Risk-Informed Safety Class (RISC)-1... holder of a license to operate a light water reactor (LWR) nuclear power plant under this part; a holder...

Tritium release during nuclear power operation in China.

PubMed

Yang, D J; Chen, X Q; Li, B

2012-06-01

Overviews were evaluated of tritium releases and related doses to the public from airborne and liquid effluents from nuclear power plants on the mainland of China before 2009. The differences between tritium releases from various nuclear power plants were also evaluated. The tritium releases are mainly from liquid pathways for pressurised water reactors, but tritium releases between airborne and liquid effluents are comparable for heavy water reactors. The airborne release from a heavy water reactor is obviously higher than that from a pressurised water reactor.

Navy Nuclear-Powered Surface Ships: Background, Issues, and Options for Congress

DTIC Science & Technology

2010-06-10

scale pressurized water reactors suitable for destroyer-sized vessels or for alternative nuclear power systems using thorium liquid salt technology...or to design a new reactor type potentially using a thorium liquid salt reactor developed for maritime use. The committee recommends an increase of...either using a pressurized water reactor or a thorium liquid salt reactor . (Page 158) Senate The Senate Armed Services Committee, in its report

An underground nuclear power station using self-regulating heat-pipe controlled reactors

DOEpatents

Hampel, V.E.

1988-05-17

A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working fluid in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast- acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor. 5 figs.

Underground nuclear power station using self-regulating heat-pipe controlled reactors

DOEpatents

Hampel, Viktor E.

1989-01-01

A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working flud in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast-acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor.

Nuclear Energy Policy

DTIC Science & Technology

2009-12-10

Small Modular Reactors Rising cost estimates for large conventional nuclear power plants—widely projected to be $6 billion or more—have contributed to growing interest in proposals for smaller, modular reactors. Ranging from about 40 to 350 megawatts of electrical capacity, such reactors would be only a fraction of the size of current commercial reactors. Several modular reactors would be installed together to make up a power block with a single control room, under most concepts. Modular reactor concepts would use a variety of technologies,

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

Adamov, E.O.; Lebedev, V.A.; Kuznetsov, Yu.N.

Zheleznogorsk is situated near the territorial center -- Krasnoyarsk on the Yenisei river. Mining and chemical complex is the main industrial enterprise of the town, which has been constructed for generation and used for isolation of weapons-grade plutonium. Heat supply to the chemical complex and town at the moment is largely provided by nuclear co-generation plant (NCGP) on the basis of the ADEh-2 dual-purpose reactor, generating 430 Gcal/h of heat and, partially, by coal backup peak-load boiler houses. NCGP also provides 73% of electric power consumed. In line with agreements between Russia and USA on strategic arms reduction and phasingmore » out of weapons-grade plutonium production, decommissioning of the ADEh-2 reactor by 2000 is planned. Thus, a problem arises relative to compensation for electric and thermal power generation for the needs of the town and industrial enterprises, which is now supplied by the reactor. A nuclear power plant constructed on the same site as a substituting power source should be considered as the most practical option. Basic requirements to the reactor of substituting nuclear power plant are as follows. It is to be a new generation reactor on the basis of verified technologies, having an operating prototype optimal for underground siting and permitting utmost utilization of the available mining workings and those being disengaged. NCGP with the reactor is to be constructed in the time period required and is to become competitive with other possible power sources. Analysis has shown that the VK-300 simplified vessel-type boiling reactor meets the requirements made in the maximum extent. Its design is based on the experience of the VK-50 reactor operation for a period of 30 years in Dimitrovgrad (Russia) and allows for experience in the development of the SBWR type reactors. The design of the reactor is discussed.« less

Experimental study of radiation dose rate at different strategic points of the BAEC TRIGA Research Reactor.

PubMed

Ajijul Hoq, M; Malek Soner, M A; Salam, M A; Haque, M M; Khanom, Salma; Fahad, S M

2017-12-01

The 3MW TRIGA Mark-II Research Reactor of Bangladesh Atomic Energy Commission (BAEC) has been under operation for about thirty years since its commissioning at 1986. In accordance with the demand of fundamental nuclear research works, the reactor has to operate at different power levels by utilizing a number of experimental facilities. Regarding the enquiry for safety of reactor operating personnel and radiation workers, it is necessary to know the radiation level at different strategic points of the reactor where they are often worked. In the present study, neutron, beta and gamma radiation dose rate at different strategic points of the reactor facility with reactor power level of 2.4MW was measured to estimate the rising level of radiation due to its operational activities. From the obtained results high radiation dose is observed at the measurement position of the piercing beam port which is caused by neutron leakage and accordingly, dose rate at the stated position with different reactor power levels was measured. This study also deals with the gamma dose rate measurements at a fixed position of the reactor pool top surface for different reactor power levels under both Natural Convection Cooling Mode (NCCM) and Forced Convection Cooling Mode (FCCM). Results show that, radiation dose rate is higher for NCCM in compared with FCCM and increasing with the increase of reactor power. Thus, concerning the radiological safety issues for working personnel and the general public, the radiation dose level monitoring and the experimental analysis performed within this paper is so much effective and the result of this work can be utilized for base line data and code verification of the nuclear reactor. Copyright © 2017 Elsevier Ltd. All rights reserved.

Project Luna Succendo: The Lunar Evolutionary Growth-Optimized (LEGO) Reactor

NASA Astrophysics Data System (ADS)

Bess, John Darrell

A final design has been established for a basic Lunar Evolutionary Growth-Optimized (LEGO) Reactor using current and near-term technologies. The LEGO Reactor is a modular, fast-fission, heatpipe-cooled, clustered-reactor system for lunar-surface power generation. The reactor is divided into subcritical units that can be safely launched within lunar shipments from the Earth, and then emplaced directly into holes drilled into the lunar regolith to form a critical reactor assembly. The regolith would not just provide radiation shielding, but serve as neutron-reflector material as well. The reactor subunits are to be manufactured using proven and tested materials for use in radiation environments, such as uranium-dioxide fuel, stainless-steel cladding and structural support, and liquid-sodium heatpipes. The LEGO Reactor system promotes reliability, safety, and ease of manufacture and testing at the cost of an increase in launch mass per overall rated power level and a reduction in neutron economy when compared to a single-reactor system. A single unshielded LEGO Reactor subunit has an estimated mass of approximately 448 kg and provides 5 kWe using a free-piston Stirling space converter. The overall envelope for a single unit with fully extended radiator panels has a height of 8.77 m and a diameter of 0.50 m. The subunits can be placed with centerline distances of approximately 0.6 m in a hexagonal-lattice pattern to provide sufficient neutronic coupling while allowing room for heat rejection and interstitial control. A lattice of six subunits could provide sufficient power generation throughout the initial stages of establishing a lunar outpost. Portions of the reactor may be neutronically decoupled to allow for reduced power production during unmanned periods of base operations. During later stages of lunar-base development, additional subunits may be emplaced and coupled into the existing LEGO Reactor network Future improvements include advances in reactor control methods, fuel form and matrix, determination of shielding requirements, as well as power conversion and heat rejection techniques to generate an even more competitive LEGO Reactor design. Further modifications in the design could provide power generative opportunities for use on other extraterrestrial surfaces such as Mars, other moons, and asteroids.

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.

Dual-mode, high energy utilization system concept for mars missions

NASA Astrophysics Data System (ADS)

El-Genk, Mohamed S.

2000-01-01

This paper describes a dual-mode, high energy utilization system concept based on the Pellet Bed Reactor (PeBR) to support future manned missions to Mars. The system uses proven Closed Brayton Cycle (CBC) engines to partially convert the reactor thermal power to electricity. The electric power generated is kept the same during the propulsion and the power modes, but the reactor thermal power in the former could be several times higher, while maintaining the reactor temperatures almost constant. During the propulsion mode, the electric power of the system, minus ~1-5 kWe for house keeping, is used to operate a Variable Specific Impulse Magnetoplasma Rocket (VASIMR). In addition, the reactor thermal power, plus more than 85% of the head load of the CBC engine radiators, are used to heat hydrogen. The hot hydrogen is mixed with the high temperature plasma in a VASIMR to provide both high thrust and Isp>35,000 N.s/kg, reducing the travel time to Mars to about 3 months. The electric power also supports surface exploration of Mars. The fuel temperature and the inlet temperatures of the He-Xe working fluid to the nuclear reactor core and the CBC turbine are maintained almost constant during both the propulsion and power modes to minimize thermal stresses. Also, the exit temperature of the He-Xe from the reactor core is kept at least 200 K below the maximum fuel design temperature. The present system has no single point failure and could be tested fully assembled in a ground facility using electric heaters in place of the nuclear reactor. Operation and design parameters of a 40-kWe prototype are presented and discussed to illustrate the operation and design principles of the proposed system. .

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.

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.

Adaptive control method for core power control in TRIGA Mark II reactor

NASA Astrophysics Data System (ADS)

Sabri Minhat, Mohd; Selamat, Hazlina; Subha, Nurul Adilla Mohd

2018-01-01

The 1MWth Reactor TRIGA PUSPATI (RTP) Mark II type has undergone more than 35 years of operation. The existing core power control uses feedback control algorithm (FCA). It is challenging to keep the core power stable at the desired value within acceptable error bands to meet the safety demand of RTP due to the sensitivity of nuclear research reactor operation. Currently, the system is not satisfied with power tracking performance and can be improved. Therefore, a new design core power control is very important to improve the current performance in tracking and regulate reactor power by control the movement of control rods. In this paper, the adaptive controller and focus on Model Reference Adaptive Control (MRAC) and Self-Tuning Control (STC) were applied to the control of the core power. The model for core power control was based on mathematical models of the reactor core, adaptive controller model, and control rods selection programming. The mathematical models of the reactor core were based on point kinetics model, thermal hydraulic models, and reactivity models. The adaptive control model was presented using Lyapunov method to ensure stable close loop system and STC Generalised Minimum Variance (GMV) Controller was not necessary to know the exact plant transfer function in designing the core power control. The performance between proposed adaptive control and FCA will be compared via computer simulation and analysed the simulation results manifest the effectiveness and the good performance of the proposed control method for core power control.

78 FR 65008 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Evolutionary...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-30

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Evolutionary Power Reactor; Notice of Meeting The ACRS Subcommittee on U.S. Evolutionary Power Reactor (U.S. EPR) will hold a meeting on November 6, 2013, Room T-2B1, 11545 Rockville Pike, Rockville...

10 CFR 50.64 - Limitations on the use of highly enriched uranium (HEU) in domestic non-power reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

... domestic non-power reactors. 50.64 Section 50.64 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... Director of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC... Director of the Office of Nuclear Reactor Regulation a written proposal for meeting the requirements of...

10 CFR 50.64 - Limitations on the use of highly enriched uranium (HEU) in domestic non-power reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

... domestic non-power reactors. 50.64 Section 50.64 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... Director of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC... Director of the Office of Nuclear Reactor Regulation a written proposal for meeting the requirements of...

10 CFR 50.64 - Limitations on the use of highly enriched uranium (HEU) in domestic non-power reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

... domestic non-power reactors. 50.64 Section 50.64 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... Director of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC... Director of the Office of Nuclear Reactor Regulation a written proposal for meeting the requirements of...

Neutrino mass hierarchy and precision physics with medium-baseline reactors: Impact of energy-scale and flux-shape uncertainties

NASA Astrophysics Data System (ADS)

Capozzi, F.; Lisi, E.; Marrone, A.

2015-11-01

Nuclear reactors provide intense sources of electron antineutrinos, characterized by few-MeV energy E and unoscillated spectral shape Φ (E ). High-statistics observations of reactor neutrino oscillations over medium-baseline distances L ˜O (50 ) km would provide unprecedented opportunities to probe both the long-wavelength mass-mixing parameters (δ m2 and θ12) and the short-wavelength ones (Δ mee 2 and θ13), together with the subtle interference effects associated with the neutrino mass hierarchy (either normal or inverted). In a given experimental setting—here taken as in the JUNO project for definiteness—the achievable hierarchy sensitivity and parameter accuracy depend not only on the accumulated statistics but also on systematic uncertainties, which include (but are not limited to) the mass-mixing priors and the normalizations of signals and backgrounds. We examine, in addition, the effect of introducing smooth deformations of the detector energy scale, E →E'(E ), and of the reactor flux shape, Φ (E )→Φ'(E ), within reasonable error bands inspired by state-of-the-art estimates. It turns out that energy-scale and flux-shape systematics can noticeably affect the performance of a JUNO-like experiment, both on the hierarchy discrimination and on precision oscillation physics. It is shown that a significant reduction of the assumed energy-scale and flux-shape uncertainties (by, say, a factor of 2) would be highly beneficial to the physics program of medium-baseline reactor projects. Our results also shed some light on the role of the inverse-beta decay threshold, of geoneutrino backgrounds, and of matter effects in the analysis of future reactor oscillation data.

Gaseous fuel reactors for power systems

NASA Technical Reports Server (NTRS)

Kendall, J. S.; Rodgers, R. J.

1977-01-01

Gaseous-fuel nuclear reactors have significant advantages as energy sources for closed-cycle power systems. The advantages arise from the removal of temperature limits associated with conventional reactor fuel elements, the wide variety of methods of extracting energy from fissioning gases, and inherent low fissile and fission product in-core inventory due to continuous fuel reprocessing. Example power cycles and their general performance characteristics are discussed. Efficiencies of gaseous fuel reactor systems are shown to be high with resulting minimal environmental effects. A technical overview of the NASA-funded research program in gaseous fuel reactors is described and results of recent tests of uranium hexafluoride (UF6)-fueled critical assemblies are presented.

Period meter for reactors

DOEpatents

Rusch, Gordon K.

1976-01-06

An improved log N amplifier type nuclear reactor period meter with reduced probability for noise-induced scrams is provided. With the reactor at low power levels a sampling circuit is provided to determine the reactor period by measuring the finite change in the amplitude of the log N amplifier output signal for a predetermined time period, while at high power levels, differentiation of the log N amplifier output signal provides an additional measure of the reactor period.

Application of Molten Salt Reactor Technology to MMW In-Space NEP and Surface Power Missions

NASA Technical Reports Server (NTRS)

Patton, Bruce; Sorensen, Kirk; Rodgers, Stephen (Technical Monitor)

2002-01-01

Anticipated manned nuclear electric propulsion (NEP) and planetary surface power missions will require multimegawatt nuclear reactors that are lightweight, operationally robust, and scalable in power for widely varying scientific mission objectives. Molten salt reactor technology meets all of these requirements and offers an interesting alternative to traditional multimegawatt gas-cooled and liquid metal concepts.

100-kWe lunar/Mars surface power utilizing the SP-100 reactor with dynamic conversion

NASA Technical Reports Server (NTRS)

Harty, Richard B.; Mason, Lee S.

1992-01-01

Results are presented from a study of the coupling of an SP-100 nuclear reactor with either a Stirling or Brayton power system, at the 100 kWe level, for a power generating system suitable for operation in the lunar and Martian surface environments. In the lunar environment, the reactor and primary coolant loop would be contained in a guard vessel to protect from a loss of primary loop containment. For Mars, all refractory components, including the reactor, coolant, and power conversion components will be contained in a vacuum vessel for protection against the CO2 environment.

Shield Design for Lunar Surface Applications

NASA Astrophysics Data System (ADS)

Johnson, Gregory A.

2006-01-01

A shielding concept for lunar surface applications of nuclear power is presented herein. The reactor, primary shield, reactor equipment and power generation module are placed in a cavity in the lunar surface. Support structure and heat rejection radiator panels are on the surface, outside the cavity. The reactor power of 1,320 kWt was sized to deliver 50 kWe from a thermoelectric power conversion subsystem. The dose rate on the surface is less than 0.6 mRem/hr at 100 meters from the reactor. Unoptimized shield mass is 1,020 kg which is much lighter than a comparable 4π shield weighing in at 17,000 kg.

Electromechanical, acoustical and thermodynamical characterization of a low-frequency sonotrode-type transducer in a small sonoreactor at different excitation levels and loading conditions.

PubMed

Petošić, Antonio; Horvat, Marko; Režek Jambrak, Anet

2017-11-01

The paper reports and compares the results of the electromechanical, acoustical and thermodynamical characterization of a low-frequency sonotrode-type ultrasonic device inside a small sonoreactor, immersed in three different loading media, namely, water, juice and milk, excited at different excitation levels, both below and above the cavitation threshold. The electroacoustic efficiency factor determined at system resonance through electromechanical characterization in degassed water as the reference medium is 88.7% for the device in question. This efficiency can be reduced up to three times due to the existence of a complex sound field in the reactor in linear driving conditions below the cavitation threshold. The behaviour of the system is more stable at higher excitation levels than in linear operating conditions. During acoustical characterization, acoustic pressure is spatially averaged, both below and above the cavitation threshold. The standing wave patterns inside the sonoreactor have a stronger influence on the variation of the spatially distributed RMS pressure in linear operating conditions. For these conditions, the variation of ±1.7dB was obtained, compared to ±1.4dB obtained in highly nonlinear regime. The acoustic power in the sonoreactor was estimated from the magnitude of the averaged RMS pressure, and from the reverberation time of the sonoreactor as the representation of the losses. The electroacoustic efficiency factors obtained through acoustical and electromechanical characterization are in a very good agreement at low excitation levels. The irradiated acoustic power estimated in nonlinear conditions differs from the dissipated acoustic power determined with the calorimetric method by several orders of magnitude. The number of negative pressure peaks that represent transient cavitation decreases over time during longer treatments of a medium with high-power ultrasound. The number of negative peaks decreases faster when the medium and the vessel are allowed to heat up. Copyright © 2017 Elsevier B.V. All rights reserved.

The Cost-Effectiveness of Nuclear Power for Navy Surface Ships

DTIC Science & Technology

2011-05-01

shipbuilding plan. 1 All of the Navy’s aircraft car- riers (and submarines) are powered by nuclear reactors ; its other surface combatants are powered by...in whether the ships were powered by conventional systems that used petroleum-based fuels or by nuclear reactors . Estimates of the relative costs...would existing ships be retrofitted with nuclear reactors . 5. Those fuel -reduction findings are based on CBO’s analysis and on data provided to CBO by

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

A Power Conversion Concept for the Jupiter Icy Moons Orbiter

NASA Technical Reports Server (NTRS)

Mason, Lee S.

2003-01-01

The Jupiter Icy Moons Orbiter (JIMO) is a bold new mission under development by the Office of Space Science at NASA Headquarters. ITMO is examining the potential of Nuclear Electric Propulsion (NEP) technology to efficiently deliver scientific payloads to three Jovian moons: Callisto, Ganymede, and Europa. A critical element of the NEP vehicle is the reactor power system, consisting of the nuclear reactor, power conversion, heat rejection, and power management and distribution (PMAD). The emphasis of this paper is on the non-nuclear elements of the reactor power system.

The startup performance and microbial distribution of an anaerobic baffled reactor (ABR) treating medium-strength synthetic industrial wastewater.

PubMed

Jiang, Hao; Nie, Hong; Ding, Jiangtao; Stinner, Walter; Sun, Kaixuan; Zhou, Hongjun

2018-01-02

In this study, an anaerobic baffled reactor (ABR) with seven chambers was applied to treat medium-strength synthetic industrial wastewater (MSIW). The performance of startup and shock test on treating MSIW was investigated. During the acclimation process, the chemical oxygen demand (COD) of MSIW gradually increased from 0 to 2,000 mg L -1 , and the COD removal finally reached 90%. At shock test, the feeding COD concentration increased by one-fifth and the reactor adapted very well with a COD removal of 82%. In a stable state, Comamonas, Smithella, Syntrophomonas and Pseudomonas were the main populations of bacteria, while the predominant methanogen was Methanobacterium. The results of chemical and microbiological analysis indicated the significant advantages of ABR, including buffering shocks, separating stages with matching microorganisms and promoting syntrophism. Meanwhile, the strategies for acclimation and operation were of great importance. Further work can test reactor performance in the treatment of actual industrial wastewater.

Electrochemical study of multi-electrode microbial fuel cells under fed-batch and continuous flow conditions

NASA Astrophysics Data System (ADS)

Ren, Lijiao; Ahn, Yongtae; Hou, Huijie; Zhang, Fang; Logan, Bruce E.

2014-07-01

Power production of four hydraulically connected microbial fuel cells (MFCs) was compared with the reactors operated using individual electrical circuits (individual), and when four anodes were wired together and connected to four cathodes all wired together (combined), in fed-batch or continuous flow conditions. Power production under these different conditions could not be made based on a single resistance, but instead required polarization tests to assess individual performance relative to the combined MFCs. Based on the power curves, power produced by the combined MFCs (2.12 ± 0.03 mW, 200 Ω) was the same as the summed power (2.13 mW, 50 Ω) produced by the four individual reactors in fed-batch mode. With continuous flow through the four MFCs, the maximum power (0.59 ± 0.01 mW) produced by the combined MFCs was slightly lower than the summed maximum power of the four individual reactors (0.68 ± 0.02 mW). There was a small parasitic current flow from adjacent anodes and cathodes, but overall performance was relatively unaffected. These findings demonstrate that optimal power production by reactors hydraulically and electrically connected can be predicted from performance by individual reactors.

77 FR 36581 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on U.S...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-06-19

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on U.S. Advanced Pressurized Power Reactor; Notice of Meeting The ACRS Subcommittee on U.S. Advanced Pressurized Power Reactor (US-APWR) will hold a meeting on July 9-10, 2012, Room T-2B3, 11545...

76 FR 44964 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on U.S...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-07-27

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on U.S. Evolutionary Power Reactor; Notice of Meeting The ACRS Subcommittee on U.S. Evolutionary Power Reactor (U.S. EPR) will hold a meeting on August 18, 2011, Room T-2B3, 11545 Rockville Pike...

75 FR 82094 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on U.S...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-12-29

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on U.S. Evolutionary Power Reactor (U.S. EPR); Notice of Meeting The ACRS Subcommittee on U.S. Evolutionary Power Reactor (U.S. EPR) will hold a meeting on January 12, 2011, Room T-2B1, 11545 Rockville Pike...

10 CFR 2.1105 - Definitions.

Code of Federal Regulations, 2010 CFR

2010-01-01

... Hearing Procedures for Expansion of Spent Nuclear Fuel Storage Capacity at Civilian Nuclear Power Reactors § 2.1105 Definitions. As used in this part: (a) Civilian nuclear power reactor means a civilian... reactor following irradiation, the constituent elements of which have not been separated by reprocessing. ...

Station Blackout Analysis of HTGR-Type Experimental Power Reactor

NASA Astrophysics Data System (ADS)

Syarip; Zuhdi, Aliq; Falah, Sabilul

2018-01-01

The National Nuclear Energy Agency of Indonesia has decided to build an experimental power reactor of high-temperature gas-cooled reactor (HTGR) type located at Puspiptek Complex. The purpose of this project is to demonstrate a small modular nuclear power plant that can be operated safely. One of the reactor safety characteristics is the reliability of the reactor to the station blackout (SBO) event. The event was observed due to relatively high disturbance frequency of electricity network in Indonesia. The PCTRAN-HTR functional simulator code was used to observe fuel and coolant temperature, and coolant pressure during the SBO event. The reactor simulated at 10 MW for 7200 s then the SBO occurred for 1-3 minutes. The analysis result shows that the reactor power decreases automatically as the temperature increase during SBO accident without operator’s active action. The fuel temperature increased by 36.57 °C every minute during SBO and the power decreased by 0.069 MW every °C fuel temperature rise at the condition of anticipated transient without reactor scram. Whilst, the maximum coolant (helium) temperature and pressure are 1004 °C and 9.2 MPa respectively. The maximum fuel temperature is 1282 °C, this value still far below the fuel temperature limiting condition i.e. 1600 °C, its mean that the HTGR has a very good inherent safety system.

10 CFR 2.501 - Notice of hearing on application under subpart F of 10 CFR part 52 for a license to manufacture...

Code of Federal Regulations, 2011 CFR

2011-01-01

... part 52 for a license to manufacture nuclear power reactors. 2.501 Section 2.501 Energy NUCLEAR... Procedures Applicable to Proceedings for the Issuance of Licenses To Manufacture Nuclear Power Reactors To Be... power reactors. (a) In the case of an application under subpart F of part 52 of this chapter for a...

10 CFR 2.501 - Notice of hearing on application under subpart F of 10 CFR part 52 for a license to manufacture...

Code of Federal Regulations, 2010 CFR

2010-01-01

... part 52 for a license to manufacture nuclear power reactors. 2.501 Section 2.501 Energy NUCLEAR... Procedures Applicable to Proceedings for the Issuance of Licenses To Manufacture Nuclear Power Reactors To Be... power reactors. (a) In the case of an application under subpart F of part 52 of this chapter for a...

Computer simulation of magnetization-controlled shunt reactors for calculating electromagnetic transients in power systems

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

Karpov, A. S.

2013-01-15

A computer procedure for simulating magnetization-controlled dc shunt reactors is described, which enables the electromagnetic transients in electric power systems to be calculated. It is shown that, by taking technically simple measures in the control system, one can obtain high-speed reactors sufficient for many purposes, and dispense with the use of high-power devices for compensating higher harmonic components.

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

The HLMA project: determination of high Δm2 LMA mixing parameters and constraint on |Ue3| with a new reactor neutrino experiment

NASA Astrophysics Data System (ADS)

Schönert, Stefan; Lasserre, Thierry; Oberauer, Lothar

2003-03-01

In the forthcoming months, the KamLAND experiment will probe the parameter space of the solar large mixing angle MSW solution as the origin of the solar neutrino deficit with ν¯e's from distant nuclear reactors. If however the solution realized in nature is such that Δm2sol>~2×10-4 eV2 (thereafter named the HLMA region), KamLAND will only observe a rate suppression but no spectral distortion and hence it will not have the optimal sensitivity to measure the mixing parameters. In this case, we propose a new medium baseline reactor experiment located at Heilbronn (Germany) to pin down the precise value of the solar mixing parameters. In this paper, we present the Heilbronn detector site, we calculate the ν¯e interaction rate and the positron spectrum expected from the surrounding nuclear power plants. We also discuss the sensitivity of such an experiment to |Ue3| in both normal and inverted neutrino mass hierarchy scenarios. We then outline the detector design, estimate background signals induced by natural radioactivity as well as by in situ cosmic ray muon interaction, and discuss a strategy to detect the anti-neutrino signal `free of background'.

Consumption of the electric power inside silent discharge reactors

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

Yehia, Ashraf, E-mail: yehia30161@yahoo.com

An experimental study was made in this paper to investigate the relation between the places of the dielectric barriers, which cover the surfaces of the electrodes in the coaxial cylindrical reactors, and the rate of change of the electric power that is consumed in forming silent discharges. Therefore, silent discharges have been formed inside three coaxial cylindrical reactors. The dielectric barriers in these reactors were pasted on both the internal surface of the outer electrode in the first reactor and the external surface of the inner electrode in the second reactor as well as the surfaces of the two electrodesmore » in the third reactor. The reactor under study has been fed by atmospheric air that flowed inside it with a constant rate at normal temperature and pressure, in parallel with the application of a sinusoidal ac voltage between the electrodes of the reactor. The electric power consumed in forming the silent discharges inside the three reactors was measured as a function of the ac peak voltage. The validity of the experimental results was investigated by applying Manley's equation on the same discharge conditions. The results have shown that the rate of consumption of the electric power relative to the ac peak voltage per unit width of the discharge gap improves by a ratio of either 26.8% or 80% or 128% depending on the places of the dielectric barriers that cover the surfaces of the electrodes inside the three reactors.« less

77 FR 40092 - License Amendment To Increase the Maximum Reactor Power Level, Florida Power & Light Company, St...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-07-06

... Increase the Maximum Reactor Power Level, Florida Power & Light Company, St. Lucie, Units 1 and 2 AGENCY... amendment for Renewed Facility Operating License Nos. DPR-67 and NPF-16, issued to Florida Power & Light... St. Lucie County, Florida. The proposed license amendment would increase the maximum thermal power...

10 CFR 100.21 - Non-seismic siting criteria.

Code of Federal Regulations, 2011 CFR

2011-01-01

... NUCLEAR REGULATORY COMMISSION (CONTINUED) REACTOR SITE CRITERIA Evaluation Factors for Stationary Power Reactor Site Applications on or After January 10, 1997 § 100.21 Non-seismic siting criteria. Applications for site approval for commercial power reactors shall demonstrate that the proposed site meets the...

10 CFR 100.21 - Non-seismic siting criteria.

Code of Federal Regulations, 2010 CFR

2010-01-01

... NUCLEAR REGULATORY COMMISSION (CONTINUED) REACTOR SITE CRITERIA Evaluation Factors for Stationary Power Reactor Site Applications on or After January 10, 1997 § 100.21 Non-seismic siting criteria. Applications for site approval for commercial power reactors shall demonstrate that the proposed site meets the...

Addition of acetate improves stability of power generation using microbial fuel cells treating domestic wastewater.

PubMed

Stager, Jennifer L; Zhang, Xiaoyuan; Logan, Bruce E

2017-12-01

Power generation using microbial fuel cells (MFCs) must provide stable, continuous conversion of organic matter in wastewaters into electricity. However, when relatively small diameter (0.8cm) graphite fiber brush anodes were placed close to the cathodes in MFCs, power generation was unstable during treatment of low strength domestic wastewater. One reactor produced 149mW/m 2 before power generation failed, while the other reactor produced 257mW/m 2 , with both reactors exhibiting severe power overshoot in polarization tests. Using separators or activated carbon cathodes did not result in stable operation as the reactors continued to exhibit power overshoot based on polarization tests. However, adding acetate (1g/L) to the wastewater produced stable performance during fed batch and continuous flow operation, and there was no power overshoot in polarization tests. These results highlight the importance of wastewater strength and brush anode size for producing stable and continuous power in compact MFCs. Copyright © 2017 Elsevier B.V. All rights reserved.

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.

A Gas-Cooled-Reactor Closed-Brayton-Cycle Demonstration with Nuclear Heating

NASA Astrophysics Data System (ADS)

Lipinski, Ronald J.; Wright, Steven A.; Dorsey, Daniel J.; Peters, Curtis D.; Brown, Nicholas; Williamson, Joshua; Jablonski, Jennifer

2005-02-01

A gas-cooled reactor may be coupled directly to turbomachinery to form a closed-Brayton-cycle (CBC) system in which the CBC working fluid serves as the reactor coolant. Such a system has the potential to be a very simple and robust space-reactor power system. Gas-cooled reactors have been built and operated in the past, but very few have been coupled directly to the turbomachinery in this fashion. In this paper we describe the option for testing such a system with a small reactor and turbomachinery at Sandia National Laboratories. Sandia currently operates the Annular Core Research Reactor (ACRR) at steady-state powers up to 4 MW and has an adjacent facility with heavy shielding in which another reactor recently operated. Sandia also has a closed-Brayton-Cycle test bed with a converted commercial turbomachinery unit that is rated for up to 30 kWe of power. It is proposed to construct a small experimental gas-cooled reactor core and attach this via ducting to the CBC turbomachinery for cooling and electricity production. Calculations suggest that such a unit could produce about 20 kWe, which would be a good power level for initial surface power units on the Moon or Mars. The intent of this experiment is to demonstrate the stable start-up and operation of such a system. Of particular interest is the effect of a negative temperature power coefficient as the initially cold Brayton gas passes through the core during startup or power changes. Sandia's dynamic model for such a system would be compared with the performance data. This paper describes the neutronics, heat transfer, and cycle dynamics of this proposed system. Safety and radiation issues are presented. The views expressed in this document are those of the author and do not necessarily reflect agreement by the government.

The use of dual mode thermionic reactors in supporting Earth orbital and space exploration missions

NASA Astrophysics Data System (ADS)

Zubrin, Robert M.; Sulmeisters, Tal K.

1993-01-01

Missions requiring large amounts of electric power to support their payload functions can be enabled through the employment of nuclear electric power reactors, which in some cases can also assist the mission by making possible the employment of high specific impulse electric propulsion. However it is found that the practicality and versality of using a power reactor to provide advanced propulsion is enormously enhanced if the reactor is configured in such a way to allow it to generate a certain amount of direct thrust as well. The use of such a system allows the creation of a common bus upper stage that can provide both high power and high impulse (with short orbit transfer times). It is shown that such a system, termed an Integral Power and Propulsion Stage (IPAPS), is optimal for supporting many Earth, Lunar, planetary and asteroidal observation, exploration, and communication support missions, and it is therefore recommended that the nuclear power reactor ultimately selected by the government for development and production be one that can be configured for such a function.

HOMOGENEOUS NUCLEAR POWER REACTOR

DOEpatents

King, L.D.P.

1959-09-01

A homogeneous nuclear power reactor utilizing forced circulation of the liquid fuel is described. The reactor does not require fuel handling outside of the reactor vessel during any normal operation including complete shutdown to room temperature, the reactor being selfregulating under extreme operating conditions and controlled by the thermal expansion of the liquid fuel. The liquid fuel utilized is a uranium, phosphoric acid, and water solution which requires no gus exhaust system or independent gas recombining system, thereby eliminating the handling of radioiytic gas.

10 CFR 100.1 - Purpose.

Code of Federal Regulations, 2011 CFR

2011-01-01

... COMMISSION (CONTINUED) REACTOR SITE CRITERIA § 100.1 Purpose. (a) The purpose of this part is to establish approval requirements for proposed sites for stationary power and testing reactors subject to part 50 or part 52 of this chapter. (b) There exists a substantial base of knowledge regarding power reactor...

Degradation of TCE using sequential anaerobic biofilm and aerobic immobilized bed reactor

NASA Technical Reports Server (NTRS)

Chapatwala, Kirit D.; Babu, G. R. V.; Baresi, Larry; Trunzo, Richard M.

1995-01-01

Bacteria capable of degrading trichloroethylene (TCE) were isolated from contaminated wastewaters and soil sites. The aerobic cultures were identified as Pseudomonas aeruginosa (four species) and Pseudomonas fluorescens. The optimal conditions for the growth of aerobic cultures were determined. The minimal inhibitory concentration values of TCE for Pseudomonas sps. were also determined. The aerobic cells were immobilized in calcium alginate in the form of beads. Degradation of TCE by the anaerobic and dichloroethylene (DCE) by aerobic cultures was studied using dual reactors - anaerobic biofilm and aerobic immobilized bed reactor. The minimal mineral salt (MMS) medium saturated with TCE was pumped at the rate of 1 ml per hour into the anaerobic reactor. The MMS medium saturated with DCE and supplemented with xylenes and toluene (3 ppm each) was pumped at the rate of 1 ml per hour into the fluidized air-uplift-type reactor containing the immobilized aerobic cells. The concentrations of TCE and DCE and the metabolites formed during their degradation by the anaerobic and aerobic cultures were monitored by GC. The preliminary study suggests that the anaerobic and aerobic cultures of our isolates can degrade TCE and DCE.

Satellite nuclear power station: An engineering analysis

NASA Technical Reports Server (NTRS)

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

1973-01-01

A nuclear-MHD power plant system which uses a compact non-breeder reactor to produce power in the multimegawatt range is analyzed. It is shown that, operated in synchronous orbit, the plant would transmit power safely to the ground by a microwave beam. Fuel reprocessing would take place in space, and no radioactive material would be returned to earth. Even the effect of a disastrous accident would have negligible effect on earth. A hydrogen moderated gas core reactor, or a colloid-core, or NERVA type reactor could also be used. The system is shown to approach closely the ideal of economical power without pollution.

TRICHLOROETHYLENE SORPTION AND OXIDATION USING A DUAL FUNCTION SORBENT/CATALYST IN A FALLING FURNACE REACTOR

EPA Science Inventory

A dual function medium (Cr-ZSM-5), capable of physisorbing trichloroethylene (TCE) at ambient temperature and catalytically oxidizing it at elevated temperature (-350 degrees C) was utilized in a novel continuous falling furnace reactor system to store and periodically destroy t...

ANALYSIS OF AN AEROBIC FLUIDIZED BED REACTOR DEGRADING MTBE AND BTEX AT REDUCED EBCTS

EPA Science Inventory

The purpose of this study was to investigate the biodegradation of MTBE and BTEX using a fluidized bed reactor (FBR) with granular activated carbon (GAC) as a biological attachment medium. Batch experiments were run to analyze the MTBE and TBA degradation kinetics of the culture ...

Reactor component automatic grapple

DOEpatents

Greenaway, Paul R.

1982-01-01

A grapple for handling nuclear reactor components in a medium such as liquid sodium which, upon proper seating and alignment of the grapple with the component as sensed by a mechanical logic integral to the grapple, automatically seizes the component. The mechanical logic system also precludes seizure in the absence of proper seating and alignment.

Reactor monitoring using antineutrino detectors

NASA Astrophysics Data System (ADS)

Bowden, N. S.

2011-08-01

Nuclear reactors have served as the antineutrino source for many fundamental physics experiments. The techniques developed by these experiments make it possible to use these weakly interacting particles for a practical purpose. The large flux of antineutrinos that leaves a reactor carries information about two quantities of interest for safeguards: the reactor power and fissile inventory. Measurements made with antineutrino detectors could therefore offer an alternative means for verifying the power history and fissile inventory of a reactor as part of International Atomic Energy Agency (IAEA) and/or other reactor safeguards regimes. Several efforts to develop this monitoring technique are underway worldwide.

The results of systems tests of the 500 kV busbar controllable shunting reactor in the Tavricheskaya substation

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

Gusev, S. I.; Karpov, V. N.; Kiselev, A. N.

2009-09-15

The results of systems tests of the 500 kV busbar magnetization-controllable shunting reactor (CSR), set up in the Tavricheskaya substation, including measurements of the quality of the electric power, the harmonic composition of the network currents of the reactor for different values of the reactive power consumed, the determination of the regulating characteristics of the reactor, the speed of response of the shunting reactor in the current and voltage stabilization modes, and also the operation of the reactor under dynamic conditions for different perturbations, are presented. The results obtained are analyzed.

Feasibility Study of a Nuclear-Stirling Power Plant for the Jupiter Icy Moons Orbiter

NASA Astrophysics Data System (ADS)

Schmitz, Paul C.; Schreiber, Jeffrey G.; Penswick, L. Barry

2005-02-01

NASA is undertaking the design of a new spacecraft to explore the planet Jupiter and its three moons Calisto, Ganymede and Europa. This proposed mission, known as Jupiter Icy Moons Orbiter (JIMO) would use a nuclear reactor and an associated electrical generation system (Reactor Power Plant - RPP) to provide power to the spacecraft. The JIMO spacecraft is envisioned to use this power for science and communications as well as Electric Propulsion (EP). Among other potential power-generating concepts, previous studies have considered Thermoelectric and Brayton power conversion systems, coupled to a liquid metal reactor for the JIMO mission. This paper will explore trades in system mass and radiator area for a nuclear reactor power conversion system, however this study will focus on Stirling power conversion. Stirling convertors have a long heritage operating in both power generation and the cooler industry, and are currently in use in a wide variety of applications. The Stirling convertor modeled in this study is based upon the Component Test Power Convertor design that was designed and operated successfully under the Civil Space Technology Initiative for use with the SP-100 nuclear reactor in the 1980's and early 1990's. The baseline RPP considered in this study consists of four dual-opposed Stirling convertors connected to the reactor by a liquid lithium loop. The study design is such that two of the four convertors would operate at any time to generate the 100 kWe while the others are held in reserve. For this study the Stirling convertors hot-side temperature is 1050 K, would operate at a temperature ratio of 2.4 for a minimum mass system and would have a system efficiency of 29%. The Stirling convertor would generate high voltage (400 volt), 100 Hz single phase AC that is supplied to the Power Management and Distribution system. The waste heat is removed from the Stirling convertors by a flowing liquid sodium-potassium eutectic and then rejected by a shared radiator. The radiator consists of two coplanar wings, which would be deployed after the reactor is in space. For this study design, the radiators would be located behind the conical radiation shield of the reactor and fan out as the radiator's distance from the reactor increases. System trades were performed to vary cycle state point temperatures and convertor design as well as power output. Other redundancy combinations were considered to understand the affects of convertor size and number of spares to the system mass.

Feasibility Study of a Nuclear-Stirling Power Plant for the Jupiter Icy Moons Orbiter

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

Schmitz, Paul C.; Schreiber, Jeffrey G.; Penswick, L. Barry

2005-02-06

NASA is undertaking the design of a new spacecraft to explore the planet Jupiter and its three moons Calisto, Ganymede and Europa. This proposed mission, known as Jupiter Icy Moons Orbiter (JIMO) would use a nuclear reactor and an associated electrical generation system (Reactor Power Plant - RPP) to provide power to the spacecraft. The JIMO spacecraft is envisioned to use this power for science and communications as well as Electric Propulsion (EP). Among other potential power-generating concepts, previous studies have considered Thermoelectric and Brayton power conversion systems, coupled to a liquid metal reactor for the JIMO mission. This papermore » will explore trades in system mass and radiator area for a nuclear reactor power conversion system, however this study will focus on Stirling power conversion. Stirling convertors have a long heritage operating in both power generation and the cooler industry, and are currently in use in a wide variety of applications. The Stirling convertor modeled in this study is based upon the Component Test Power Convertor design that was designed and operated successfully under the Civil Space Technology Initiative for use with the SP-100 nuclear reactor in the 1980's and early 1990's. The baseline RPP considered in this study consists of four dual-opposed Stirling convertors connected to the reactor by a liquid lithium loop. The study design is such that two of the four convertors would operate at any time to generate the 100 kWe while the others are held in reserve. For this study the Stirling convertors hot-side temperature is 1050 K, would operate at a temperature ratio of 2.4 for a minimum mass system and would have a system efficiency of 29%. The Stirling convertor would generate high voltage (400 volt), 100 Hz single phase AC that is supplied to the Power Management and Distribution system. The waste heat is removed from the Stirling convertors by a flowing liquid sodium-potassium eutectic and then rejected by a shared radiator. The radiator consists of two coplanar wings, which would be deployed after the reactor is in space. For this study design, the radiators would be located behind the conical radiation shield of the reactor and fan out as the radiator's distance from the reactor increases. System trades were performed to vary cycle state point temperatures and convertor design as well as power output. Other redundancy combinations were considered to understand the affects of convertor size and number of spares to the system mass.« less

Measurement instruments for automatically monitoring the water chemistry of reactor coolant at nuclear power stations equipped with VVER reactors. Selection of measurement instruments and experience gained from their operation at Russian and foreign NPSs

NASA Astrophysics Data System (ADS)

Ivanov, Yu. A.

2007-12-01

An analytical review is given of Russian and foreign measurement instruments employed in a system for automatically monitoring the water chemistry of the reactor coolant circuit and used in the development of projects of nuclear power stations equipped with VVER-1000 reactors and the nuclear station project AES 2006. The results of experience gained from the use of such measurement instruments at nuclear power stations operating in Russia and abroad are presented.

A study of increasing radical density and etch rate using remote plasma generator system

NASA Astrophysics Data System (ADS)

Lee, Jaewon; Kim, Kyunghyun; Cho, Sung-Won; Chung, Chin-Wook

2013-09-01

To improve radical density without changing electron temperature, remote plasma generator (RPG) is applied. Multistep dissociation of the polyatomic molecule was performed using RPG system. RPG is installed to inductively coupled type processing reactor; electrons, positive ions, radicals and polyatomic molecule generated in RPG and they diffused to processing reactor. The processing reactor dissociates the polyatomic molecules with inductively coupled power. The polyatomic molecules are dissociated by the processing reactor that is operated by inductively coupled power. Therefore, the multistep dissociation system generates more radicals than single-step system. The RPG was composed with two cylinder type inductively coupled plasma (ICP) using 400 kHz RF power and nitrogen gas. The processing reactor composed with two turn antenna with 13.56 MHz RF power. Plasma density, electron temperature and radical density were measured with electrical probe and optical methods.

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.

10 CFR 72.2 - Scope.

Code of Federal Regulations, 2011 CFR

2011-01-01

... RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Provisions § 72.2 Scope. (a) Except..., packaging, and possession of: (1) Power reactor spent fuel to be stored in a complex that is designed and constructed specifically for storage of power reactor spent fuel aged for at least one year, other radioactive...

10 CFR 72.2 - Scope.

Code of Federal Regulations, 2010 CFR

2010-01-01

... RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Provisions § 72.2 Scope. (a) Except..., packaging, and possession of: (1) Power reactor spent fuel to be stored in a complex that is designed and constructed specifically for storage of power reactor spent fuel aged for at least one year, other radioactive...

77 FR 41206 - Guidelines for Preparing and Reviewing Licensing Applications for Instrumentation and Control...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-07-12

... Applications for Instrumentation and Control Upgrades for Non-Power Reactors AGENCY: Nuclear Regulatory... (NRC or the Commission) is requesting public comment on Chapter 7, Section 7.3, Reactor Control System...-Power Reactors: Format and Content,'' for instrumentation and control (I&C) upgrades and NUREG-1537...

75 FR 62610 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Plant...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-10-12

...: Draft Regulatory Guide DG-1237, ``Guidance on Making Changes to Emergency Plans for Nuclear Power Reactors,'' Interim Staff Guidance (ISG) NSIR/DPR-ISG-01, ``Emergency Planning for Nuclear Power Plants... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS...

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.

77 FR 3009 - Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2012-01-20

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0010] Knowledge and Abilities Catalog for Nuclear Power... comment a draft NUREG, NUREG-2104, Revision 0, ``Knowledge and Abilities Catalog for Nuclear Power Plant... developed using this Catalog along with the Operator Licensing Examination Standards for Power Reactors...

75 FR 76498 - Firstenergy Nuclear Operating Company, Davis-Besse Nuclear Power Station; Environmental...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-12-08

... Company, Davis-Besse Nuclear Power Station; Environmental Assessment And Finding of No Significant Impact... operation of the Davis-Besse Nuclear Power Station, Unit 1 (DBNPS), located in Ottawa County, Ohio. In... the reactor coolant pressure boundary of light-water nuclear power reactors provide adequate margins...

78 FR 24438 - Evaluations of Explosions Postulated To Occur at Nearby Facilities and on Transportation Routes...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-04-25

... Nearby Facilities and on Transportation Routes Near Nuclear Power Plants AGENCY: Nuclear Regulatory... Nearby Facilities and on Transportation Routes Near Nuclear Power Plants.'' This regulatory guide describes for applicants seeking nuclear power reactor licenses and licensees of nuclear power reactors...

REVIEW OF POWER AND HEAT REACTOR DESIGNS. Domestic and Foreign

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

Appleby, E.R., comp

1963-10-01

Unclassified information from domestic and foreign literature from January 1952 through September 1963 is compiled. Design characteristics and current information on the status of the individual designs are given, along with references for the associated literature. SNAP systems, proposed reactors, and chemonuclear and test reactors with characteristics similar to power reactors are included. The designs are indexed by name, location, type, and some special characteristics. (D.C.W.)

PBF Reactor Building (PER620). Camera is in cab of electricpowered ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

PBF Reactor Building (PER-620). Camera is in cab of electric-powered rail crane and facing east. Reactor pit and storage canal have been shaped. Floors for wings on east and west side are above and below reactor in view. Photographer: Larry Page. Date: August 23, 1967. INEEL negative no. 67-4403 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Small Reactor for Deep Space Exploration

ScienceCinema

none,

2018-06-06

This is the first demonstration of a space nuclear reactor system to produce electricity in the United States since 1965, and an experiment demonstrated the first use of a heat pipe to cool a small nuclear reactor and then harvest the heat to power a Stirling engine at the Nevada National Security Site's Device Assembly Facility confirms basic nuclear reactor physics and heat transfer for a simple, reliable space power system.

Characteristics and Dose Levels for Spent Reactor Fuels

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

Coates, Cameron W

2007-01-01

Current guidance considers highly radioactive special nuclear materials to be those materials that, unshielded, emit a radiation dose [rate] measured at 1 m which exceeds 100 rem/h. Smaller, less massive fuel assemblies from research reactors can present a challenge from the point of view of self protection because of their size (lower dose, easier to handle) and the desirability of higher enrichments; however, a follow-on study to cross-compare dose trends of research reactors and power reactors was deemed useful to confirm/verify these trends. This paper summarizes the characteristics and dose levels of spent reactor fuels for both research reactors andmore » power reactors and extends previous studies aimed at quantifying expected dose rates from research reactor fuels worldwide.« less

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.

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

Flow reversal power limit for the HFBR

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

Cheng, L.Y.; Tichler, P.R.

The High Flux Beam Reactor (HFBR) is a pressurized heavy water moderated and cooled research reactor that began operation at 40 MW. The reactor was subsequently upgraded to 60 MW and operated at that level for several years. The reactor undergoes a buoyancy-driven reversal of flow in the reactor core following certain postulated accidents. Questions which were raised about the afterheat removal capability during the flow reversal transition led to a reactor shutdown and subsequent resumption of operation at a reduced power of 30 MW. An experimental and analytical program to address these questions is described in this report. Themore » experiments were single channel flow reversal tests under a range of conditions. The analytical phase involved simulations of the tests to benchmark the physical models and development of a criterion for dryout. The criterion is then used in simulations of reactor accidents to determine a safe operating power level. It is concluded that the limit on the HFBR operating power with respect to the issue of flow reversal is in excess of 60 MW. Direct use of the experimental results and an understanding of the governing phenomenology supports this conclusion.« less

Direct-Drive Gas-Cooled Reactor Power System: Concept and Preliminary Testing

NASA Technical Reports Server (NTRS)

Wright, S. A.; Lipinski, R. J.; Godfroy, T. J.; Bragg-Sitton, S. M.; VanDyke, M. K.

2002-01-01

This paper describes the concept and preliminary component testing of a gas-cooled, UN-fueled, pin-type reactor which uses He/Xe gas that goes directly into a recuperated Brayton system to produce electricity for nuclear electric propulsion. This Direct-Drive Gas-Cooled Reactor (DDG) is designed to be subcritical under water or wet- sand immersion in case of a launch accident. Because the gas-cooled reactor can directly drive the Brayton turbomachinery, it is possible to configure the system such that there are no external surfaces or pressure boundaries that are refractory metal, even though the gas delivered to the turbine is 1144 K. The He/Xe gas mixture is a good heat transport medium when flowing, and a good insulator when stagnant. Judicious use of stagnant cavities as insulating regions allows transport of the 1144-K gas while keeping all external surfaces below 900 K. At this temperature super-alloys (Hastelloy or Inconel) can be used instead of refractory metals. Super-alloys reduce the technology risk because they are easier to fabricate than refractory metals, we have a much more extensive knowledge base on their characteristics, and, because they have a greater resistance to oxidation, system testing is eased. The system is also relatively simple in its design: no additional coolant pumps, heat exchanger, or freeze-thaw systems are required. Key to success of this concept is a good knowledge of the heat transfer between the fuel pins and the gas, as well as the pressure drop through the system. This paper describes preliminary testing to obtain this key information, as well as experience in demonstrating electrically heated testing of simulated reactor components.

JPRS Report, Science & Technology, China: Energy.

DTIC Science & Technology

1992-03-30

breeder reactors should become...the primary type of reactors . In developing breeder reactors , we should follow the path of using metal fuel. Breeder reactors give us more time to...first reactor used for power generation was a fast reactor : the " Breeder 1" reactor at the Idaho National Reactor Test Center which was used to

Uranium Dioxides and Debris Fragments Released to the Environment with Cesium-Rich Microparticles from the Fukushima Daiichi Nuclear Power Plant.

PubMed

Ochiai, Asumi; Imoto, Junpei; Suetake, Mizuki; Komiya, Tatsuki; Furuki, Genki; Ikehara, Ryohei; Yamasaki, Shinya; Law, Gareth T W; Ohnuki, Toshihiko; Grambow, Bernd; Ewing, Rodney C; Utsunomiya, Satoshi

2018-03-06

Trace U was released from the Fukushima Daiichi Nuclear Power Plant (FDNPP) during the meltdowns, but the speciation of the released components of the nuclear fuel remains unknown. We report, for the first time, the atomic-scale characteristics of nanofragments of the nuclear fuels that were released from the FDNPP into the environment. Nanofragments of an intrinsic U-phase were discovered to be closely associated with radioactive cesium-rich microparticles (CsMPs) in paddy soils collected ∼4 km from the FDNPP. The nanoscale fuel fragments were either encapsulated by or attached to CsMPs and occurred in two different forms: (i) UO 2+X nanocrystals of ∼70 nm size, which are embedded into magnetite associated with Tc and Mo on the surface and (ii) Isometric (U,Zr)O 2+X nanocrystals of ∼200 nm size, with the U/(U+Zr) molar ratio ranging from 0.14 to 0.91, with intrinsic pores (∼6 nm), indicating the entrapment of vapors or fission-product gases during crystallization. These results document the heterogeneous physical and chemical properties of debris at the nanoscale, which is a mixture of melted fuel and reactor materials, reflecting the complex thermal processes within the FDNPP reactor during meltdown. Still CsMPs are an important medium for the transport of debris fragments into the environment in a respirable form.

Pratt & Whitney ESCORT derivative for mars surface power

NASA Astrophysics Data System (ADS)

Feller, Gerald J.; Joyner, Russell

1999-01-01

The purpose of this paper is to address the applicability of a common reactor system design from the Pratt & Whitney ESCORT nuclear thermal rocket engine concept to support current NASA mars surface-based power requirements. The ESCORT is a bimodal engine capable of supporting a wide range of propulsive thermal and vehicle electrical power requirements. The ESCORT engine is powered by a fast-spectrum beryllium-reflected CERMET-fueled nuclear reactor. In addition to an expander cycle propulsive mode, the ESCORT is capable of operating in an electrical power mode. In this mode, the reactor is used to heat a mixture of helium and xenon to drive a closed-loop Brayton cycle in order to generate electrical energy. Recent Design Reference Mission requirements (DRM) from NASA Johnson Space Center and NASA Lewis Research Center studies in 1997 and 1998 have detailed upgraded requirements for potential mars transfer missions. The current NASA DRM requires a nuclear thermal propulsion system capable of delivering total mission requirements of 200170 N (45000 lbf) thrust and 50 kWe of spacecraft electrical power. Additionally, these requirements detailed a surface power system capable of providing approximately 160 kW of electrical energy over an approximate 10 year period within a given weight and volume envelope. Current NASA studies use a SP-100 reactor (0.8 MT) and a NERVA derivative (1.6 MT) as baseline systems. A mobile power cart of approximate dimensions 1.7 m×4.5 m×4.4 m has been conceptualized to transport the reactor power system on the Mars Surface. The 63.25 cm diameter and 80.25 cm height of the ESCORT and its 1.3 MT of weight fit well within the current weight and volume target range of the NASA DRM requirements. The modifications required to the ESCORT reactor system to support this upgraded electrical power requirements along with operation in the Martian atmospheric conditions are addressed in this paper. Sufficient excess reactivity and burnup capability were designed into the ESCORT reactor system to support these upgraded requirements. Only slight modifications to reactor hardware were required to address any environmental considerations. These modifications involved sealing any refractory metal alloy components from the CO2 in the Martian Atmosphere. Also, the Brayton cycle Power Conversion Unit (PCU) hardware was modified to support the upgraded requirements. This paper discusses the design analysis performed and provides information on the final common reactor concept to be used on the Mars surface to support manned missions.

The United States Naval Nuclear Propulsion Program - Over 151 Million Miles Safely Steamed on Nuclear Power

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

None, None

NNSA’s third mission pillar is supporting the U.S. Navy’s ability to protect and defend American interests across the globe. The Naval Reactors Program remains at the forefront of technological developments in naval nuclear propulsion and ensures a commanding edge in warfighting capabilities by advancing new technologies and improvements in naval reactor performance and reliability. In 2015, the Naval Nuclear Propulsion Program pioneered advances in nuclear reactor and warship design – such as increasing reactor lifetimes, improving submarine operational effectiveness, and reducing propulsion plant crewing. The Naval Reactors Program continued its record of operational excellence by providing the technical expertise requiredmore » to resolve emergent issues in the Nation’s nuclear-powered fleet, enabling the Fleet to safely steam more than two million miles. Naval Reactors safely maintains, operates, and oversees the reactors on the Navy’s 82 nuclear-powered warships, constituting more than 45 percent of the Navy’s major combatants.« less

Megawatt Class Nuclear Space Power Systems (MCNSPS) conceptual design and evaluation report. Volume 1: Objectives, summary results and introduction

NASA Technical Reports Server (NTRS)

Wetch, J. R.

1988-01-01

The objective was to determine which reactor, conversion, and radiator technologies would best fulfill future Megawatt Class Nuclear Space Power System Requirements. Specifically, the requirement was 10 megawatts for 5 years of full power operation and 10 years systems life on orbit. A variety of liquid metal and gas cooled reactors, static and dynamic conversion systems, and passive and dynamic radiators were considered. Four concepts were selected for more detailed study. The concepts are: a gas cooled reactor with closed cycle Brayton turbine-alternator conversion with heat pipe and pumped tube-fin heat rejection; a lithium cooled reactor with a free piston Stirling engine-linear alternator and a pumped tube-fin radiator; a lithium cooled reactor with potassium Rankine turbine-alternator and heat pipe radiator; and a lithium cooled incore thermionic static conversion reactor with a heat pipe radiator. The systems recommended for further development to meet a 10 megawatt long life requirement are the lithium cooled reactor with the K-Rankine conversion and heat pipe radiator, and the lithium cooled incore thermionic reactor with heat pipe radiator.

10 CFR 72.22 - Contents of application: General and financial information.

Code of Federal Regulations, 2011 CFR

2011-01-01

... INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN... of spent fuel, high-level radioactive waste, and/or reactor-related GTCC waste from storage. (f) Each applicant for a license under this part to receive, transfer, and possess power reactor spent fuel, power...

76 FR 54808 - Agency Information Collection Activities: Submission for the Office of Management and Budget (OMB...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-09-02

... the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste and Reactor-Related... receive, transfer, package and possess power reactor spent fuel, high-level waste, and other radioactive..., package, and possess power reactor spent fuel and high-level radioactive waste, and other associated...

76 FR 63668 - Guidelines for Preparing and Reviewing Licensing Applications for the Production of Radioisotopes

Federal Register 2010, 2011, 2012, 2013, 2014

2011-10-13

... Licensing of Non-Power Reactors: Format and Content,'' for the Production of Radioisotopes and NUREG-1537, part 2, ``Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors... production facility and the Research and Test Reactor Licensing Branch (PRLB) of the Division of Policy and...

10 CFR 50.44 - Combustible gas control for nuclear power reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 1 2014-01-01 2014-01-01 false Combustible gas control for nuclear power reactors. 50.44... FACILITIES Standards for Licenses, Certifications, and Regulatory Approvals § 50.44 Combustible gas control... capability for ensuring a mixed atmosphere. (2) Combustible gas control. (i) All boiling water reactors with...

10 CFR 50.44 - Combustible gas control for nuclear power reactors.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Combustible gas control for nuclear power reactors. 50.44... FACILITIES Standards for Licenses, Certifications, and Regulatory Approvals § 50.44 Combustible gas control... capability for ensuring a mixed atmosphere. (2) Combustible gas control. (i) All boiling water reactors with...

10 CFR 50.44 - Combustible gas control for nuclear power reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 1 2012-01-01 2012-01-01 false Combustible gas control for nuclear power reactors. 50.44... FACILITIES Standards for Licenses, Certifications, and Regulatory Approvals § 50.44 Combustible gas control... capability for ensuring a mixed atmosphere. (2) Combustible gas control. (i) All boiling water reactors with...

76 FR 32240 - Advisory Committee on Reactor Safeguards (ACRS) Meeting on the ACRS Subcommittee on Power Uprates

Federal Register 2010, 2011, 2012, 2013, 2014

2011-06-03

... Expanded Operating Domains-Power Distribution Validation and Pin-by-Pin Gamma Scan). The Subcommittee will... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting on the ACRS Subcommittee on Power Uprates Notice of Meeting The ACRS Subcommittee on Power Uprates will hold a meeting on...

Preliminary assessment of high power, NERVA-class dual-mode space nuclear propulsion and power systems

NASA Astrophysics Data System (ADS)

Buksa, John J.; Kirk, William L.; Cappiello, Michael W.

A preliminary assessment of the technical feasibility and mass competitiveness of a dual-mode nuclear propulsion and power system based on the NERVA rocket engine has been completed. Results indicate that the coupling of the Rover reactor to a direct Brayton power conversion system can be accomplished through a number of design features. Furthermore, based on previously published and independently calculated component masses, the dual-mode system was found to have the potential to be mass competitive with propulsion/power systems that use separate reactors. The uncertainties of reactor design modification and shielding requirements were identified as important issues requiring future investigation.

POWER-BURST FACILITY (PBF) CONCEPTUAL DESIGN

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

Wasserman, A.A.; Johnson, S.O.; Heffner, R.E.

1963-06-21

A description is presented of the conceptual design of a high- performance, pulsed reactor called the Power Burst Facility (PBF). This reactor is designed to generate power bursts with initial asymptotic periods as short as 1 msec, producing energy releases large enough to destroy entire fuel subassemblies placed in a capsule or flow loop mounted in the reactor, all without damage to the reactor itself. It will be used primarily to evaluate the consequences and hazards of very rapid destructive accidents in reactors representing the entire range of current nuclear technology as applied to power generation, propulsion, and testing. Itmore » will also be used to carry out detailed studies of nondestructive reactivity feedback mechanisms in the shortperiod domain. The facility was designed to be sufficiently flexible to accommodate future cores of even more advanced design. The design for the first reactor core is based upon proven technology; hence, completion of the final design of this core will involve no significant development delays. Construction of the PBF is proposed to begin in September 1984, and is expected to take approximately 20 months to complete. (auth)« less

Analysis and Down Select of Flow Passages for Thermal Hydraulic Testing of a SNAP Derived Reactor

NASA Technical Reports Server (NTRS)

Godfroy, T. J.; Sadasivan, P.; Masterson, S.

2007-01-01

As past of the Vision for Space Exploration, man will return to the moon. To enable safe and productive time on the lunar surface will require adequate power resources. To provide the needed power and to give mission planners all landing site possibilities, including a permanently dark crater, a nuclear reactor provides the most options. Designed to be l00kWt providing approx. 25kWe this power plants would be very effective in delivering dependable, site non-specific power to crews or robotic missions on the lunar surface. An affordable reference reactor based upon the successful SNAP program of the 1960's and early 1970's has been designed by Los Alamos National Laboratory that will meet such a requirement. Considering current funding, environmental, and schedule limitations this lunar surface power reactor will be tested using non-nuclear simulators to simulate the heat from fission reactions. Currently a 25kWe surface power SNAP derivative reactor is in the early process of design and testing with collaboration between Los Alamos National Laboratory, Idaho National Laboratory, Glenn Research Center, Marshall Space Flight Center, and Sandia National Laboratory to ensure that this new design is affordable and can be tested using non-nuclear methods as have proven so effective in the past. This paper will discuss the study and down selection of a flow passage concept for a approx. 25kWe lunar surface power reactor. Several different flow passages designs were evaluated using computational fluid dynamics to determine pressure drop and a structural assessment to consider thermal and stress of the passage walls. The reactor design basis conditions are discussed followed by passage problem setup and results for each concept. A recommendation for passage design is made with rationale for selection.

10 CFR 100.1 - Purpose.

Code of Federal Regulations, 2010 CFR

2010-01-01

... approval requirements for proposed sites for stationary power and testing reactors subject to part 50 or part 52 of this chapter. (b) There exists a substantial base of knowledge regarding power reactor... approach incorporates the appropriate standards and criteria for approval of stationary power and testing...

Nuclear Engineering Technologists in the Nuclear Power Era

ERIC Educational Resources Information Center

Wang, C. H.; And Others

1974-01-01

Describes manpower needs in nuclear engineering in the areas of research and development, architectural engineering and construction supervision, power reactor operations, and regulatory tasks. Outlines a suitable curriculum to prepare students for the tasks related to construction and operation of power reactors. (GS)

Computer modeling and simulators as part of university training for NPP operating personnel

NASA Astrophysics Data System (ADS)

Volman, M.

2017-01-01

This paper considers aspects of a program for training future nuclear power plant personnel developed by the NPP Department of Ivanovo State Power Engineering University. Computer modeling is used for numerical experiments on the kinetics of nuclear reactors in Mathcad. Simulation modeling is carried out on the computer and full-scale simulator of water-cooled power reactor for the simulation of neutron-physical reactor measurements and the start-up - shutdown process.

The Sustainable Nuclear Future: Fission and Fusion E.M. Campbell Logos Technologies

NASA Astrophysics Data System (ADS)

Campbell, E. Michael

2010-02-01

Global industrialization, the concern over rising CO2 levels in the atmosphere and other negative environmental effects due to the burning of hydrocarbon fuels and the need to insulate the cost of energy from fuel price volatility have led to a renewed interest in nuclear power. Many of the plants under construction are similar to the existing light water reactors but incorporate modern engineering and enhanced safety features. These reactors, while mature, safe and reliable sources of electrical power have limited efficiency in converting fission power to useful work, require significant amounts of water, and must deal with the issues of nuclear waste (spent fuel), safety, and weapons proliferation. If nuclear power is to sustain its present share of the world's growing energy needs let alone displace carbon based fuels, more than 1000 reactors will be needed by mid century. For this to occur new reactors that are more efficient, versatile in their energy markets, require minimal or no water, produce less waste and more robust waste forms, are inherently safe and minimize proliferation concerns will be necessary. Graphite moderated, ceramic coated fuel, and He cooled designs are reactors that can satisfy these requirements. Along with other generation IV fast reactors that can further reduce the amounts of spent fuel and extend fuel resources, such a nuclear expansion is possible. Furthermore, facilities either in early operations or under construction should demonstrate the next step in fusion energy development in which energy gain is produced. This demonstration will catalyze fusion energy development and lead to the ultimate development of the next generation of nuclear reactors. In this presentation the role of advanced fission reactors and future fusion reactors in the expansion of nuclear power will be discussed including synergies with the existing worldwide nuclear fleet. )

Experiences in utilization of research reactors in Yugoslavia

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

Copic, M.; Gabrovsek, Z.; Pop-Jordanov, J.

1971-06-15

The nuclear institutes in Yugoslavia possess three research reactors. Since 1958, two heavy-water reactors have been in operation at the 'Boris Kidric' Institute, a zero-power reactor RB and a 6. 5-MW reactor RA. At the Jozef Stefan Institute, a 250-kW TRIGA Mark II reactor has been operating since 1966. All reactors are equipped with the necessary experimental facilities. The main activities based on these reactors are: (1) fundamental research in solid-state and nuclear physics; (2) R and D activities related to nuclear power program; and (3) radioisotope production. In fundamental physics, inelastic neutron scattering and diffraction phenomena are studied bymore » means of the neutron beam tubes and applied to investigations of the structures of solids and liquids. Valuable results are also obtained in n - γ reaction studies. Experiments connected with the fuel -element development program, owing to the characteristics of the existing reactors, are limited to determination of the fuel element parameters, to studies on the purity of uranium, and to a small number of capsule irradiations. All three reactors are also used for the verification of different methods applied in the analysis of power reactors, particularly concerning neutron flux distributions, the optimization of reactor core configurations and the shielding effects. An appreciable irradiation space in the reactors is reserved for isotope production. Fruitful international co-operation has been established in all these activities, on the basis of either bilateral or multilateral arrangements. The paper gives a critical analysis of the utilization of research reactors in a developing country such as Yugoslavia. The investments in and the operational costs of research reactors are compared with the benefits obtained in different areas of reactor application. The impact on the general scientific, technological and educational level in the country is also considered. In particular, an attempt is made ro envisage the role of research reactors in the promotion of nuclear power programs in relation to the size of the program, the competence of domestic industries and the degree of independence where fuel supply is concerned. (author)« less

Inherently Safe and Long-Life Fission Power System for Lunar Outposts

NASA Astrophysics Data System (ADS)

Schriener, T. M.; El-Genk, Mohamed S.

Power requirements for future lunar outposts, of 10's to 100's kWe, can be fulfilled using nuclear reactor power systems. In addition to the long life and operation reliability, safety is paramount in all phases, including fabrication and assembly, launch, emplacement below grade on the lunar surface, operation, post-operation decay heat removal and long-term storage and eventual retrieval. This paper introduces the Solid Core-Sectored Compact Reactor (SC-SCoRe) and power system with static components and no single point failures. They ensure reliable continuous operation for ~21 years and fulfill the safety requirements. The SC-SCoRe nominally generates 1.0 MWth at liquid NaK-56 coolant inlet and exit temperatures of 850 K and 900 K and the power system provides 38 kWe at high DC voltage using SiGe thermoelectric (TE) conversion assemblies. In case of a loss of coolant or cooling in a reactor core sector, the power system continues to operate; generating ~4 kWe to the outpost for emergency life support needs. The post-operation storage of the reactor below grade on the lunar surface is a safe and practical choice. The total radioactivity in the reactor drops from ~1 million Ci, immediately at shutdown, to below 164 Ci after 300 years of storage. At such time, the reactor is retrieved safely with no contamination or environmental concerns.

Multi-Megawatt Power System Trade Study

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

Longhurst, Glen Reed; Schnitzler, Bruce Gordon; Parks, Benjamin Travis

2001-11-01

As part of a larger task, the Idaho National Engineering and Environmental Laboratory (INEEL) was tasked to perform a trade study comparing liquid-metal cooled reactors having Rankine power conversion systems with gas-cooled reactors having Brayton power conversion systems. This report summarizes the approach, the methodology, and the results of that trade study. Findings suggest that either approach has the possibility to approach the target specific mass of 3-5 kg/kWe for the power system, though it appears either will require improvements to achieve that. Higher reactor temperatures have the most potential for reducing the specific mass of gas-cooled reactors but domore » not necessarily have a similar effect for liquid-cooled Rankine systems. Fuels development will be the key to higher reactor operating temperatures. Higher temperature turbines will be important for Brayton systems. Both replacing lithium coolant in the primary circuit with gallium and replacing potassium with sodium in the power loop for liquid systems increase system specific mass. Changing the feed pump turbine to an electric motor in Rankine systems has little effect. Key technologies in reducing specific mass are high reactor and radiator operating temperatures, low radiator areal density, and low turbine/generator system masses. Turbine/generator mass tends to dominate overall power system mass for Rankine systems. Radiator mass was dominant for Brayton systems.« less

Pyrolysis of cassava rhizome in a counter-rotating twin screw reactor unit.

PubMed

Sirijanusorn, Somsak; Sriprateep, Keartisak; Pattiya, Adisak

2013-07-01

A counter-rotating twin screw reactor unit was investigated for its behaviour in the pyrolysis of cassava rhizome biomass. Several parameters such as pyrolysis temperature in the range of 500-700°C, biomass particle size of <0.6mm, the use of sand as heat transfer medium, nitrogen flow rate of 4-10 L/min and nitrogen pressure of 1-3 bar were thoroughly examined. It was found that the pyrolysis temperature of 550°C could maximise the bio-oil yield (50 wt.%). The other optimum parameters for maximising the bio-oil yield were the biomass particle size of 0.250-0.425 mm, the nitrogen flow rate of 4 L/min and the nitrogen pressure of 2 bar. The use of the heat transfer medium could increase the bio-oil yield to a certain extent. Moreover, the water content of bio-oil produced with the counter-rotating twin screw reactor was relatively low, whereas the solids content was relatively high, compared to some other reactor configurations. Copyright © 2013 Elsevier Ltd. All rights reserved.

Wide-range structurally optimized channel for monitoring the certified power of small-core reactors

NASA Astrophysics Data System (ADS)

Koshelev, A. S.; Kovshov, K. N.; Ovchinnikov, M. A.; Pikulina, G. N.; Sokolov, A. B.

2016-12-01

The results of tests of a prototype version of a channel for monitoring the certified power of small-core reactors performed at the BR-K1 reactor at the All-Russian Scientific Research Institute of Experimental Physics are reported. An SNM-11 counter and commercial KNK-4 and KNK-3 compensated ion chambers were used as neutron detectors in the tested channel, and certified NCMM and CCMM measurement modules controlled by a PC with specialized software were used as measuring instruments. The specifics of metrological assurance of calibration of the channel in the framework of reactor power monitoring are discussed.

Wide-range structurally optimized channel for monitoring the certified power of small-core reactors

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

Koshelev, A. S., E-mail: alexsander.coshelev@yandex.ru; Kovshov, K. N.; Ovchinnikov, M. A.

The results of tests of a prototype version of a channel for monitoring the certified power of small-core reactors performed at the BR-K1 reactor at the All-Russian Scientific Research Institute of Experimental Physics are reported. An SNM-11 counter and commercial KNK-4 and KNK-3 compensated ion chambers were used as neutron detectors in the tested channel, and certified NCMM and CCMM measurement modules controlled by a PC with specialized software were used as measuring instruments. The specifics of metrological assurance of calibration of the channel in the framework of reactor power monitoring are discussed.

Neutron source, linear-accelerator fuel enricher and regenerator and associated methods

DOEpatents

Steinberg, Meyer; Powell, James R.; Takahashi, Hiroshi; Grand, Pierre; Kouts, Herbert

1982-01-01

A device for producing fissile material inside of fabricated nuclear elements so that they can be used to produce power in nuclear power reactors. Fuel elements, for example, of a LWR are placed in pressure tubes in a vessel surrounding a liquid lead-bismuth flowing columnar target. A linear-accelerator proton beam enters the side of the vessel and impinges on the dispersed liquid lead-bismuth columns and produces neutrons which radiate through the surrounding pressure tube assembly or blanket containing the nuclear fuel elements. These neutrons are absorbed by the natural fertile uranium-238 elements and are transformed to fissile plutonium-239. The fertile fuel is thus enriched in fissile material to a concentration whereby they can be used in power reactors. After use in the power reactors, dispensed depleted fuel elements can be reinserted into the pressure tubes surrounding the target and the nuclear fuel regenerated for further burning in the power reactor.

Flow reversal power limit for the HFBR

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

Cheng, Lap Y.; Tichler, P.R.

The High Flux Beam Reactor (HFBR) undergoes a buoyancy-driven reversal of flow in the reactor core following certain postulated accidents. Uncertainties about the afterheat removal capability during the flow reversal has limited the reactor operating power to 30 MW. An experimental and analytical program to address these uncertainties is described in this report. The experiments were single channel flow reversal tests under a range of conditions. The analytical phase involved simulations of the tests to benchmark the physical models and development of a criterion for dryout. The criterion is then used in simulations of reactor accidents to determine a safemore » operating power level. It is concluded that the limit on the HFBR operating power with respect to the issue of flow reversal is in excess of 60 MW.« less

Spherical torus fusion reactor

DOEpatents

Martin Peng, Y.K.M.

1985-10-03

The object of this invention is to provide a compact torus fusion reactor with dramatic simplification of plasma confinement design. Another object of this invention is to provide a compact torus fusion reactor with low magnetic field and small aspect ratio stable plasma confinement. In accordance with the principles of this invention there is provided a compact toroidal-type plasma confinement fusion reactor in which only the indispensable components inboard of a tokamak type of plasma confinement region, mainly a current conducting medium which carries electrical current for producing a toroidal magnet confinement field about the toroidal plasma region, are retained.

Low-power lead-cooled fast reactor loaded with MOX-fuel

NASA Astrophysics Data System (ADS)

Sitdikov, E. R.; Terekhova, A. M.

2017-01-01

Fast reactor for the purpose of implementation of research, education of undergraduate and doctoral students in handling innovative fast reactors and training specialists for atomic research centers and nuclear power plants (BRUTs) was considered. Hard neutron spectrum achieved in the fast reactor with compact core and lead coolant. Possibility of prompt neutron runaway of the reactor is excluded due to the low reactivity margin which is less than the effective fraction of delayed neutrons. The possibility of using MOX fuel in the BRUTs reactor was examined. The effect of Keff growth connected with replacement of natural lead coolant to 208Pb coolant was evaluated. The calculations and reactor core model were performed using the Serpent Monte Carlo code.

Safety and Environment aspects of Tokamak- type Fusion Power Reactor- An Overview

NASA Astrophysics Data System (ADS)

Doshi, Bharat; Reddy, D. Chenna

2017-04-01

Naturally occurring thermonuclear fusion reaction (of light atoms to form a heavier nucleus) in the sun and every star in the universe, releases incredible amounts of energy. Demonstrating the controlled and sustained reaction of deuterium-tritium plasma should enable the development of fusion as an energy source here on Earth. The promising fusion power reactors could be operated on the deuterium-tritium fuel cycle with fuel self-sufficiency. The potential impact of fusion power on the environment and the possible risks associated with operating large-scale fusion power plants is being studied by different countries. The results show that fusion can be a very safe and sustainable energy source. A fusion power plant possesses not only intrinsic advantages with respect to safety compared to other sources of energy, but also a negligible long term impact on the environment provided certain precautions are taken in its design. One of the important considerations is in the selection of low activation structural materials for reactor vessel. Selection of the materials for first wall and breeding blanket components is also important from safety issues. It is possible to fully benefit from the advantages of fusion energy if safety and environmental concerns are taken into account when considering the conceptual studies of a reactor design. The significant safety hazards are due to the tritium inventory and energetic neutron fluence induced activity in the reactor vessel, first wall components, blanket system etc. The potential of release of radioactivity under operational and accident conditions needs attention while designing the fusion reactor. Appropriate safety analysis for the quantification of the risk shall be done following different methods such as FFMEA (Functional Failure Modes and Effects Analysis) and HAZOP (Hazards and operability). Level of safety and safety classification such as nuclear safety and non-nuclear safety is very important for the FPR (Fusion Power Reactor). This paper describes an overview of safety and environmental merits of fusion power reactor, issues and design considerations and need for R&D on safety and environmental aspects of Tokamak type fusion reactor.

Developing the European Center of Competence on VVER-Type Nuclear Power Reactors

ERIC Educational Resources Information Center

Geraskin, Nikolay; Pironkov, Lyubomir; Kulikov, Evgeny; Glebov, Vasily

2017-01-01

This paper presents the results of the European educational projects CORONA and CORONA-II which are dedicated to preserving and further developing nuclear knowledge and competencies in the area of VVER-type nuclear power reactors technologies (Water-Water Energetic Reactor, WWER or VVER). The development of the European Center of Competence for…

MODERATOR ELEMENTS FOR UNIFORM POWER NUCLEAR REACTOR

DOEpatents

Balent, R.

1963-03-12

This patent describes a method of obtaining a flatter flux and more uniform power generation across the core of a nuclear reactor. The method comprises using moderator elements having differing moderating strength. The elements have an increasing amount of the better moderating material as a function of radial and/or axial distance from the reactor core center. (AEC)

76 FR 72007 - ZionSolutions, LLC; Zion Nuclear Power Station, Units 1 and 2; Exemption From Certain Security...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-11-21

...; Zion Nuclear Power Station, Units 1 and 2; Exemption From Certain Security Requirements 1.0 Background Zion Nuclear Power Station (ZNPS or Zion), Unit 1, is a Westinghouse 3250 MWt Pressurized Water Reactor... activities in nuclear power reactors against radiological sabotage,'' paragraph (b)(1) states, ``The licensee...

Adaptive Neural Network Algorithm for Power Control in Nuclear Power Plants

NASA Astrophysics Data System (ADS)

Masri Husam Fayiz, Al

2017-01-01

The aim of this paper is to design, test and evaluate a prototype of an adaptive neural network algorithm for the power controlling system of a nuclear power plant. The task of power control in nuclear reactors is one of the fundamental tasks in this field. Therefore, researches are constantly conducted to ameliorate the power reactor control process. Currently, in the Department of Automation in the National Research Nuclear University (NRNU) MEPhI, numerous studies are utilizing various methodologies of artificial intelligence (expert systems, neural networks, fuzzy systems and genetic algorithms) to enhance the performance, safety, efficiency and reliability of nuclear power plants. In particular, a study of an adaptive artificial intelligent power regulator in the control systems of nuclear power reactors is being undertaken to enhance performance and to minimize the output error of the Automatic Power Controller (APC) on the grounds of a multifunctional computer analyzer (simulator) of the Water-Water Energetic Reactor known as Vodo-Vodyanoi Energetichesky Reaktor (VVER) in Russian. In this paper, a block diagram of an adaptive reactor power controller was built on the basis of an intelligent control algorithm. When implementing intelligent neural network principles, it is possible to improve the quality and dynamic of any control system in accordance with the principles of adaptive control. It is common knowledge that an adaptive control system permits adjusting the controller’s parameters according to the transitions in the characteristics of the control object or external disturbances. In this project, it is demonstrated that the propitious options for an automatic power controller in nuclear power plants is a control system constructed on intelligent neural network algorithms.

Planetary surface reactor shielding using indigenous materials

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

Houts, Michael G.; Poston, David I.; Trellue, Holly R.

The exploration and development of Mars will require abundant surface power. Nuclear reactors are a low-cost, low-mass means of providing that power. A significant fraction of the nuclear power system mass is radiation shielding necessary for protecting humans and/or equipment from radiation emitted by the reactor. For planetary surface missions, it may be desirable to provide some or all of the required shielding from indigenous materials. This paper examines shielding options that utilize either purely indigenous materials or a combination of indigenous and nonindigenous materials.

ENGINEERING AND CONSTRUCTING THE HALLAM NUCLEAR POWER FACILITY REACTOR STRUCTURE

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

Mahlmeister, J E; Haberer, W V; Casey, D F

1960-12-15

The Hallam Nuclear Power Facility reactor structure, including the cavity liner, is described, and the design philosophy and special design requirements which were developed during the preliminary and final engineering phases of the project are explained. The structure was designed for 600 deg F inlet and 1000 deg F outlet operating sodium temperatures and fabricated of austenitic and ferritic stainless steels. Support for the reactor core components and adequate containment for biological safeguards were readily provided even though quite conservative design philosophy was used. The calculated operating characteristics, including heat generation, temperature distributions and stress levels for full-power operation, aremore » summarized. Ship fabrication and field installation experiences are also briefly related. Results of this project have established that the sodium graphite reactor permits practical and economical fabrication and field erection procedures; considerably higher operating design temperatures are believed possible without radical design changes. Also, larger reactor structures can be similarly constructed for higher capacity (300 to 1000 Mwe) nuclear power plants. (auth)« less

The energy release and temperature field in the ultracold neutron source of the WWR-M reactor at the Petersburg Nuclear Physics Institute

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

Serebrov, A. P., E-mail: serebrov@pnpi.spb.ru; Kislitsin, B. V.; Onegin, M. S.

2016-12-15

Results of calculations of energy releases and temperature fields in the ultracold neutron source under design at the WWR-M reactor are presented. It is shown that, with the reactor power of 18 MW, the power of energy release in the 40-L volume of the source with superfluid helium will amount to 28.5 W, while 356 W will be released in a liquid-deuterium premoderator. The lead shield between the reactor core and the source reduces the radiative heat release by an order of magnitude. A thermal power of 22 kW is released in it, which is removed by passage of water.more » The distribution of temperatures in all components of the vacuum structure is presented, and the temperature does not exceed 100°C at full reactor power. The calculations performed make it possible to go to design of the source.« less

The role of nuclear reactors in space exploration and development

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

Lipinski, R.J.

2000-07-01

The United States has launched more than 20 radioisotopic thermoelectric generators (RTGs) into space over the past 30 yr but has launched only one nuclear reactor, and that was in 1965. Russia has launched more than 30 reactors. The RTGs use the heat of alpha decay of {sup 238}Pu for power and typically generate <1 kW of electricity. Apollo, Pioneer, Voyager, Viking, Galileo, Ulysses, and Cassini all used RTGs. Space reactors use the fission energy of {sup 235}U; typical designs are for 100 to 1000 kW of electricity. The only US space reactor launch (SNAP-10A) was a demonstration mission. Onemore » reason for the lack of space reactor use by the United States was the lack of space missions that required high power. But, another was the assumed negative publicity that would accompany a reactor launch. The net result is that all space reactor programs after 1970 were terminated before an operating space reactor could be developed, and they are now many years from recovering the ability to build them. Two major near-term needs for space reactors are the human exploration of Mars and advanced missions to and beyond the orbit of Jupiter. To help obtain public acceptance of space reactors, one must correct some of the misconceptions concerning space reactors and convey the following facts to the public and to decision makers: Space reactors are 1000 times smaller in power and size than a commercial power reactor. A space reactor at launch is only as radioactive as a pile of dirt 60 m (200 ft) across. A space reactor contains no plutonium at launch. It does not become significantly radioactive until it is turned on, and it will be engineered so that no launch accident can turn it on, even if that means fueling it after launch. The reactor will not be turned on until it is in a high stable orbit or even on an earth-escape trajectory for some missions. The benefits of space reactors are that they give humanity a stairway to the planets and perhaps the stars. They open a new frontier for their children and their grandchildren. They pave the way for all life on earth to move out into the solar system. At one time, humans built and flew space reactors; it is time to do so again.« less

Nuclear space power safety and facility guidelines study

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

Mehlman, W.F.

1995-09-11

This report addresses safety guidelines for space nuclear reactor power missions and was prepared by The Johns Hopkins University Applied Physics Laboratory (JHU/APL) under a Department of Energy grant, DE-FG01-94NE32180 dated 27 September 1994. This grant was based on a proposal submitted by the JHU/APL in response to an {open_quotes}Invitation for Proposals Designed to Support Federal Agencies and Commercial Interests in Meeting Special Power and Propulsion Needs for Future Space Missions{close_quotes}. The United States has not launched a nuclear reactor since SNAP 10A in April 1965 although many Radioisotope Thermoelectric Generators (RTGs) have been launched. An RTG powered system ismore » planned for launch as part of the Cassini mission to Saturn in 1997. Recently the Ballistic Missile Defense Office (BMDO) sponsored the Nuclear Electric Propulsion Space Test Program (NEPSTP) which was to demonstrate and evaluate the Russian-built TOPAZ II nuclear reactor as a power source in space. As of late 1993 the flight portion of this program was canceled but work to investigate the attributes of the reactor were continued but at a reduced level. While the future of space nuclear power systems is uncertain there are potential space missions which would require space nuclear power systems. The differences between space nuclear power systems and RTG devices are sufficient that safety and facility requirements warrant a review in the context of the unique features of a space nuclear reactor power system.« less

Design and performance of a trickle-bed bioreactor with immobilized hybridoma cells.

PubMed

Phillips, H A; Scharer, J M; Bols, N C; Moo-Young, M

1992-01-01

A trickle-bed system employing inert matrices of vermiculite or polyurethane foam packed in the downcomer section of a split-flow air-lift reactor has been developed for hybridoma culture to enhance antibody productivity. This quiescent condition favoured occlusion and allowed the cells to achieve densities twelve fold greater (12.8 x 10(6) cells/ml reactor for polyurethane foam) than in free cell suspension. The reactor was operated in a cyclic batch mode whereby defined volumes of medium were periodically withdrawn and replaced with equal volumes of fresh medium. The pH of the medium was used as the indicator of the feeding schedule. Glucose, lactate and ammonia concentrations reached a stationary value after 5 days. With vermiculite packing, a monoclonal antibody (MAb) concentration of 2.4 mg/l was achieved after 12 days. The MAb concentration declined then increased to a value of 1.8 mg/l. In the polyurethane foam average monoclonal antibody (MAb) concentrations reached a stationary value of 1.1 mg/l in the first 20 days and increased to a new stationary state value of 2.1 mg/l for the remainder of the production. MAb productivity in the trickle-bed reactor was 0.3 mg/l.d (polyurethane foam) and 0.18 mg/l.d (vermiculite) in comparison to 0.12 mg/l.d for free cell suspension. This trickle-bed system seems to be an attractive way of increasing MAb productivity in culture.

Growth of plant root cultures in liquid- and gas-dispersed reactor environments.

PubMed

McKelvey, S A; Gehrig, J A; Hollar, K A; Curtis, W R

1993-01-01

The growth of Agrobacterium transformed "hairy root" cultures of Hyoscyamus muticus was examined in various liquid- and gas-dispersed bioreactor configurations. Reactor runs were replicated to provide statistical comparisons of nutrient availability on culture performance. Accumulated tissue mass in submerged air-sparged reactors was 31% of gyratory shake-flask controls. Experiments demonstrate that poor performance of sparged reactors is not due to bubble shear damage, carbon dioxide stripping, settling, or flotation of roots. Impaired oxygen transfer due to channeling and stagnation of the liquid phase are the apparent causes of poor growth. Roots grown on a medium-perfused inclined plane grew at 48% of gyratory controls. This demonstrates the ability of cultures to partially compensate for poor liquid distribution through vascular transport of nutrients. A reactor configuration in which the medium is sprayed over the roots and permitted to drain down through the root tissue was able to provide growth rates which are statistically indistinguishable (95% T-test) from gyratory shake-flask controls. In this type of spray/trickle-bed configuration, it is shown that distribution of the roots becomes a key factor in controlling the rate of growth. Implications of these results regarding design and scale-up of bioreactors to produce fine chemicals from root cultures are discussed.

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

The Angra Neutrino Project: precise measurement of {theta}{sub 13} and safeguards applications of neutrino detectors

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

Casimiro, E.; Anjos, J. C.

2009-04-20

We present an introduction to the Angra Neutrino Project. The goal of the project is to explore the use of neutrino detectors to monitor the reactor activity. The Angra Project, willl employ as neutrino sources the reactors of the nuclear power complex in Brazil, located in Angra dos Reis, some 150 Km south from the city of Rio de Janeiro. The Angra collaboration will develop and operate a low-mass neutrino detector to monitor the nuclear reactor activity, in particular to measure the reactor thermal power and the reactor fuel isotopic composition.

The Angra Neutrino Project: precise measurement of θ13 and safeguards applications of neutrino detectors

NASA Astrophysics Data System (ADS)

Casimiro, E.; Anjos, J. C.

2009-04-01

We present an introduction to the Angra Neutrino Project. The goal of the project is to explore the use of neutrino detectors to monitor the reactor activity. The Angra Project, willl employ as neutrino sources the reactors of the nuclear power complex in Brazil, located in Angra dos Reis, some 150 Km south from the city of Rio de Janeiro. The Angra collaboration will develop and operate a low-mass neutrino detector to monitor the nuclear reactor activity, in particular to measure the reactor thermal power and the reactor fuel isotopic composition.

Beyond ITER: neutral beams for a demonstration fusion reactor (DEMO) (invited).

PubMed

McAdams, R

2014-02-01

In the development of magnetically confined fusion as an economically sustainable power source, International Tokamak Experimental Reactor (ITER) is currently under construction. Beyond ITER is the demonstration fusion reactor (DEMO) programme in which the physics and engineering aspects of a future fusion power plant will be demonstrated. DEMO will produce net electrical power. The DEMO programme will be outlined and the role of neutral beams for heating and current drive will be described. In particular, the importance of the efficiency of neutral beam systems in terms of injected neutral beam power compared to wallplug power will be discussed. Options for improving this efficiency including advanced neutralisers and energy recovery are discussed.

Target-fueled nuclear reactor for medical isotope production

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

Coats, Richard L.; Parma, Edward J.

A small, low-enriched, passively safe, low-power nuclear reactor comprises a core of target and fuel pins that can be processed to produce the medical isotope .sup.99Mo and other fission product isotopes. The fuel for the reactor and the targets for the .sup.99Mo production are the same. The fuel can be low enriched uranium oxide, enriched to less than 20% .sup.235U. The reactor power level can be 1 to 2 MW. The reactor is passively safe and maintains negative reactivity coefficients. The total radionuclide inventory in the reactor core is minimized since the fuel/target pins are removed and processed after 7more » to 21 days.« less

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.

Code Development and Assessment for Reactor Outage Thermal-Hydraulic and Safety Analysis - Midloop Operation with Loss of Residual Heat Removal

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

Liang, Thomas K.S.; Ko, F.-K

Although only a few percent of residual power remains during plant outages, the associated risk of core uncovery and corresponding fuel overheating has been identified to be relatively high, particularly under midloop operation (MLO) in pressurized water reactors. However, to analyze the system behavior during outages, the tools currently available, such as RELAP5, RETRAN, etc., cannot easily perform the task. Therefore, a medium-sized program aiming at reactor outage simulation and evaluation, such as MLO with the loss of residual heat removal (RHR), was developed. All important thermal-hydraulic processes involved during MLO with the loss of RHR will be properly simulatedmore » by the newly developed reactor outage simulation and evaluation (ROSE) code. Important processes during MLO with loss of RHR involve a pressurizer insurge caused by the hot-leg flooding, reflux condensation, liquid holdup inside the steam generator, loop-seal clearance, core-level depression, etc. Since the accuracy of the pressure distribution from the classical nodal momentum approach will be degraded when the system is stratified and under atmospheric pressure, the two-region approach with a modified two-fluid model will be the theoretical basis of the new program to analyze the nuclear steam supply system during plant outages. To verify the analytical model in the first step, posttest calculations against the closed integral midloop experiments with loss of RHR were performed. The excellent simulation capacity of the ROSE code against the Institute of Nuclear Energy Research Integral System Test Facility (IIST) test data is demonstrated.« less

10 CFR 50.70 - Inspections.

Code of Federal Regulations, 2013 CFR

2013-01-01

... Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate, provide... New Reactors, or the Director, Office of Nuclear Reactor Regulation. All furniture, supplies and... construction permit holder (nuclear power reactor only) shall ensure that the arrival and presence of an NRC...

10 CFR 50.70 - Inspections.

Code of Federal Regulations, 2011 CFR

2011-01-01

... Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate, provide... New Reactors, or the Director, Office of Nuclear Reactor Regulation. All furniture, supplies and... construction permit holder (nuclear power reactor only) shall ensure that the arrival and presence of an NRC...

10 CFR 50.70 - Inspections.

Code of Federal Regulations, 2012 CFR

2012-01-01

... Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate, provide... New Reactors, or the Director, Office of Nuclear Reactor Regulation. All furniture, supplies and... construction permit holder (nuclear power reactor only) shall ensure that the arrival and presence of an NRC...

10 CFR 50.70 - Inspections.

Code of Federal Regulations, 2010 CFR

2010-01-01

... Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate, provide... New Reactors, or the Director, Office of Nuclear Reactor Regulation. All furniture, supplies and... construction permit holder (nuclear power reactor only) shall ensure that the arrival and presence of an NRC...

10 CFR 50.70 - Inspections.

Code of Federal Regulations, 2014 CFR

2014-01-01

... Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate, provide... New Reactors, or the Director, Office of Nuclear Reactor Regulation. All furniture, supplies and... construction permit holder (nuclear power reactor only) shall ensure that the arrival and presence of an NRC...

Nuclear Security: Action May Be Needed to Reassess the Security of NRC-Licensed Research Reactors. Report to the Ranking Member, Subcommittee on National Security and Foreign Affairs, Committee on Oversight and Government Reform, House of Representatives. GAO-08-403

ERIC Educational Resources Information Center

Aloise, Gene

2008-01-01

There are 37 research reactors in the United States, mostly located on college campuses. Of these, 33 reactors are licensed and regulated by the Nuclear Regulatory Commission (NRC). Four are operated by the Department of Energy (DOE) and are located at three national laboratories. Although less powerful than commercial nuclear power reactors,…

Self-teaching neural network learns difficult reactor control problem

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

Jouse, W.C.

1989-01-01

A self-teaching neural network used as an adaptive controller quickly learns to control an unstable reactor configuration. The network models the behavior of a human operator. It is trained by allowing it to operate the reactivity control impulsively. It is punished whenever either the power or fuel temperature stray outside technical limits. Using a simple paradigm, the network constructs an internal representation of the punishment and of the reactor system. The reactor is constrained to small power orbits.

A review of engineering aspects of intensification of chemical synthesis using ultrasound.

PubMed

Sancheti, Sonam V; Gogate, Parag R

2017-05-01

Cavitation generated using ultrasound can enhance the rates of several chemical reactions giving better selectivity based on the physical and chemical effects. The present review focuses on overview of the different reactions that can be intensified using ultrasound followed by the discussion on the chemical kinetics for ultrasound assisted reactions, engineering aspects related to reactor designs and effect of operating parameters on the degree of intensification obtained for chemical synthesis. The cavitational effects in terms of magnitudes of collapse temperatures and collapse pressure, number of free radicals generated and extent of turbulence are strongly dependent on the operating parameters such as ultrasonic power, frequency, duty cycle, temperature as well as physicochemical parameters of liquid medium which controls the inception of cavitation. Guidelines have been presented for the optimum selection based on the critical analysis of the existing literature so that maximum process intensification benefits can be obtained. Different reactor designs have also been analyzed with guidelines for efficient scale up of the sonochemical reactor, which would be dependent on the type of reaction, controlling mechanism of reaction, catalyst and activation energy requirements. Overall, it has been established that sonochemistry offers considerable potential for green and sustainable processing and efficient scale up procedures are required so as to harness the effects at actual commercial level. Copyright © 2016 Elsevier B.V. All rights reserved.

Numerical analysis of biomass torrefaction reactor with recirculation of heat carrier

NASA Astrophysics Data System (ADS)

Director, L. B.; Ivanin, O. A.; Sinelshchikov, V. A.

2018-01-01

In this paper, results of numerical analysis of the energy-technological complex consisting of the gas piston power plant, the torrefaction reactor with recirculation of gaseous heat carrier and the heat recovery boiler are presented. Calculations of the reactor without recirculation and with recirculation of the heat carrier in torrefaction zone at different frequencies of unloading of torrefied biomass were held. It was shown that in recirculation mode the power of the gas piston power plant, required for providing given reactor productivity, is reduced several times and the consumption of fuel gas, needed for combustion of volatile torrefaction products in the heat recovery boiler, is reduced by an order.

Megawatt Class Nuclear Space Power Systems (MCNSPS) conceptual design and evaluation report. Volume 4: Concepts selection, conceptual designs, recommendations

NASA Technical Reports Server (NTRS)

Wetch, J. R.

1988-01-01

A study was conducted by NASA Lewis Research Center for the Triagency SP-100 program office. The objective was to determine which reactor, conversion and radiator technologies would best fulfill future Megawatt Class Nuclear Space Power System Requirements. The requirement was 10 megawatts for 5 years of full power operation and 10 years system life on orbit. A variety of liquid metal and gas cooled reactors, static and dynamic conversion systems, and passive and dynamic radiators were considered. Four concepts were selected for more detailed study: (1) a gas cooled reactor with closed cycle Brayton turbine-alternator conversion with heatpipe and pumped tube fin rejection, (2) a Lithium cooled reactor with a free piston Stirling engine-linear alternator and a pumped tube-fin radiator,(3) a Lithium cooled reactor with a Potassium Rankine turbine-alternator and heat pipe radiator, and (4) a Lithium cooled incore thermionic static conversion reactor with a heat pipe radiator. The systems recommended for further development to meet a 10 megawatt long life requirement are the Lithium cooled reactor with the K-Rankine conversion and heat pipe radiator, and the Lithium cooled incore thermionic reactor with heat pipe radiator.

The Ongoing Impact of the U.S. Fast Reactor Integral Experiments Program

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

John D. Bess; Michael A. Pope; Harold F. McFarlane

2012-11-01

The creation of a large database of integral fast reactor physics experiments advanced nuclear science and technology in ways that were unachievable by less capital intensive and operationally challenging approaches. They enabled the compilation of integral physics benchmark data, validated (or not) analytical methods, and provided assurance of future rector designs The integral experiments performed at Argonne National Laboratory (ANL) represent decades of research performed to support fast reactor design and our understanding of neutronics behavior and reactor physics measurements. Experiments began in 1955 with the Zero Power Reactor No. 3 (ZPR-3) and terminated with the Zero Power Physics Reactormore » (ZPPR, originally the Zero Power Plutonium Reactor) in 1990 at the former ANL-West site in Idaho, which is now part of the Idaho National Laboratory (INL). Two additional critical assemblies, ZPR-6 and ZPR-9, operated at the ANL-East site in Illinois. A total of 128 fast reactor assemblies were constructed with these facilities [1]. The infrastructure and measurement capabilities are too expensive to be replicated in the modern era, making the integral database invaluable as the world pushes ahead with development of liquid metal cooled reactors.« less

ADVANCED COURSE ON FUEL ELEMENTS FOR WATER COOLED POWER REACTORS, ORGANIZED BY THE NETHERLANDS'-NORWEGIAN REACTOR SCHOOL AT INSTITUTT FOR ATOMENERGI, KJELLER, NORWAY, 22nd AUGUST-3rd SEPTEMBER,1960. VOLUME III

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

Aas, S.; Barendregt, T.J.; Chesne, A.

1960-07-01

A series of lectures on fuel elements for water-cooled power reactors are presented. Topics covered include fabrication, properties, cladding, radiation damage, design, cycling, storage and transpont, and reprocessing. Separate records have been prepared for each section.

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.

Applications of plasma core reactors to terrestrial energy systems

NASA Technical Reports Server (NTRS)

Latham, T. S.; Biancardi, F. R.; Rodgers, R. J.

1974-01-01

Plasma core reactors offer several new options for future energy needs in addition to space power and propulsion applications. Power extraction from plasma core reactors with gaseous nuclear fuel allows operation at temperatures higher than conventional reactors. Highly efficient thermodynamic cycles and applications employing direct coupling of radiant energy are possible. Conceptual configurations of plasma core reactors for terrestrial applications are described. Closed-cycle gas turbines, MHD systems, photo- and thermo-chemical hydrogen production processes, and laser systems using plasma core reactors as prime energy sources are considered. Cycle efficiencies in the range of 50 to 65 percent are calculated for closed-cycle gas turbine and MHD electrical generators. Reactor advantages include continuous fuel reprocessing which limits inventory of radioactive by-products and thorium-U-233 breeder configurations with about 5-year doubling times.-

Fuel burnup analysis for IRIS reactor using MCNPX and WIMS-D5 codes

NASA Astrophysics Data System (ADS)

Amin, E. A.; Bashter, I. I.; Hassan, Nabil M.; Mustafa, S. S.

2017-02-01

International Reactor Innovative and Secure (IRIS) reactor is a compact power reactor designed with especial features. It contains Integral Fuel Burnable Absorber (IFBA). The core is heterogeneous both axially and radially. This work provides the full core burn up analysis for IRIS reactor using MCNPX and WIMDS-D5 codes. Criticality calculations, radial and axial power distributions and nuclear peaking factor at the different stages of burnup were studied. Effective multiplication factor values for the core were estimated by coupling MCNPX code with WIMS-D5 code and compared with SAS2H/KENO-V code values at different stages of burnup. The two calculation codes show good agreement and correlation. The values of radial and axial powers for the full core were also compared with published results given by SAS2H/KENO-V code (at the beginning and end of reactor operation). The behavior of both radial and axial power distribution is quiet similar to the other data published by SAS2H/KENO-V code. The peaking factor values estimated in the present work are close to its values calculated by SAS2H/KENO-V code.

HISTORICAL AMERICAN ENGINEERING RECORD - IDAHO NATIONAL ENGINEERING AND ENVIRONMENTAL LABORATORY, TEST AREA NORTH, HAER NO. ID-33-E

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

Susan Stacy; Hollie K. Gilbert

2005-02-01

Test Area North (TAN) was a site of the Aircraft Nuclear Propulsion (ANP) Project of the U.S. Air Force and the Atomic Energy Commission. Its Cold War mission was to develop a turbojet bomber propelled by nuclear power. The project was part of an arms race. Test activities took place in five areas at TAN. The Assembly & Maintenance area was a shop and hot cell complex. Nuclear tests ran at the Initial Engine Test area. Low-power test reactors operated at a third cluster. The fourth area was for Administration. A Flight Engine Test facility (hangar) was built to housemore » the anticipated nuclear-powered aircraft. Experiments between 1955-1961 proved that a nuclear reactor could power a jet engine, but President John F. Kennedy canceled the project in March 1961. ANP facilities were adapted for new reactor projects, the most important of which were Loss of Fluid Tests (LOFT), part of an international safety program for commercial power reactors. Other projects included NASA's Systems for Nuclear Auxiliary Power and storage of Three Mile Island meltdown debris. National missions for TAN in reactor research and safety research have expired; demolition of historic TAN buildings is underway.« less

Converting Maturing Nuclear Sites to Integrated Power Production Islands

DOE PAGES

Solbrig, Charles W.

2011-01-01

Nuclear islands, which are integrated power production sites, could effectively sequester and safeguard the US stockpile of plutonium. A nuclear island, an evolution of the integral fast reactor, utilizes all the Transuranics (Pu plus minor actinides) produced in power production, and it eliminates all spent fuel shipments to and from the site. This latter attribute requires that fuel reprocessing occur on each site and that fast reactors be built on-site to utilize the TRU. All commercial spent fuel shipments could be eliminated by converting all LWR nuclear power sites to nuclear islands. Existing LWR sites have the added advantage ofmore » already possessing a license to produce nuclear power. Each could contribute to an increase in the nuclear power production by adding one or more fast reactors. Both the TRU and the depleted uranium obtained in reprocessing would be used on-site for fast fuel manufacture. Only fission products would be shipped to a repository for storage. The nuclear island concept could be used to alleviate the strain of LWR plant sites currently approaching or exceeding their spent fuel pool storage capacity. Fast reactor breeding ratio could be designed to convert existing sites to all fast reactors, or keep the majority thermal.« 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...

Boiling water neutronic reactor incorporating a process inherent safety design

DOEpatents

Forsberg, C.W.

1985-02-19

A boiling-water reactor core is positioned within a prestressed concrete reactor vessel of a size which will hold a supply of coolant water sufficient to submerge and cool the reactor core by boiling for a period of at least one week after shutdown. Separate volumes of hot, clean (nonborated) water for cooling during normal operation and cool highly borated water for emergency cooling and reactor shutdown are separated by an insulated wall during normal reactor operation with contact between the two water volumes being maintained at interfaces near the top and bottom ends of the reactor vessel. Means are provided for balancing the pressure of the two water volumes at the lower interface zone during normal operation to prevent entry of the cool borated water into the reactor core region, for detecting the onset of excessive power to coolant flow conditions in the reactor core and for detecting low water levels of reactor coolant. Cool borated water is permitted to flow into the reactor core when low reactor coolant levels or excessive power to coolant flow conditions are encountered.

Boiling water neutronic reactor incorporating a process inherent safety design

DOEpatents

Forsberg, Charles W.

1987-01-01

A boiling-water reactor core is positioned within a prestressed concrete reactor vessel of a size which will hold a supply of coolant water sufficient to submerge and cool the reactor core by boiling for a period of at least one week after shutdown. Separate volumes of hot, clean (non-borated) water for cooling during normal operation and cool highly borated water for emergency cooling and reactor shutdown are separated by an insulated wall during normal reactor operation with contact between the two water volumes being maintained at interfaces near the top and bottom ends of the reactor vessel. Means are provided for balancing the pressure of the two volumes at the lower interface zone during normal operation to prevent entry of the cool borated water into the reactor core region, for detecting the onset of excessive power to coolant flow conditions in the reactor core and for detecting low water levels of reactor coolant. Cool borated water is permitted to flow into the reactor core when low reactor coolant levels or excessive power to coolant flow conditions are encountered.

The startup of the Dodewaard natural circulation boiling water reactor -- Experiences

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

Nissen, W.H.M.; Van Der Voet, J.; Karuza, J.

1994-07-01

Because of its similarity to the simplified boiling water reactor (SBWR), the Dodewaard natural circulation boiling water reactor (BWR) is of special interest to further development of the SBWR design. It has become especially important to gain more insight into the Dodewaard BWR behavior during startup, paying special attention to its stability. Therefore, special instrumentation was used by means of which a series of measurements were taken during the two startups in February and June 1992. The results obtained from these measurements are used to deepen insight into the recirculation flow and the stability of the reactor during startup undermore » conditions with a normal pressure/power trajectory. They have already shown a very early recirculation flow onset during low-power operation and no indication of reactor instability. Furthermore, they will be used as a basis for the research program investigating the reactor behavior under different pressure/power conditions, which is scheduled for next year.« less

Applications of power beaming from space-based nuclear power stations. [Laser beaming to airplanes; microwave beaming to ground

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

Powell, J.R.; Botts, T.E.; Hertzberg, A.

1981-01-01

Power beaming from space-based reactor systems is examined using an advanced compact, lightweight Rotating Bed Reactor (RBR). Closed Brayton power conversion efficiencies in the range of 30 to 40% can be achieved with turbines, with reactor exit temperatures on the order of 2000/sup 0/K and a liquid drop radiator to reject heat at temperatures of approx. 500/sup 0/K. Higher RBR coolant temperatures (up to approx. 3000/sup 0/K) are possible, but gains in power conversion efficiency are minimal, due to lower expander efficiency (e.g., a MHD generator). Two power beaming applications are examined - laser beaming to airplanes and microwave beamingmore » to fixed ground receivers. Use of the RBR greatly reduces system weight and cost, as compared to solar power sources. Payback times are a few years at present prices for power and airplane fuel.« less

Proposed Guidance for Preparing and Reviewing Molten Salt Nonpower Reactor Licence Applications (NUREG-1537)

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

Belles, Randy; Flanagan, George F.; Voth, Marcus

Development of non-power molten salt reactor (MSR) test facilities is under consideration to support the analyses needed for development of a full-scale MSR. These non-power MSR test facilities will require review by the US Nuclear Regulatory Commission (NRC) staff. This report proposes chapter adaptations for NUREG-1537 in the form of interim staff guidance to address preparation and review of molten salt non-power reactor license applications. The proposed adaptations are based on a previous regulatory gap analysis of select chapters from NUREG-1537 for their applicability to non-power MSRs operating with a homogeneous fuel salt mixture.

Gas core reactors for actinide transmutation and breeder applications

NASA Technical Reports Server (NTRS)

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

1978-01-01

This work consists of design power plant studies for four types of reactor systems: uranium plasma core breeder, uranium plasma core actinide transmuter, UF6 breeder and UF6 actinide transmuter. The plasma core systems can be coupled to MHD generators to obtain high efficiency electrical power generation. A 1074 MWt UF6 breeder reactor was designed with a breeding ratio of 1.002 to guard against diversion of fuel. Using molten salt technology and a superheated steam cycle, an efficiency of 39.2% was obtained for the plant and the U233 inventory in the core and heat exchangers was limited to 105 Kg. It was found that the UF6 reactor can produce high fluxes (10 to the 14th power n/sq cm-sec) necessary for efficient burnup of actinide. However, the buildup of fissile isotopes posed severe heat transfer problems. Therefore, the flux in the actinide region must be decreased with time. Consequently, only beginning-of-life conditions were considered for the power plant design. A 577 MWt UF6 actinide transmutation reactor power plant was designed to operate with 39.3% efficiency and 102 Kg of U233 in the core and heat exchanger for beginning-of-life conditions.

Gaseous-fuel nuclear reactor research for multimegawatt power in space

NASA Technical Reports Server (NTRS)

Thom, K.; Schneider, R. T.; Helmick, H. H.

1977-01-01

In the gaseous-fuel reactor concept, the fissile material is contained in a moderator-reflector cavity and exists in the form of a flowing gas or plasma separated from the cavity walls by means of fluid mechanical forces. Temperatures in excess of structural limitations are possible for low-specific-mass power and high-specific-impulse propulsion in space. Experiments have been conducted with a canister filled with enriched UF6 inserted into a beryllium-reflected cavity. A theoretically predicted critical mass of 6 kg was measured. The UF6 was also circulated through this cavity, demonstrating stable reactor operation with the fuel in motion. Because the flowing gaseous fuel can be continuously processed, the radioactive waste in this type of reactor can be kept small. Another potential of fissioning gases is the possibility of converting the kinetic energy of fission fragments directly into coherent electromagnetic radiation, the nuclear pumping of lasers. Numerous nuclear laser experiments indicate the possibility of transmitting power in space directly from fission energy. The estimated specific mass of a multimegawatt gaseous-fuel reactor power system is from 1 to 5 kg/kW while the companion laser-power receiver station would be much lower in specific mass.

A PC-based high temperature gas reactor simulator for Indonesian conceptual HTR reactor basic training

NASA Astrophysics Data System (ADS)

Syarip; Po, L. C. C.

2018-05-01

In planning for nuclear power plant construction in Indonesia, helium cooled high temperature reactor (HTR) is favorable for not relying upon water supply that might be interrupted by earthquake. In order to train its personnel, BATAN has cooperated with Micro-Simulation Technology of USA to develop a 200 MWt PC-based simulation model PCTRAN/HTR. It operates in Win10 environment with graphic user interface (GUI). Normal operation of startup, power maneuvering, shutdown and accidents including pipe breaks and complete loss of AC power have been conducted. A sample case of safety analysis simulation to demonstrate the inherent safety features of HTR was done for helium pipe break malfunction scenario. The analysis was done for the variation of primary coolant pipe break i.e. from 0,1% - 0,5 % and 1% - 10 % helium gas leakages, while the reactor was operated at the maximum constant power of 10 MWt. The result shows that the highest temperature of HTR fuel centerline and coolant were 1150 °C and 1296 °C respectively. With 10 kg/s of helium flow in the reactor core, the thermal power will back to the startup position after 1287 s of helium pipe break malfunction.

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.

High Temperature Fusion Reactor Cooling Using Brayton Cycle Based Partial Energy Conversion

NASA Technical Reports Server (NTRS)

Juhasz, Albert J.; Sawicki, Jerzy T.

2003-01-01

For some future space power systems using high temperature nuclear heat sources most of the output energy will be used in other than electrical form, and only a fraction of the total thermal energy generated will need to be converted to electrical work. The paper describes the conceptual design of such a partial energy conversion system, consisting of a high temperature fusion reactor operating in series with a high temperature radiator and in parallel with dual closed cycle gas turbine (CCGT) power systems, also referred to as closed Brayton cycle (CBC) systems, which are supplied with a fraction of the reactor thermal energy for conversion to electric power. Most of the fusion reactor's output is in the form of charged plasma which is expanded through a magnetic nozzle of the interplanetary propulsion system. Reactor heat energy is ducted to the high temperature series radiator utilizing the electric power generated to drive a helium gas circulation fan. In addition to discussing the thermodynamic aspects of the system design the authors include a brief overview of the gas turbine and fan rotor-dynamics and proposed bearing support technology along with performance characteristics of the three phase AC electric power generator and fan drive motor.

Thermal characteristics analysis of microwaves reactor for pyrolysis of used cooking oil

NASA Astrophysics Data System (ADS)

Anis, Samsudin; Shahadati, Laily; Sumbodo, Wirawan; Wahyudi

2017-03-01

The research is objected to develop microwave reactor for pyrolysis of used cooking oil. The effect of microwave power as well as addition of char as absorber towards its thermal characteristic were investigated. Domestic microwave was modified and used to test the thermal characteristic of used cooking oil in the terms of temperature evolution, heating rate, and thermal efficiency. The samples were examined under various microwave power of 347W, 399W, 572W and 642W for 25 minutes of irradiation time. The char loading was tested in the level of 0, 50, and 100 g. Microwave reactor consists of microwave unit with a maximum power of 642W, a ceramic reactor, and a condenser equipped with temperature measurement system was successfully developed. It was found that microwave power and addition of absorber significantly influenced the thermal characteristic of microwave reactor. Under investigated condition, the optimum result was obtained at microwave power of 642W and 100 g of char. The condition was able to provide temperature of 480°C, heating rate of 18.2°C/min and thermal efficiency of 53% that is suitable to pyrolyze used cooking oil.

High Temperature Fusion Reactor Cooling Using Brayton Cycle Based Partial Energy Conversion

NASA Astrophysics Data System (ADS)

Juhasz, Albert J.; Sawicki, Jerzy T.

2004-02-01

For some future space power systems using high temperature nuclear heat sources most of the output energy will be used in other than electrical form, and only a fraction of the total thermal energy generated will need to be converted to electrical work. The paper describes the conceptual design of such a ``partial energy conversion'' system, consisting of a high temperature fusion reactor operating in series with a high temperature radiator and in parallel with dual closed cycle gas turbine (CCGT) power systems, also referred to as closed Brayton cycle (CBC) systems, which are supplied with a fraction of the reactor thermal energy for conversion to electric power. Most of the fusion reactor's output is in the form of charged plasma which is expanded through a magnetic nozzle of the interplanetary propulsion system. Reactor heat energy is ducted to the high temperature series radiator utilizing the electric power generated to drive a helium gas circulation fan. In addition to discussing the thermodynamic aspects of the system design the authors include a brief overview of the gas turbine and fan rotor-dynamics and proposed bearing support technology along with performance characteristics of the three phase AC electric power generator and fan drive motor.

Common cause evaluations in applied risk analysis of nuclear power plants. [PWR

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

Taniguchi, T.; Ligon, D.; Stamatelatos, M.

1983-04-01

Qualitative and quantitative approaches were developed for the evaluation of common cause failures (CCFs) in nuclear power plants and were applied to the analysis of the auxiliary feedwater systems of several pressurized water reactors (PWRs). Key CCF variables were identified through a survey of experts in the field and a review of failure experience in operating PWRs. These variables were classified into categories of high, medium, and low defense against a CCF. Based on the results, a checklist was developed for analyzing CCFs of systems. Several known techniques for quantifying CCFs were also reviewed. The information provided valuable insights inmore » the development of a new model for estimating CCF probabilities, which is an extension of and improvement over the Beta Factor method. As applied to the analysis of the PWR auxiliary feedwater systems, the method yielded much more realistic values than the original Beta Factor method for a one-out-of-three system.« less

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.

Preliminary Comparison of Radioactive Waste Disposal Cost for Fusion and Fission Reactors

NASA Astrophysics Data System (ADS)

Seki, Yasushi; Aoki, Isao; Yamano, Naoki; Tabara, Takashi

1997-09-01

The environmental and economic impact of radioactive waste (radwaste) generated from fusion power reactors using five types of structural materials and a fission reactor has been evaluated and compared. Possible radwaste disposal scenario of fusion radwaste in Japan is considered. The exposure doses were evaluated for the skyshine of gamma-ray during the disposal operation, groundwater migration scenario during the institutional control period of 300 years and future site use scenario after the institutional period. The radwaste generated from a typical light water fission reactor was evaluated using the same methodology as for the fusion reactors. It is found that radwaste from the fusion reactors using F82H and SiC/SiC composites without impurities could be disposed by the shallow land disposal presently applied to the low level waste in Japan. The disposal cost of radwaste from five fusion power reactors and a typical light water reactor were roughly evaluated and compared.

Planetary surface reactor shielding using indigenous materials

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

Houts, Michael G.; Poston, David I.; Trellue, Holly R.

The exploration and development of Mars will require abundant surface power. Nuclear reactors are a low-cost, low-mass means of providing that power. A significant fraction of the nuclear power system mass is radiation shielding necessary for protecting humans and/or equipment from radiation emitted by the reactor. For planetary surface missions, it may be desirable to provide some or all of the required shielding from indigenous materials. This paper examines shielding options that utilize either purely indigenous materials or a combination of indigenous and nonindigenous materials. {copyright} {ital 1999 American Institute of Physics.}

FALCON nuclear-reactor-pumped laser program and wireless power transmission

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

Lipinski, R.J.; Pickard, P.S.

1992-12-31

FALCON is a high-power, reactor-pumped laser concept. The major strengths of a reactor-pumped laser are (1) simple, modular construction, (2) long-duration, closed-cycle capability, (3) self-contained power, (4) compact size, and (5) a variety of wavelengths (from visible to infrared). Reactor-pumped lasing has been demonstrated experimentally in various mixtures of xenon, argon, neon, and helium at wavelengths of 585, 703, 725, 1271, 1733, 1792, 2032, 2630, 2650, and 3370 nm with intrinsic efficiency as high as 2.5%. Powers up to 300 W for 2 ms have been demonstrated. Projected beam quality for FALCON is good enough that frequency doubling at reasonablemore » efficiencies could be expected to yield wavelengths at 353, 363, 636, 867, 896, 1016, 1315, 1325, and 1685 nm. Appropriate missions for FALCON are described and include power beaming to satellites, the moon, and unmanned surveillance planes; lunar mapping; space debris removal; and laser propulsion.« less

FALCON nuclear-reactor-pumped laser program and wireless power transmission

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

Lipinski, R.J.; Pickard, P.S.

1992-01-01

FALCON is a high-power, reactor-pumped laser concept. The major strengths of a reactor-pumped laser are (1) simple, modular construction, (2) long-duration, closed-cycle capability, (3) self-contained power, (4) compact size, and (5) a variety of wavelengths (from visible to infrared). Reactor-pumped lasing has been demonstrated experimentally in various mixtures of xenon, argon, neon, and helium at wavelengths of 585, 703, 725, 1271, 1733, 1792, 2032, 2630, 2650, and 3370 nm with intrinsic efficiency as high as 2.5%. Powers up to 300 W for 2 ms have been demonstrated. Projected beam quality for FALCON is good enough that frequency doubling at reasonablemore » efficiencies could be expected to yield wavelengths at 353, 363, 636, 867, 896, 1016, 1315, 1325, and 1685 nm. Appropriate missions for FALCON are described and include power beaming to satellites, the moon, and unmanned surveillance planes; lunar mapping; space debris removal; and laser propulsion.« less

Advanced Thermal Simulator Testing: Thermal Analysis and Test Results

NASA Technical Reports Server (NTRS)

Bragg-Sitton, Shannon M.; Dickens, Ricky; Dixon, David; Reid, Robert; Adams, Mike; Davis, Joe

2008-01-01

Work at the NASA Marshall Space Flight Center seeks to develop high fidelity, electrically heated thermal simulators that represent fuel elements in a nuclear reactor design to support non-nuclear testing applicable to the development of a space nuclear power or propulsion system. Comparison between the fuel pins and thermal simulators is made at the outer fuel clad surface, which corresponds to the outer sheath surface in the thermal simulator. The thermal simulators that are currently being tested correspond to a SNAP derivative reactor design that could be applied for Lunar surface power. These simulators are designed to meet the geometric and power requirements of a proposed surface power reactor design, accommodate testing of various axial power profiles, and incorporate imbedded instrumentation. This paper reports the results of thermal simulator analysis and testing in a bare element configuration, which does not incorporate active heat removal, and testing in a water-cooled calorimeter designed to mimic the heat removal that would be experienced in a reactor core.

Cavity temperature and flow characteristics in a gas-core test reactor

NASA Technical Reports Server (NTRS)

Putre, H. A.

1973-01-01

A test reactor concept for conducting basic studies on a fissioning uranium plasma and for testing various gas-core reactor concepts is analyzed. The test reactor consists of a conventional fuel-element region surrounding a 61-cm-(2-ft-) diameter cavity region which contains the plasma experiment. The fuel elements provide the neutron flux for the cavity region. The design operating conditions include 60-MW reactor power, 2.7-MW cavity power, 200-atm cavity pressure, and an average uranium plasma temperature of 15,000 K. The analytical results are given for cavity radiant heat transfer, hydrogen transpiration cooling, and uranium wire or powder injection.

Perspectives of The Interagency Nuclear Safety Review Panel (INSRP) on future nuclear powered space missions

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

Gray, L.B.; Pyatt, D.W.; Sholtis, J.A.

1993-01-10

The Interagency Nuclear Safety Review Panel (INSRP) has provided reviews of all nuclear powered spacecraft launched by the United States. The two most recent launches were Ulysses in 1990 and Galileo in 1989. One reactor was launched in 1965 (SNAP-10A). All other U.S. space missions have utilized radioisotopic thermoelectric generators (RTGs). There are several missions in the next few years that are to be nuclear powered, including one that would utilize the Topaz II reactor purchased from Russia. INSRP must realign itself to perform parallel safety assessments of a reactor powered space mission, which has not been done in aboutmore » thirty years, and RTG powered missions.« less

77 FR 9707 - Advisory Committee on Reactor Safeguards Meeting of the ACRS Subcommittee on Power Uprates...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-02-17

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards Meeting of the ACRS Subcommittee on Power Uprates; Revision to February 23, 2012, ACRS Meeting Federal Register Notice The Federal Register Notice for the ACRS Subcommittee meeting on Power Uprates, scheduled to be held on February 23...

CIRFT Data Update and Data Analyses for Spent Nuclear Fuel Vibration Reliability Study

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

Wang, Jy-An John; Wang, Hong

The objective of this research is to collect experimental data on spent nuclear fuel (SNF) from pressurized water reactors (PWRs), including the H. B. Robinson Nuclear Power Station (HBR), Catawba Nuclear Station, North Anna Nuclear Power Station (NA), and the Limerick Nuclear Power Station (LMK) boiling water reactor (BWR).

A Spherical Torus Nuclear Fusion Reactor Space Propulsion Vehicle Concept for Fast Interplanetary Travel

NASA Technical Reports Server (NTRS)

Williams, Craig H.; Borowski, Stanley K.; Dudzinski, Leonard A.; Juhasz, Albert J.

1998-01-01

A conceptual vehicle design enabling fast outer solar system travel was produced predicated on a small aspect ratio spherical torus nuclear fusion reactor. Initial requirements were for a human mission to Saturn with a greater than 5% payload mass fraction and a one way trip time of less than one year. Analysis revealed that the vehicle could deliver a 108 mt crew habitat payload to Saturn rendezvous in 235 days, with an initial mass in low Earth orbit of 2,941 mt. Engineering conceptual design, analysis, and assessment was performed on all ma or systems including payload, central truss, nuclear reactor (including divertor and fuel injector), power conversion (including turbine, compressor, alternator, radiator, recuperator, and conditioning), magnetic nozzle, neutral beam injector, tankage, start/re-start reactor and battery, refrigeration, communications, reaction control, and in-space operations. Detailed assessment was done on reactor operations, including plasma characteristics, power balance, power utilization, and component design.

The diversity and unit of reactor noise theory

NASA Astrophysics Data System (ADS)

Kuang, Zhifeng

The study of reactor noise theory concerns questions about cause and effect relationships, and utilisation of random noise in nuclear reactor systems. The diversity of reactor noise theory arises from the variety of noise sources, the various mathematical treatments applied and various practical purposes. The neutron noise in zero- energy systems arises from the fluctuations in the number of neutrons per fission, the time between nuclear events, and the type of reactions. It can be used to evaluate system parameters. The mathematical treatment is based on the master equation of stochastic branching processes. The noise in power reactor systems is given rise by random processes of technological origin such as vibration of mechanical parts, boiling of the coolant, fluctuations of temperature and pressure. It can be used to monitor reactor behaviour with the possibility of detecting malfunctions at an early stage. The mathematical treatment is based on the Langevin equation. The unity of reactor noise theory arises from the fact that useful information from noise is embedded in the second moments of random variables, which lends the possibility of building up a unified mathematical description and analysis of the various reactor noise sources. Exploring such possibilities is the main subject among the three major topics reported in this thesis. The first subject is within the zero power noise in steady media, and we reported on the extension of the existing theory to more general cases. In Paper I, by use of the master equation approach, we have derived the most general Feynman- and Rossi-alpha formulae so far by taking the full joint statistics of the prompt and all the six groups of delayed neutron precursors, and a multiple emission source into account. The involved problems are solved with a combination of effective analytical techniques and symbolic algebra codes (Mathematica). Paper II gives a numerical evaluation of these formulae. An assessment of the contribution of the terms that are novel as compared to the traditional formulae has been made. The second subject treats a problem in power reactor noise with the Langevin formalism. With a very few exceptions, in all previous work the diffusion approximation was used. In order to extend the treatment to transport theory, in Paper III, we introduced a novel method, i.e. Padé approximation via Lanczos algorithm to calculate the transfer function of a finite slab reactor described by one-group transport equation. It was found that the local-global decomposition of the neutron noise, formerly only reproduced in at least 2- group theory, can be reconstructed. We have also showed the existence of a boundary layer of the neutron noise close to the boundary. Finally, we have explored the possibility of building up a unified theory to account for the coexistence of zero power and power reactor noise in a system. In Paper IV, a unified description of the neutron noise is given by the use of backward master equations in a model where the cross section fluctuations are given as a simple binary pseudorandom process. The general solution contains both the zero power and power reactor noise concurrently, and they can be extracted individually as limiting cases of the general solution. It justified the separate treatments of zero power and power reactor noise. The result was extended to the case including one group of delayed neutron precursors in Paper V.

A model for the release, dispersion and environmental impact of a postulated reactor accident from a submerged commercial nuclear power plant

NASA Astrophysics Data System (ADS)

Bertch, Timothy Creston

1998-12-01

Nuclear power plants are inherently suitable for submerged applications and could provide power to the shore power grid or support future underwater applications. The technology exists today and the construction of a submerged commercial nuclear power plant may become desirable. A submerged reactor is safer to humans because the infinite supply of water for heat removal, particulate retention in the water column, sedimentation to the ocean floor and inherent shielding of the aquatic environment would significantly mitigate the effects of a reactor accident. A better understanding of reactor operation in this new environment is required to quantify the radioecological impact and to determine the suitability of this concept. The impact of release to the environment from a severe reactor accident is a new aspect of the field of marine radioecology. Current efforts have been centered on radioecological impacts of nuclear waste disposal, nuclear weapons testing fallout and shore nuclear plant discharges. This dissertation examines the environmental impact of a severe reactor accident in a submerged commercial nuclear power plant, modeling a postulated site on the Atlantic continental shelf adjacent to the United States. This effort models the effects of geography, decay, particle transport/dispersion, bioaccumulation and elimination with associated dose commitment. The use of a source term equivalent to the release from Chernobyl allows comparison between the impacts of that accident and the postulated submerged commercial reactor plant accident. All input parameters are evaluated using sensitivity analysis. The effect of the release on marine biota is determined. Study of the pathways to humans from gaseous radionuclides, consumption of contaminated marine biota and direct exposure as contaminated water reaches the shoreline is conducted. The model developed by this effort predicts a significant mitigation of the radioecological impact of the reactor accident release with a submerged commercial nuclear power plant. The two box models predict the most of the radio-ecological impact occurs during the first eight days after release. The most significant risk to humans is from consumption of biota. The reduction in impact to humans from a large radioactive release makes the concept worthy of further study.

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. .

10 CFR 50.30 - Filing of application; oath or affirmation.

Code of Federal Regulations, 2012 CFR

2012-01-01

... Reactor Regulation, Director, Office of New Reactors, or Director, Office of Nuclear Material Safety and... Director, Office of New Reactors, or the Director, Office of Nuclear Reactor Regulation, or the Director..., operating license, early site permit, combined license, or manufacturing license for a nuclear power reactor...

10 CFR 50.30 - Filing of application; oath or affirmation.

Code of Federal Regulations, 2011 CFR

2011-01-01

... Reactor Regulation, Director, Office of New Reactors, or Director, Office of Nuclear Material Safety and... Director, Office of New Reactors, or the Director, Office of Nuclear Reactor Regulation, or the Director..., operating license, early site permit, combined license, or manufacturing license for a nuclear power reactor...

10 CFR 50.30 - Filing of application; oath or affirmation.

Code of Federal Regulations, 2014 CFR

2014-01-01

... Reactor Regulation, Director, Office of New Reactors, or Director, Office of Nuclear Material Safety and... Director, Office of New Reactors, or the Director, Office of Nuclear Reactor Regulation, or the Director..., operating license, early site permit, combined license, or manufacturing license for a nuclear power reactor...

10 CFR 50.30 - Filing of application; oath or affirmation.

Code of Federal Regulations, 2013 CFR

2013-01-01

... Reactor Regulation, Director, Office of New Reactors, or Director, Office of Nuclear Material Safety and... Director, Office of New Reactors, or the Director, Office of Nuclear Reactor Regulation, or the Director..., operating license, early site permit, combined license, or manufacturing license for a nuclear power reactor...

Manned space flight nuclear system safety. Volume 3: Reactor system preliminary nuclear safety analysis. Part 1: Reference Design Document (RDD)

NASA Technical Reports Server (NTRS)

1972-01-01

The Reference Design Document, of the Preliminary Safety Analysis Report (PSAR) - Reactor System provides the basic design and operations data used in the nuclear safety analysis of the Rector Power Module as applied to a Space Base program. A description of the power module systems, facilities, launch vehicle and mission operations, as defined in NASA Phase A Space Base studies is included. Each of two Zirconium Hydride Reactor Brayton power modules provides 50 kWe for the nominal 50 man Space Base. The INT-21 is the prime launch vehicle. Resupply to the 500 km orbit over the ten year mission is provided by the Space Shuttle. At the end of the power module lifetime (nominally five years), a reactor disposal system is deployed for boost into a 990 km high altitude (long decay time) earth orbit.

IECEC '83; Proceedings of the Eighteenth Intersociety Energy Conversion Engineering Conference, Orlando, FL, August 21-26, 1983. Volume 1 - Thermal energy systems

NASA Astrophysics Data System (ADS)

Among the topics discussed are the nuclear fuel cycle, advanced nuclear reactor designs, developments in central status power reactors, space nuclear reactors, magnetohydrodynamic devices, thermionic devices, thermoelectric devices, geothermal systems, solar thermal energy conversion systems, ocean thermal energy conversion (OTEC) developments, and advanced energy conversion concepts. Among the specific questions covered under these topic headings are a design concept for an advanced light water breeder reactor, energy conversion in MW-sized space power systems, directionally solidified cermet electrodes for thermionic energy converters, boron-based high temperature thermoelectric materials, geothermal energy commercialization, solar Stirling cycle power conversion, and OTEC production of methanol. For individual items see A84-30027 to A84-30055

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

None

Progress is reported on fundamental research in: crystal physics, reactions at metal surfaces, spectroscopy of ionic media, structure of metals, theory of alloying, physical properties, sintering, deformation of crystalline solids, x ray diffraction, metallurgy of superconducting materials, and electron microscope studies. Long-randge applied research studies were conducted for: zirconium metallurgy, materials compatibility, solid reactions, fuel element development, mechanical properties, non-destructive testing, and high-temperature materials. Reactor development support work was carried out for: gas-cooled reactor program, molten-salt reactor, high-flux isotope reactor, space-power program, thorium-utilization program, advanced-test reactor, Army Package Power Reactor, Enrico Fermi fast-breeder reactor, and water desalination program. Other programmore » activities, for which research was conducted, included: thermonuclear project, transuraniunn program, and post-irradiation examination laboratory. Separate abstracts were prepared for 30 sections of the report. (B.O.G.)« 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.

Movable-molybdenum-reflector reactivity experiments for control studies of compact space power reactor concepts

NASA Technical Reports Server (NTRS)

Fox, T. A.

1973-01-01

An experimental reflector reactivity study was made with a compact cylindrical reactor using a uranyl fluoride - water fuel solution. The reactor was axially unreflected and radially reflected with segments of molybdenum. The reflector segments were displaced incrementally in both the axial and radial dimensions, and the shutdown of each configuration was measured by using the pulsed-neutron source technique. The reactivity effects for axial and radial displacement of reflector segments are tabulated separately and compared. The experiments provide data for control-system studies of compact-space-power-reactor concepts.

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.

Continuous production of butanol from starch-based packing peanuts.

PubMed

Ezeji, Thaddeus C; Groberg, Marisa; Qureshi, Nasib; Blaschek, Hans P

2003-01-01

Acetone, butanol, ethanol (ABE, or solvents) were produced from starch-based packing peanuts in batch and continuous reactors. In a batch reactor, 18.9 g/L of total ABE was produced from 80 g/L packing peanuts in 110 h of fermentation. The initial and final starch concentrations were 69.6 and 11.1 g/L, respectively. In this fermentation, ABE yield and productivity of 0.32 and 0.17 g/(L h) were obtained, respectively. Compared to the batch fermentation, continuous fermentation of 40 g/L of starchbased packing peanuts in P2 medium resulted in a maximum solvent production of 8.4 g/L at a dilution rate of 0.033 h-1. This resulted in a productivity of 0.27 g/(L h). However, the reactor was not stable and fermentation deteriorated with time. Continuous fermentation of 35 g/L of starch solution resulted in a similar performance. These studies were performed in a vertical column reactor using Clostridium beijerinckii BA101 and P2 medium. It is anticipated that prolonged exposure of culture to acrylamide, which is formed during boiling/autoclaving of starch, affects the fermentation negatively.

NASA's Kilopower Reactor Development and the Path to Higher Power Missions

NASA Technical Reports Server (NTRS)

Gibson, Marc A.; Oleson, Steven R.; Poston, David I.; McClure, Patrick

2017-01-01

The development of NASAs Kilopower fission reactor is taking large strides toward flight development with several successful tests completed during its technology demonstration trials. The Kilopower reactors are designed to provide 1-10 kW of electrical power to a spacecraft which could be used for additional science instruments as well as the ability to power electric propulsion systems. Power rich nuclear missions have been excluded from NASA proposals because of the lack of radioisotope fuel and the absence of a flight qualified fission system. NASA has partnered with the Department of Energy's National Nuclear Security Administration to develop the Kilopower reactor using existing facilities and infrastructure to determine if the design is ready for flight development. The 3-year Kilopower project started in 2015 with a challenging goal of building and testing a full-scale flight prototypic nuclear reactor by the end of 2017. As the date approaches, the engineering team shares information on the progress of the technology as well as the enabling capabilities it provides for science and human exploration.

NASA's Kilopower Reactor Development and the Path to Higher Power Missions

NASA Technical Reports Server (NTRS)

Gibson, Marc A.; Oleson, Steven R.; Poston, Dave I.; McClure, Patrick

2017-01-01

The development of NASA's Kilopower fission reactor is taking large strides toward flight development with several successful tests completed during its technology demonstration trials. The Kilopower reactors are designed to provide 1-10 kW of electrical power to a spacecraft which could be used for additional science instruments as well as the ability to power electric propulsion systems. Power rich nuclear missions have been excluded from NASA proposals because of the lack of radioisotope fuel and the absence of a flight qualified fission system. NASA has partnered with the Department of Energy's National Nuclear Security Administration to develop the Kilopower reactor using existing facilities and infrastructure to determine if the design is ready for flight development. The 3-year Kilopower project started in 2015 with a challenging goal of building and testing a full-scale flight prototypic nuclear reactor by the end of 2017. As the date approaches, the engineering team shares information on the progress of the technology as well as the enabling capabilities it provides for science and human exploration.

Fission Surface Power Technology Development Update

NASA Technical Reports Server (NTRS)

Palac, Donald T.; Mason, Lee S.; Houts, Michael G.; Harlow, Scott

2011-01-01

Power is a critical consideration in planning exploration of the surfaces of the Moon, Mars, and places beyond. Nuclear power is an important option, especially for locations in the solar system where sunlight is limited or environmental conditions are challenging (e.g., extreme cold, dust storms). NASA and the Department of Energy are maintaining the option for fission surface power for the Moon and Mars by developing and demonstrating technology for a fission surface power system. The Fission Surface Power Systems project has focused on subscale component and subsystem demonstrations to address the feasibility of a low-risk, low-cost approach to space nuclear power for surface missions. Laboratory demonstrations of the liquid metal pump, reactor control drum drive, power conversion, heat rejection, and power management and distribution technologies have validated that the fundamental characteristics and performance of these components and subsystems are consistent with a Fission Surface Power preliminary reference concept. In addition, subscale versions of a non-nuclear reactor simulator, using electric resistance heating in place of the reactor fuel, have been built and operated with liquid metal sodium-potassium and helium/xenon gas heat transfer loops, demonstrating the viability of establishing system-level performance and characteristics of fission surface power technologies without requiring a nuclear reactor. While some component and subsystem testing will continue through 2011 and beyond, the results to date provide sufficient confidence to proceed with system level technology readiness demonstration. To demonstrate the system level readiness of fission surface power in an operationally relevant environment (the primary goal of the Fission Surface Power Systems project), a full scale, 1/4 power Technology Demonstration Unit (TDU) is under development. The TDU will consist of a non-nuclear reactor simulator, a sodium-potassium heat transfer loop, a power conversion unit with electrical controls, and a heat rejection system with a multi-panel radiator assembly. Testing is planned at the Glenn Research Center Vacuum Facility 6 starting in 2012, with vacuum and liquid-nitrogen cold walls to provide simulation of operationally relevant environments. A nominal two-year test campaign is planned including a Phase 1 reactor simulator and power conversion test followed by a Phase 2 integrated system test with radiator panel heat rejection. The testing is expected to demonstrate the readiness and availability of fission surface power as a viable power system option for NASA's exploration needs. In addition to surface power, technology development work within this project is also directly applicable to in-space fission power and propulsion systems.

Mars power system concept definition study. Volume 1: Study results

NASA Technical Reports Server (NTRS)

Littman, Franklin D.

1994-01-01

A preliminary top level study was completed to define power system concepts applicable to Mars surface applications. This effort included definition of power system requirements and selection of power systems with the potential for high commonality. These power systems included dynamic isotope, Proton Exchange Membrane (PEM) regenerative fuel cell, sodium sulfur battery, photovoltaic, and reactor concepts. Design influencing factors were identified. Characterization studies were then done for each concept to determine system performance, size/volume, and mass. Operations studies were done to determine emplacement/deployment maintenance/servicing, and startup/shutdown requirements. Technology development roadmaps were written for each candidate power system (included in Volume 2). Example power system architectures were defined and compared on a mass basis. The dynamic isotope power system and nuclear reactor power system architectures had significantly lower total masses than the photovoltaic system architectures. Integrated development and deployment time phasing plans were completed for an example DIPS and reactor architecture option to determine the development strategies required to meet the mission scenario requirements.

A fungal biofilm reactor based on metal structured packing improves the quality of a Gla::GFP fusion protein produced by Aspergillus oryzae.

PubMed

Zune, Q; Delepierre, A; Gofflot, S; Bauwens, J; Twizere, J C; Punt, P J; Francis, F; Toye, D; Bawin, T; Delvigne, F

2015-08-01

Fungal biofilm is known to promote the excretion of secondary metabolites in accordance with solid-state-related physiological mechanisms. This work is based on the comparative analysis of classical submerged fermentation with a fungal biofilm reactor for the production of a Gla::green fluorescent protein (GFP) fusion protein by Aspergillus oryzae. The biofilm reactor comprises a metal structured packing allowing the attachment of the fungal biomass. Since the production of the target protein is under the control of the promoter glaB, specifically induced in solid-state fermentation, the biofilm mode of culture is expected to enhance the global productivity. Although production of the target protein was enhanced by using the biofilm mode of culture, we also found that fusion protein production is also significant when the submerged mode of culture is used. This result is related to high shear stress leading to biomass autolysis and leakage of intracellular fusion protein into the extracellular medium. Moreover, 2-D gel electrophoresis highlights the preservation of fusion protein integrity produced in biofilm conditions. Two fungal biofilm reactor designs were then investigated further, i.e. with full immersion of the packing or with medium recirculation on the packing, and the scale-up potentialities were evaluated. In this context, it has been shown that full immersion of the metal packing in the liquid medium during cultivation allows for a uniform colonization of the packing by the fungal biomass and leads to a better quality of the fusion protein.

Current status of SPINNORs designs

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

Su'ud, Zaki

2010-06-22

This study discuss about the SPINNOR (Small Power Reactor, Indonesia, No On-site Refuelling) and the VSPINNOR (Very Small Power Reactor, Indonesia, No On-site Refuelling) which are small lead-bismuth cooled nuclear power reactors with fast neutron spectrum that could be operated for more than 10 or 15 years without on-site refuelling. They are based on the concept of a long-life core reactor developed in Indonesia since early 1990 in collaboration with the Research Laboratory for Nuclear Reactors of the Tokyo Institute of Technology (RLNR TITech). The reactor cores are designed to have near zero (less then one effective delayed neutron fraction)more » burn-up reactivity swing during the whole course of their operation to avoid a possibility of prompt criticality accident. The basic concept is that central region of the reactor core is filled with fertile (blanket) material. During the reactor operation fissile material accumulates in this central region, which helps to compensate fissile material loss in the peripheral core region and also contributes to negative coolant loss reactivity effect. A concept of high fuel volume fraction in the core is applied to achieve smaller size of a critical reactor. In this paper we consider to add Np-237 to the fuel to enhance non proliferation characteristics of the systems. The effect of Np-237 amount variation is discussed.« less

Pressure polymerization of polyester

DOEpatents

Maurer, Charles J.; Shaw, Gordon; Smith, Vicky S.; Buelow, Steven J.; Tumas, William; Contreras, Veronica; Martinez, Ronald J.

2000-08-29

A process is disclosed for the preparation of a polyester polymer or polyester copolymer under superatmospheric pressure conditions in a pipe or tubular reaction under turbannular flow conditions. Reaction material having a glycol equivalents to carboxylic acid equivalents mole ratio of from 1.0:1 to 1.2:1, together with a superatmospheric dense gaseous medium are fed co-currently to the reactor. Dicarboxylic acid and/or diol raw materials may be injected into any of the reaction zones in the process during operation to achieve the overall desired mole ratio balance. The process operates at temperatures of from about 220.degree. C. to about 320.degree. C., with turbannular flow achieved before the polymer product and gas exit the reactor process. The pressure in the reaction zones can be in the range from 15 psia to 2500 psia. A polymer product having a DP of a greater than 40, more preferably at least about 70, is achieved by the transfer of water from the reacting material polymer melt to the gaseous medium in the reactor.

Green synthesis of isopropyl myristate in novel single phase medium Part II: Packed bed reactor (PBR) studies.

PubMed

Vadgama, Rajeshkumar N; Odaneth, Annamma A; Lali, Arvind M

2015-12-01

Isopropyl myristate is a useful functional molecule responding to the requirements of numerous fields of application in cosmetic, pharmaceutical and food industry. In the present work, lipase-catalyzed production of isopropyl myristate by esterification of myristic acid with isopropyl alcohol (molar ratio of 1:15) in the homogenous reaction medium was performed on a bench-scale packed bed reactors, in order to obtain suitable reaction performance data for upscaling. An immobilized lipase B from Candida antartica was used as the biocatalyst based on our previous study. The process intensification resulted in a clean and green synthesis process comprising a series of packed bed reactors of immobilized enzyme and water dehydrant. In addition, use of the single phase reaction system facilitates efficient recovery of the product with no effluent generated and recyclability of unreacted substrates. The single phase reaction system coupled with a continuous operating bioreactor ensures a stable operational life for the enzyme.

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 efficiencies in the range of 40 to 50% can be achieved.« less

Multi-physics design and analyses of long life reactors for lunar outposts

NASA Astrophysics Data System (ADS)

Schriener, Timothy M.

Future human exploration of the solar system is likely to include establishing permanent outposts on the surface of the Moon. These outposts will require reliable sources of electrical power in the range of 10's to 100's of kWe to support exploration and resource utilization activities. This need is best met using nuclear reactor power systems which can operate steadily throughout the long ˜27.3 day lunar rotational period, irrespective of location. Nuclear power systems can potentially open up the entire lunar surface for future exploration and development. Desirable features of nuclear power systems for the lunar surface include passive operation, the avoidance of single point failures in reactor cooling and the integrated power system, moderate operating temperatures to enable the use of conventional materials with proven irradiation experience, utilization of the lunar regolith for radiation shielding and as a supplemental neutron reflector, and safe post-operation decay heat removal and storage for potential retrieval. In addition, it is desirable for the reactor to have a long operational life. Only a limited number of space nuclear reactor concepts have previously been developed for the lunar environment, and these designs possess only a few of these desirable design and operation features. The objective of this research is therefore to perform design and analyses of long operational life lunar reactors and power systems which incorporate the desirable features listed above. A long reactor operational life could be achieved either by increasing the amount of highly enriched uranium (HEU) fuel in the core or by improving the neutron economy in the reactor through reducing neutron leakage and parasitic absorption. The amount of fuel in surface power reactors is constrained by the launch safety requirements. These include ensuring that the bare reactor core remains safely subcritical when submerged in water or wet sand and flooded with seawater in the unlikely event of a launch abort accident. Increasing the amount of fuel in the reactor core, and hence its operational life, would be possible by launching the reactor unfueled and fueling it on the Moon. Such a reactor would, thus, not be subject to launch criticality safety requirements. However, loading the reactor with fuel on the Moon presents a challenge, requiring special designs of the core and the fuel elements, which lend themselves to fueling on the lunar surface. This research investigates examples of both a solid core reactor that would be fueled at launch as well as an advanced concept which could be fueled on the Moon. Increasing the operational life of a reactor fueled at launch is exercised for the NaK-78 cooled Sectored Compact Reactor (SCoRe). A multi-physics design and analyses methodology is developed which iteratively couples together detailed Monte Carlo neutronics simulations with 3-D Computational Fluid Dynamics (CFD) and thermal-hydraulics analyses. Using this methodology the operational life of this compact, fast spectrum reactor is increased by reconfiguring the core geometry to reduce neutron leakage and parasitic absorption, for the same amount of HEU in the core, and meeting launch safety requirements. The multi-physics analyses determine the impacts of the various design changes on the reactor's neutronics and thermal-hydraulics performance. The option of increasing the operational life of a reactor by loading it on the Moon is exercised for the Pellet Bed Reactor (PeBR). The PeBR uses spherical fuel pellets and is cooled by He-Xe gas, allowing the reactor core to be loaded with fuel pellets and charged with working fluid on the lunar surface. The performed neutronics analyses ensure the PeBR design achieves a long operational life, and develops safe launch canister designs to transport the spherical fuel pellets to the lunar surface. The research also investigates loading the PeBR core with fuel pellets on the Moon using a transient Discrete Element Method (DEM) analysis in lunar gravity. In addition, this research addresses the post-operation storage of the SCoRe and PeBR concepts, below the lunar surface, to determine the time required for the radioactivity in the used fuel to decrease to a low level to allow for its safe recovery. The SCoRe and PeBR concepts are designed to operate at coolant temperatures ≤ 900 K and use conventional stainless steels and superalloys for the structure in the reactor core and power system. They are emplaced below grade on the Moon to take advantage of the regolith as a supplemental neutron reflector and as shielding of the lunar outpost from the reactors' neutron and gamma radiation.

Liquid Metal Pump Technologies for Nuclear Surface Power

NASA Technical Reports Server (NTRS)

Polzin, Kurt A.

2007-01-01

Multiple liquid metal pump options are reviewed for the purpose of determining the technologies that are best suited for inclusion in a nuclear reactor thermal simulator intended to rest prototypical space nuclear surface power system components. Conduction, induction and thermoelectric electromagnetic pumps are evaluated based on their performance characteristics and the technical issues associated with incorporation into a reactor system. A thermoelectric electromagnetic pump is selected as the best option for use in NASA-MSFC's Fission Surface Power-Primary Test Circuit reactor simulator based on its relative simplicity, low power supply mass penalty, flight heritage, and the promise of increased pump efficiency over those earlier pump designs through the use of skutterudite thermoelectric elements.

Threshold self-powered gamma detector for use as a monitor of power in a nuclear reactor

DOEpatents

LeVert, Francis E.; Cox, Samson A.

1978-01-01

A self-powered gamma monitor for placement near the core of a nuclear reactor comprises a lead prism surrounded by a coaxial thin nickel sheet, the combination forming a collector. A coaxial polyethylene electron barrier encloses the collector and is separated from the nickel sheet by a vacuum region. The electron barrier is enclosed by a coaxial stainless steel emitter which, in turn, is enclosed within a lead casing. When the detector is placed in a flux of gamma rays, a measure of the current flow in an external circuit between emitter and collector provides a measure of the power level of the reactor.

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

Trianti, Nuri, E-mail: nuri.trianti@gmail.com; Nurjanah,; Su’ud, Zaki

Thermalhydraulic of reactor core is the thermal study on fluids within the core reactor, i.e. analysis of the thermal energy transfer process produced by fission reaction from fuel to the reactor coolant. This study include of coolant temperature and reactor power density distribution. The purposes of this analysis in the design of nuclear power plant are to calculate the coolant temperature distribution and the chimney height so natural circulation could be occurred. This study was used boiling water reactor (BWR) with cylinder type reactor core. Several reactor core properties such as linear power density, mass flow rate, coolant density andmore » inlet temperature has been took into account to obtain distribution of coolant density, flow rate and pressure drop. The results of calculation are as follows. Thermal hydraulic calculations provide the uniform pressure drop of 1.1 bar for each channels. The optimum mass flow rate to obtain the uniform pressure drop is 217g/s. Furthermore, from the calculation it could be known that outlet temperature is 288°C which is the saturated fluid’s temperature within the system. The optimum chimney height for natural circulation within the system is 14.88 m.« less

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 temperature between 550 C and 850 C. The UniSim models used realistic component parameters and operating conditions to model the complete reactor and power conversion systems. CO2 properties were evaluated, and the operating ranges of the cycles were adjusted to take advantage of the rapidly changing properties of CO2 near the critical point. The results of the analyses showed that, for the direct supercritical CO2 power cycle, thermal efficiencies in the range of 40 to 50% can be achieved. For the indirect supercritical CO2 power cycle, thermal efficiencies were approximately 10% lower than those obtained for the direct cycle over the same reactor outlet temperature range.« less

76 FR 39922 - Office of New Reactors; Proposed Revision 4 to Standard Review Plan Section 8.1 on Electric Power...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-07-07

... the proposed Revision 4 to Standard Review Plan Section 8.1 on ``Electric Power-- Introduction.'' The... NUCLEAR REGULATORY COMMISSION [NRC-2011-0119] Office of New Reactors; Proposed Revision 4 to Standard Review Plan Section 8.1 on Electric Power--Introduction, Correction AGENCY: Nuclear Regulatory...

COST FUNCTION STUDIES FOR POWER REACTORS

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

Heestand, J.; Wos, L.T.

1961-11-01

A function to evaluate the cost of electricity produced by a nuclear power reactor was developed. The basic equation, revenue = capital charges + profit + operating expenses, was expanded in terms of various cost parameters to enable analysis of multiregion nuclear reactors with uranium and/or plutonium for fuel. A corresponding IBM 704 computer program, which will compute either the price of electricity or the value of plutonium, is presented in detail. (auth)

Compact power reactor

DOEpatents

Wetch, Joseph R.; Dieckamp, Herman M.; Wilson, Lewis A.

1978-01-01

There is disclosed a small compact nuclear reactor operating in the epithermal neutron energy range for supplying power at remote locations, as for a satellite. The core contains fuel moderator elements of Zr hydride with 7 w/o of 93% enriched uranium alloy. The core has a radial beryllium reflector and is cooled by liquid metal coolant such as NaK. The reactor is controlled and shut down by moving portions of the reflector.

Numerical investigation of flow and heat transfer in a novel configuration multi-tubular fixed bed reactor for propylene to acrolein process

NASA Astrophysics Data System (ADS)

Jiang, Bin; Hao, Li; Zhang, Luhong; Sun, Yongli; Xiao, Xiaoming

2015-01-01

In the present contribution, a numerical study of fluid flow and heat transfer performance in a pilot-scale multi-tubular fixed bed reactor for propylene to acrolein oxidation reaction is presented using computational fluid dynamics (CFD) method. Firstly, a two-dimensional CFD model is developed to simulate flow behaviors, catalytic oxidation reaction, heat and mass transfer adopting porous medium model on tube side to achieve the temperature distribution and investigate the effect of operation parameters on hot spot temperature. Secondly, based on the conclusions of tube-side, a novel configuration multi-tubular fixed-bed reactor comprising 790 tubes design with disk-and-doughnut baffles is proposed by comparing with segmental baffles reactor and their performance of fluid flow and heat transfer is analyzed to ensure the uniformity condition using molten salt as heat carrier medium on shell-side by three-dimensional CFD method. The results reveal that comprehensive performance of the reactor with disk-and-doughnut baffles is better than that of with segmental baffles. Finally, the effects of operating conditions to control the hot spots are investigated. The results show that the flow velocity range about 0.65 m/s is applicable and the co-current cooling system flow direction is better than counter-current flow to control the hottest temperature.

Simplifying microbial electrosynthesis reactor design.

PubMed

Giddings, Cloelle G S; Nevin, Kelly P; Woodward, Trevor; Lovley, Derek R; Butler, Caitlyn S

2015-01-01

Microbial electrosynthesis, an artificial form of photosynthesis, can efficiently convert carbon dioxide into organic commodities; however, this process has only previously been demonstrated in reactors that have features likely to be a barrier to scale-up. Therefore, the possibility of simplifying reactor design by both eliminating potentiostatic control of the cathode and removing the membrane separating the anode and cathode was investigated with biofilms of Sporomusa ovata. S. ovata reduces carbon dioxide to acetate and acts as the microbial catalyst for plain graphite stick cathodes as the electron donor. In traditional 'H-cell' reactors, where the anode and cathode chambers were separated with a proton-selective membrane, the rates and columbic efficiencies of microbial electrosynthesis remained high when electron delivery at the cathode was powered with a direct current power source rather than with a potentiostat-poised cathode utilized in previous studies. A membrane-less reactor with a direct-current power source with the cathode and anode positioned to avoid oxygen exposure at the cathode, retained high rates of acetate production as well as high columbic and energetic efficiencies. The finding that microbial electrosynthesis is feasible without a membrane separating the anode from the cathode, coupled with a direct current power source supplying the energy for electron delivery, is expected to greatly simplify future reactor design and lower construction costs.

Neutronic design studies of a conceptual DCLL fusion reactor for a DEMO and a commercial power plant

NASA Astrophysics Data System (ADS)

Palermo, I.; Veredas, G.; Gómez-Ros, J. M.; Sanz, J.; Ibarra, A.

2016-01-01

Neutronic analyses or, more widely, nuclear analyses have been performed for the development of a dual-coolant He/LiPb (DCLL) conceptual design reactor. A detailed three-dimensional (3D) model has been examined and optimized. The design is based on the plasma parameters and functional materials of the power plant conceptual studies (PPCS) model C. The initial radial-build for the detailed model has been determined according to the dimensions established in a previous work on an equivalent simplified homogenized reactor model. For optimization purposes, the initial specifications established over the simplified model have been refined on the detailed 3D design, modifying material and dimension of breeding blanket, shield and vacuum vessel in order to fulfil the priority requirements of a fusion reactor in terms of the fundamental neutronic responses. Tritium breeding ratio, energy multiplication factor, radiation limits in the TF coils, helium production and displacements per atom (dpa) have been calculated in order to demonstrate the functionality and viability of the reactor design in guaranteeing tritium self-sufficiency, power efficiency, plasma confinement, and re-weldability and structural integrity of the components. The paper describes the neutronic design improvements of the DCLL reactor, obtaining results for both DEMO and power plant operational scenarios.

Feasibility study of a magnetic fusion production reactor

NASA Astrophysics Data System (ADS)

Moir, R. W.

1986-12-01

A magnetic fusion reactor can produce 10.8 kg of tritium at a fusion power of only 400 MW —an order of magnitude lower power than that of a fission production reactor. Alternatively, the same fusion reactor can produce 995 kg of plutonium. Either a tokamak or a tandem mirror production plant can be used for this purpose; the cost is estimated at about 1.4 billion (1982 dollars) in either case. (The direct costs are estimated at 1.1 billion.) The production cost is calculated to be 22,000/g for tritium and 260/g for plutonium of quite high purity (1%240Pu). Because of the lack of demonstrated technology, such a plant could not be constructed today without significant risk. However, good progress is being made in fusion technology and, although success in magnetic fusion science and engineering is hard to predict with assurance, it seems possible that the physics basis and much of the needed technology could be demonstrated in facilities now under construction. Most of the remaining technology could be demonstrated in the early 1990s in a fusion test reactor of a few tens of megawatts. If the Magnetic Fusion Energy Program constructs a fusion test reactor of approximately 400 MW of fusion power as a next step in fusion power development, such a facility could be used later as a production reactor in a spinoff application. A construction decision in the late 1980s could result in an operating production reactor in the late 1990s. A magnetic fusion production reactor (MFPR) has four potential advantages over a fission production reactor: (1) no fissile material input is needed; (2) no fissioning exists in the tritium mode and very low fissioning exists in the plutonium mode thus avoiding the meltdown hazard; (3) the cost will probably be lower because of the smaller thermal power required; (4) and no reprocessing plant is needed in the tritium mode. The MFPR also has two disadvantages: (1) it will be more costly to operate because it consumes rather than sells electricity, and (2) there is a risk of not meeting the design goals.

When Do Commercial Reactors Permanently Shut Down?

EIA Publications

2011-01-01

For those wishing to obtain current data, the following resources are available: U.S. reactors, go to the Energy Information Administration's nuclear reactor shutdown list. (Note: As of April 30, 2010, the last U.S. reactor to permanently shut down was Big Rock Point in 1997.) Foreign Reactors, go to the Power Reactor Information System (PRIS) on the International Atomic Energy Agency's website.

10 CFR 140.11 - Amounts of financial protection for certain reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

...,000,000 for each nuclear reactor he is authorized to operate at a thermal power level not exceeding ten kilowatts; (2) In the amount of $1,500,000 for each nuclear reactor he is authorized to operate at... amount of $2,500,000 for each nuclear reactor other than a testing reactor or a reactor licensed under...

10 CFR 140.11 - Amounts of financial protection for certain reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

...,000,000 for each nuclear reactor he is authorized to operate at a thermal power level not exceeding ten kilowatts; (2) In the amount of $1,500,000 for each nuclear reactor he is authorized to operate at... amount of $2,500,000 for each nuclear reactor other than a testing reactor or a reactor licensed under...

10 CFR 140.11 - Amounts of financial protection for certain reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

...,000,000 for each nuclear reactor he is authorized to operate at a thermal power level not exceeding ten kilowatts; (2) In the amount of $1,500,000 for each nuclear reactor he is authorized to operate at... amount of $2,500,000 for each nuclear reactor other than a testing reactor or a reactor licensed under...

A fission-fusion hybrid reactor in steady-state L-mode tokamak configuration with natural uranium

NASA Astrophysics Data System (ADS)

Reed, Mark; Parker, Ronald R.; Forget, Benoit

2012-06-01

This work develops a conceptual design for a fusion-fission hybrid reactor operating in steady-state L-mode tokamak configuration with a subcritical natural or depleted uranium pebble bed blanket. A liquid lithium-lead alloy breeds enough tritium to replenish that consumed by the D-T fusion reaction. The fission blanket augments the fusion power such that the fusion core itself need not have a high power gain, thus allowing for fully non-inductive (steady-state) low confinement mode (L-mode) operation at relatively small physical dimensions. A neutron transport Monte Carlo code models the natural uranium fission blanket. Maximizing the fission power gain while breeding sufficient tritium allows for the selection of an optimal set of blanket parameters, which yields a maximum prudent fission power gain of approximately 7. A 0-D tokamak model suffices to analyze approximate tokamak operating conditions. This fission blanket would allow the fusion component of a hybrid reactor with the same dimensions as ITER to operate in steady-state L-mode very comfortably with a fusion power gain of 6.7 and a thermal fusion power of 2.1 GW. Taking this further can determine the approximate minimum scale for a steady-state L-mode tokamak hybrid reactor, which is a major radius of 5.2 m and an aspect ratio of 2.8. This minimum scale device operates barely within the steady-state L-mode realm with a thermal fusion power of 1.7 GW. Basic thermal hydraulic analysis demonstrates that pressurized helium could cool the pebble bed fission blanket with a flow rate below 10 m/s. The Brayton cycle thermal efficiency is 41%. This reactor, dubbed the Steady-state L-mode non-Enriched Uranium Tokamak Hybrid (SLEUTH), with its very fast neutron spectrum, could be superior to pure fission reactors in terms of breeding fissile fuel and transmuting deleterious fission products. It would likely function best as a prolific plutonium breeder, and the plutonium it produces could actually be more proliferation-resistant than that bred by conventional fast reactors. Furthermore, it can maintain constant total hybrid power output as burnup proceeds by varying the neutron source strength.

10 CFR 73.60 - Additional requirements for physical protection at nonpower reactors.

Code of Federal Regulations, 2010 CFR

2010-01-01

... nonpower reactors. 73.60 Section 73.60 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION... requirements for physical protection at nonpower reactors. Each nonpower reactor licensee who, pursuant to the... nonpower reactors licensed to operate at or above a power level of 2 megawatts thermal. [38 FR 35430, Dec...

10 CFR 73.60 - Additional requirements for physical protection at nonpower reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... nonpower reactors. 73.60 Section 73.60 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION... requirements for physical protection at nonpower reactors. Each nonpower reactor licensee who, pursuant to the... nonpower reactors licensed to operate at or above a power level of 2 megawatts thermal. [38 FR 35430, Dec...

PBF Reactor Building (PER620). Camera faces north into highbay/reactor pit ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

PBF Reactor Building (PER-620). Camera faces north into high-bay/reactor pit area. Inside from for reactor enclosure is in place. Photographer: John Capek. Date: March 15, 1967. INEEL negative no. 67-1769 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

World Energy Data System (WENDS). Volume XI. Nuclear fission program summaries

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

Not Available

1979-06-01

Brief management and technical summaries of nuclear fission power programs are presented for nineteen countries. The programs include the following: fuel supply, resource recovery, enrichment, fuel fabrication, light water reactors, heavy water reactors, gas cooled reactors, breeder reactors, research and test reactors, spent fuel processing, waste management, and safety and environment. (JWR)

Analysis of the Gas Core Actinide Transmutation Reactor (GCATR)

NASA Technical Reports Server (NTRS)

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

1977-01-01

Design power plant studies were carried out for two applications of the plasma core reactor: (1) As a breeder reactor, (2) As a reactor able to transmute actinides effectively. In addition to the above applications the reactor produced electrical power with a high efficiency. A reactor subsystem was designed for each of the two applications. For the breeder reactor, neutronics calculations were carried out for a U-233 plasma core with a molten salt breeding blanket. A reactor was designed with a low critical mass (less than a few hundred kilograms U-233) and a breeding ratio of 1.01. The plasma core actinide transmutation reactor was designed to transmute the nuclear waste from conventional LWR's. The spent fuel is reprocessed during which 100% of Np, Am, Cm, and higher actinides are separated from the other components. These actinides are then manufactured as oxides into zirconium clad fuel rods and charged as fuel assemblies in the reflector region of the plasma core actinide transmutation reactor. In the equilibrium cycle, about 7% of the actinides are directly fissioned away, while about 31% are removed by reprocessing.

MONTE CARLO SIMULATIONS OF PERIODIC PULSED REACTOR WITH MOVING GEOMETRY PARTS

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

Cao, Yan; Gohar, Yousry

2015-11-01

In a periodic pulsed reactor, the reactor state varies periodically from slightly subcritical to slightly prompt supercritical for producing periodic power pulses. Such periodic state change is accomplished by a periodic movement of specific reactor parts, such as control rods or reflector sections. The analysis of such reactor is difficult to perform with the current reactor physics computer programs. Based on past experience, the utilization of the point kinetics approximations gives considerable errors in predicting the magnitude and the shape of the power pulse if the reactor has significantly different neutron life times in different zones. To accurately simulate themore » dynamics of this type of reactor, a Monte Carlo procedure using the transfer function TRCL/TR of the MCNP/MCNPX computer programs is utilized to model the movable reactor parts. In this paper, two algorithms simulating the geometry part movements during a neutron history tracking have been developed. Several test cases have been developed to evaluate these procedures. The numerical test cases have shown that the developed algorithms can be utilized to simulate the reactor dynamics with movable geometry parts.« less

The Simulator Development for RDE Reactor

NASA Astrophysics Data System (ADS)

Subekti, Muhammad; Bakhri, Syaiful; Sunaryo, Geni Rina

2018-02-01

BATAN is proposing the construction of experimental power reactor (RDE reactor) for increasing the public acceptance on NPP development plan, proofing the safety level of the most advanced reactor by performing safety demonstration on the accidents such as Chernobyl and Fukushima, and owning the generation fourth (G4) reactor technology. For owning the reactor technology, the one of research activities is RDE’s simulator development that employing standard equation. The development utilizes standard point kinetic and thermal equation. The examination of the simulator carried out comparison in which the simulation’s calculation result has good agreement with assumed parameters and ChemCAD calculation results. The transient simulation describes the characteristic of the simulator to respond the variation of power increase of 1.5%/min, 2.5%/min, and 3.5%/min.

Neutron flux and power in RTP core-15

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

Rabir, Mohamad Hairie, E-mail: m-hairie@nuclearmalaysia.gov.my; Zin, Muhammad Rawi Md; Usang, Mark Dennis

PUSPATI TRIGA Reactor achieved initial criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes. This paper describes the reactor parameters calculation for the PUSPATI TRIGA REACTOR (RTP); focusing on the application of the developed reactor 3D model for criticality calculation, analysis of power and neutron flux distribution of TRIGA core. The 3D continuous energy Monte Carlo code MCNP was used to develop a versatile and accurate full model of the TRIGA reactor. The model represents in detailed all important components of the core withmore » literally no physical approximation. The consistency and accuracy of the developed RTP MCNP model was established by comparing calculations to the available experimental results and TRIGLAV code calculation.« less

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

From the first nuclear power plant to fourth-generation nuclear power installations [on the 60th anniversary of the World's First nuclear power plant

NASA Astrophysics Data System (ADS)

Rachkov, V. I.; Kalyakin, S. G.; Kukharchuk, O. F.; Orlov, Yu. I.; Sorokin, A. P.

2014-05-01

Successful commissioning in the 1954 of the World's First nuclear power plant constructed at the Institute for Physics and Power Engineering (IPPE) in Obninsk signaled a turn from military programs to peaceful utilization of atomic energy. Up to the decommissioning of this plant, the AM reactor served as one of the main reactor bases on which neutron-physical investigations and investigations in solid state physics were carried out, fuel rods and electricity generating channels were tested, and isotope products were bred. The plant served as a center for training Soviet and foreign specialists on nuclear power plants, the personnel of the Lenin nuclear-powered icebreaker, and others. The IPPE development history is linked with the names of I.V. Kurchatov, A.I. Leipunskii, D.I. Blokhintsev, A.P. Aleksandrov, and E.P. Slavskii. More than 120 projects of various nuclear power installations were developed under the scientific leadership of the IPPE for submarine, terrestrial, and space applications, including two water-cooled power units at the Beloyarsk NPP in Ural, the Bilibino nuclear cogeneration station in Chukotka, crawler-mounted transportable TES-3 power station, the BN-350 reactor in Kazakhstan, and the BN-600 power unit at the Beloyarsk NPP. Owing to efforts taken on implementing the program for developing fast-neutron reactors, Russia occupied leading positions around the world in this field. All this time, IPPE specialists worked on elaborating the principles of energy supertechnologies of the 21st century. New large experimental installations have been put in operation, including the nuclear-laser setup B, the EGP-15 accelerator, the large physical setup BFS, the high-pressure setup SVD-2; scientific, engineering, and technological schools have been established in the field of high- and intermediate-energy nuclear physics, electrostatic accelerators of multicharge ions, plasma processes in thermionic converters and nuclear-pumped lasers, physics of compact nuclear reactors and radiation protection, thermal physics, physical chemistry and technology of liquid metal coolants, and physics of radiation-induced defects, and radiation materials science. The activity of the institute is aimed at solving matters concerned with technological development of large-scale nuclear power engineering on the basis of a closed nuclear fuel cycle with the use of fast-neutron reactors (referred to henceforth as fast reactors), development of innovative nuclear and conventional technologies, and extension of their application fields.

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

10 CFR 52.93 - Exemptions and variances.

Code of Federal Regulations, 2010 CFR

2010-01-01

... referencing a nuclear power reactor manufactured under a manufacturing license issued under subpart F of this... NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSES, CERTIFICATIONS, AND APPROVALS FOR NUCLEAR POWER PLANTS..., site parameters, terms and conditions, or approved design of the manufactured reactor. The Commission...

Operating manual for the Bulk Shielding Reactor

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

Not Available

1983-04-01

The BSR is a pool-type reactor. It has the capabilities of continuous operation at a power level of 2 MW or at any desired lower power level. This manual presents descriptive and operational information. The reactor and its auxillary facilities are described from physical and operational viewpoints. Detailed operating procedures are included which are applicable from source-level startup to full-power operation. Also included are procedures relative to the safety of personnel and equipment in the areas of experiments, radiation and contamination control, emergency actions, and general safety. This manual supercedes all previous operating manuals for the BSR.

Operating manual for the Bulk Shielding Reactor

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

Not Available

1987-03-01

The BSR is a pool-type reactor. It has the capabilities of continuous operation at a power level of 2 MW or at any desired lower power level. This manual presents descriptive and operational information. The reactor and its auxiliary facilities are described from physical and operational viewpoints. Detailed operating procedures are included which are applicable from source-level startup to full-power operation. Also included are procedures relative to the safety of personnel and equipment in the areas of experiments, radiation and contamination control, emergency actions, and general safety. This manual supersedes all previous operating manuals for the BSR.

Thermal-Hydraulic Design of a Fluoride High-Temperature Demonstration Reactor

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

Carbajo, Juan J; Qualls, A L

2016-01-01

INTRODUCTION The Fluoride High-Temperature Reactor (FHR) named the Demonstration Reactor (DR) is a novel reactor concept using molten salt coolant and TRIstructural ISOtropic (TRISO) fuel that is being developed at Oak Ridge National Laboratory (ORNL). The objective of the FHR DR is to advance the technology readiness level of FHRs. The FHR DR will demonstrate technologies needed to close remaining gaps to commercial viability. The FHR DR has a thermal power of 100 MWt, very similar to the SmAHTR, another FHR ORNL concept (Refs. 1 and 2) with a power of 125 MWt. The FHR DR is also a smallmore » version of the Advanced High Temperature Reactor (AHTR), with a power of 3400 MWt, cooled by a molten salt and also being developed at ORNL (Ref. 3). The FHR DR combines three existing technologies: (1) high-temperature, low-pressure molten salt coolant, (2) high-temperature coated-particle TRISO fuel, (3) and passive decay heat cooling systems by using Direct Reactor Auxiliary Cooling Systems (DRACS). This paper presents FHR DR thermal-hydraulic design calculations.« 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.

Developing the European Center of Competence on VVER-type nuclear power reactors

NASA Astrophysics Data System (ADS)

Geraskin, Nikolay; Pironkov, Lyubomir; Kulikov, Evgeny; Glebov, Vasily

2017-09-01

This paper presents the results of the European educational projects CORONA and CORONA-II which are dedicated to preserving and further developing nuclear knowledge and competencies in the area of VVER-type nuclear power reactors technologies (Water-Water Energetic Reactor, WWER or VVER). The development of the European Center of Competence for VVER-technology is focused on master's degree programmes. The specifics of a systematic approach to training in the area of VVER-type nuclear power reactors technologies are analysed. This paper discusses enhancement of the training opportunities of the European Center that have arisen from advances in methodology and distance education. With a special attention paid to the European Nuclear Education Network (ENEN), the possibilities of further development of the international cooperation between European countries and educational institutions are examined.

Closed Brayton Cycle power system with a high temperature pellet bed reactor heat source for NEP applications

NASA Technical Reports Server (NTRS)

Juhasz, Albert J.; El-Genk, Mohamed S.; Harper, William B., Jr.

1992-01-01

Capitalizing on past and future development of high temperature gas reactor (HTGR) technology, a low mass 15 MWe closed gas turbine cycle power system using a pellet bed reactor heating helium working fluid is proposed for Nuclear Electric Propulsion (NEP) applications. Although the design of this directly coupled system architecture, comprising the reactor/power system/space radiator subsystems, is presented in conceptual form, sufficient detail is included to permit an assessment of overall system performance and mass. Furthermore, an attempt is made to show how tailoring of the main subsystem design characteristics can be utilized to achieve synergistic system level advantages that can lead to improved reliability and enhanced system life while reducing the number of parasitic load driven peripheral subsystems.

Update on reactors and reactor instruments in Asia

NASA Astrophysics Data System (ADS)

Rao, K. R.

1991-10-01

The 1980s have seen the commissioning of several medium flux (∼10 14 neutrons/cm 2s) research reactors in Asia. The reactors are based on indigenous design and development in India and China. At Dhruva reactor (India), a variety of neutron spectrometers have been established that have provided useful data related to the structure of high- Tc materials, phonon density of states, magnetic moment distributions and micellar aggregation during the last couple of years. Polarised neutron analysis, neutron interferometry and neutron spin echo methods are some of the new techniques under development. The spectrometers and associated automaton, detectors and neutron guides have all been indigenously developed. This paper summarises the developments and on-going activities in Bangladesh, China, India, Indonesia, Korea, Malaysia, the Philippines and Thailand.

Fusion pumped laser

DOEpatents

Pappas, D.S.

1987-07-31

The apparatus of this invention may comprise a system for generating laser radiation from a high-energy neutron source. The neutron source is a tokamak fusion reactor generating a long pulse of high-energy neutrons and having a temperature and magnetic field effective to generate a neutron flux of at least 10/sup 15/ neutrons/cm/sup 2//center dot/s. Conversion means are provided adjacent the fusion reactor at a location operable for converting the high-energy neutrons to an energy source with an intensity and energy effective to excite a preselected lasing medium. A lasing medium is spaced about and responsive to the energy source to generate a population inversion effective to support laser oscillations for generating output radiation. 2 figs., 2 tabs.

TOPAZ-2 reactor distribution during descent in the atmosphere and at the impact with the Earth surface

NASA Astrophysics Data System (ADS)

Grinberg, Eduard I.; Nikolaev, Vadim S.; Sokolov, Nikolai A.; Doschatov, Vitaly V.; Usov, Veniamin A.; Gulidov, Aleksander I.

1995-01-01

The paper presents results of more accurate computational analysis of the TOPAZ-2 system reactor core aerodynamic disruption at an inadvertent reentry. Given are preliminary results on the pattern of disruption of the core partially burnt during its descent in the atmosphere at its impact on the surface of water and sandstone (medium density concrete).

Quenching of a highly superheated porous medium by injection of water

NASA Astrophysics Data System (ADS)

Fichot, F.; Bachrata, A.; Repetto, G.; Fleurot, J.; Quintard, M.

2012-11-01

Understanding of two-phase flow through porous medium with intense phase change is of interest in many situations, including nuclear, chemical or geophysical applications. Intense boiling occurs when the liquid is injected into a highly superheated medium. Under such conditions, the heat flux extracted by the fluid from the porous medium is mainly governed by the nucleation of bubbles and by the evaporation of thin liquid films. Both configurations are possible, depending on local flow conditions and on the ratio of bubble size to pore size. The present study is motivated by the safety evaluation of light water nuclear reactors in case of a severe accident scenario, such as the one that happened in Fukushima Dai-ichi plant in March, 2011. If water sources are not available for a long period of time, the reactor core heats up due to the residual power and eventually becomes significantly damaged due to intense oxidation of metals and fragmentation of fuel rods resulting in the formation of a porous medium where the particles have a characteristic length-scale of 1 to 5 mm. The coolability of the porous medium will depend on the water flow rate which can enter the medium under the available driving head and on the geometrical features of the porous matrix (average pore size, porosity). Therefore, it is of high interest to evaluate the conditions for which the injection of water in such porous medium is likely to stop the progression of the accident. The present paper addresses the issue of modelling two-phase flow and heat transfers in a porous medium initially dry, where water is injected. The medium is initially at a temperature well above the saturation temperature of water. In a first part, a summary of existing knowledge is provided, showing the scarcity of models and experimental data. In a second part, new experimental results obtained in an IRSN facility are analysed. The experiment consists in a bed of steel particles that are heated up to 700°Cbefore injecting water. The facility is briefly described. The velocity of the "quench front" (location where particles are quickly cooled down) and the total pressure drop across the medium are estimated. The dependencies of those quantities with respect to the inlet water flow rate, the initial temperature of the medium and the diameter of particles are obtained. In a third part, a model is proposed, based on a previously developed model which is improved in order to take into account intense boiling regimes (in particular nucleate boiling). The model includes a function that takes into account the contact area between water and the particles which depends on the temperature of particles and on the void fraction. That function affects the local intensity of phase change. The model involves a few parameters which cannot be evaluated analytically. Those parameters are bounded, following the analysis of experimental data. Finally, the model is assessed by comparison of calculations with those new experimental data. The satisfactory agreement shows that the model is almost predictive in the range of parameters studied. The experimental results also show that the quench front becomes unstable under certain conditions. This is also analysed and compared with the predictions of the model.

Feasibility Study of a Nuclear-Stirling Plant for the Jupiter Icy Moons Orbiter

NASA Technical Reports Server (NTRS)

Schmitz, Paul C.; Schreiber, Jeffrey G.; Penswick, L. Barry

2005-01-01

NASA is undertaking the design of a new spacecraft to explore the planet Jupiter and its three moons Calisto, Ganymede and Europa. This proposed mission, known as Jupiter Icy Moons Orbiter (JIMO) would use a nuclear reactor and an associated electrical generation system (Reactor Power Plant-RPP) to provide power to the spacecraft. The JIMO spacecraft is envisioned to use this power for science and communications as well as Electric Propulsion (EP). Among other potential power-generating concepts, previous studies have considered Thermoelectric and Brayton Power conversion systems, coupled to a liquid metal reactor for the JIMO mission. This paper will explore trades in system mass and radiator area for a nuclear reactor power conversion system, however this study will focus on Stirling power conversion. The Stirling convertor modeled in this study is based upon the Component Test Power Convertor design that was designed and operated successfully under the Civil Space Technology Initiative for use with the SP-100 nuclear reactor i the 1980's and early 1990's. The study design is such that two of the four convertors would operate at any time to generate the 100 kWe while the others are held in reserve. For this study the Stirling convertors hot-side temperature is 1050 K, would operate at a temperature ratio of 2.4 for a minimum mass system and would have a system efficiency of 29%. The Stirling convertor would generate high voltage (400 volt), 100 Hz single phase AC that is supplied to the Power Management and Distribution system. The waste hear is removed from the Stirling convertors by a flowing liquid sodium-potassium eutectic and then rejected by a shared radiator. The radiator consists of two coplanar wings, which would be deployed after the reactor is in space. System trades were performed to vary cycle state point temperatures and convertor design as well as power output. Other redundancy combinations were considered to understand the affects of convertor size and number of spares to the system mass.

Transmutation of Isotopes --- Ecological and Energy Production Aspects

NASA Astrophysics Data System (ADS)

Gudowski, Waclaw

2000-01-01

This paper describes principles of Accelerator-Driven Transmutation of Nuclear Wastes (ATW) and gives some flavour of the most important topics which are today under investigations in many countries. An assessment of the potential impact of ATW on a future of nuclear energy is also given. Nuclear reactors based on self-sustained fission reactions --- after spectacular development in fifties and sixties, that resulted in deployment of over 400 power reactors --- are wrestling today more with public acceptance than with irresolvable technological problems. In a whole spectrum of reasons which resulted in today's opposition against nuclear power few of them are very relevant for the nuclear physics community and they arose from the fact that development of nuclear power had been handed over to the nuclear engineers and technicians with some generically unresolved problems, which should have been solved properly by nuclear scientists. In a certain degree of simplification one can say, that most of the problems originate from very specific features of a fission phenomenon: self-sustained chain reaction in fissile materials and very strong radioactivity of fission products and very long half-life of some of the fission and activation products. And just this enormous concentration of radioactive fission products in the reactor core is the main problem of managing nuclear reactors: it requires unconditional guarantee for the reactor core integrity in order to avoid radioactive contamination of the environment; it creates problems to handle decay heat in the reactor core and finally it makes handling and/or disposal of spent fuel almost a philosophical issue, due to unimaginable long time scales of radioactive decay of some isotopes. A lot can be done to improve the design of conventional nuclear reactors (like Light Water Reactors); new, better reactors can be designed but it seems today very improbable to expect any radical change in the public perception of conventional nuclear power. In this context a lot of hopes and expectations have been expressed for novel systems called Accelerator-Driven Systems, Accelerator-Driven Transmutation of Waste or just Hybrid Reactors. All these names are used for description of the same nuclear system combining a powerful particle accelerator with a subcritical reactor. A careful analysis of possible environmental impact of ATW together with limitation of this technology is presented also in this paper.

10 CFR 100.3 - Definitions.

Code of Federal Regulations, 2010 CFR

2010-01-01

... COMMISSION (CONTINUED) REACTOR SITE CRITERIA § 100.3 Definitions. As used in this part: Combined license... power facilities. Exclusion area means that area surrounding the reactor, in which the reactor licensee.... Activities unrelated to operation of the reactor may be permitted in an exclusion area under appropriate...

10 CFR 100.3 - Definitions.

Code of Federal Regulations, 2011 CFR

2011-01-01

... COMMISSION (CONTINUED) REACTOR SITE CRITERIA § 100.3 Definitions. As used in this part: Combined license... power facilities. Exclusion area means that area surrounding the reactor, in which the reactor licensee.... Activities unrelated to operation of the reactor may be permitted in an exclusion area under appropriate...

Evaluation of performance of select fusion experiments and projected reactors

NASA Technical Reports Server (NTRS)

Miley, G. H.

1978-01-01

The performance of NASA Lewis fusion experiments (SUMMA and Bumpy Torus) is compared with other experiments and that necessary for a power reactor. Key parameters cited are gain (fusion power/input power) and the time average fusion power, both of which may be more significant for real fusion reactors than the commonly used Lawson parameter. The NASA devices are over 10 orders of magnitude below the required powerplant values in both gain and time average power. The best experiments elsewhere are also as much as 4 to 5 orders of magnitude low. However, the NASA experiments compare favorably with other alternate approaches that have received less funding than the mainline experiments. The steady-state character and efficiency of plasma heating are strong advantages of the NASA approach. The problem, though, is to move ahead to experiments of sufficient size to advance in gain and average power parameters.

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.

Multi-Megawatt Space Nuclear Power Generation

DTIC Science & Technology

1993-06-28

electric generation, both for open- and closed-cycle opera- tion. These reactors use the particulate fuel of the type developed for HTGR reactors. What...commercial HTGR power reactors, the particles are held in place and directly cooled. Figure 2.7 shows the two types of fuel particles developed for...of MW(e), for pulsed energy devices. The FBR would use HTGR -type particle fuel , contained in a annular bed be- tween two porous frits. Helium would

Solar Power Satellites - A Review of the Space Transportation Options.

DTIC Science & Technology

1980-03-01

already exists with such systems, gained mainly through liquid-metal breeder reactor programmes. 0 For example, inlet temperatures of 970 C can be handled...alternatives exist. In addition, there would be extreme reluctance on the part of most governments to allow large C- reactors , producing gigawatts of power, to...antenna. The reactors employed are high-temperature gas- cooled breeders , which convert U238 into fissile plutonium. Each of the modules includes a

EXPERIMENTAL MOLTEN-SALT-FUELED 30-Mw POWER REACTOR

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

Alexander, L.G.; Kinyon, B.W.; Lackey, M.E.

1960-03-24

A preliminary design study was made of an experimental molten-salt- fueled power reactor. The reactor considered is a single-region homogeneous burner coupled with a Loeffler steam-generating cycle. Conceptual plant layouts, basic information on the major fuel circuit components, a process flowsheet, and the nuclear characteristics of the core are presented. The design plant electrical output is 10 Mw, and the total construction cost is estimated to be approximately ,000,000. (auth)

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

Sklenka, L.; Rataj, J.; Frybort, J.

Research reactors play an important role in providing key personnel of nuclear power plants a hands-on experience from operation and experiments at nuclear facilities. Training of NPP (Nuclear Power Plant) staff is usually deeply theoretical with an extensive utilisation of simulators and computer visualisation. But a direct sensing of the reactor response to various actions can only improve the personnel awareness of important aspects of reactor operation. Training Reactor VR-1 and its utilization for training of NPP operators and other professionals from Czech Republic and Slovakia is described. Typical experimental exercises and good practices in organization of a training programmore » are demonstrated. (authors)« less

10 CFR 51.52 - Environmental effects of transportation of fuel and waste-Table S-4.

Code of Federal Regulations, 2011 CFR

2011-01-01

... thermal power level not exceeding 3,800 megawatts; (2) The reactor fuel is in the form of sintered uranium... Nuclear Power Plants,” WASH-1238, December 1972, and Supp. 1 NUREG-75/038 April 1975. Both documents are...-water-cooled nuclear power reactor, and submitted after February 4, 1975, shall contain a statement...

10 CFR 51.52 - Environmental effects of transportation of fuel and waste-Table S-4.

Code of Federal Regulations, 2014 CFR

2014-01-01

... thermal power level not exceeding 3,800 megawatts; (2) The reactor fuel is in the form of sintered uranium... Nuclear Power Plants,” WASH-1238, December 1972, and Supp. 1 NUREG-75/038 April 1975. Both documents are...-water-cooled nuclear power reactor, and submitted after February 4, 1975, shall contain a statement...

10 CFR 51.52 - Environmental effects of transportation of fuel and waste-Table S-4.

Code of Federal Regulations, 2012 CFR

2012-01-01

... thermal power level not exceeding 3,800 megawatts; (2) The reactor fuel is in the form of sintered uranium... Nuclear Power Plants,” WASH-1238, December 1972, and Supp. 1 NUREG-75/038 April 1975. Both documents are...-water-cooled nuclear power reactor, and submitted after February 4, 1975, shall contain a statement...

10 CFR 51.52 - Environmental effects of transportation of fuel and waste-Table S-4.

Code of Federal Regulations, 2013 CFR

2013-01-01

... thermal power level not exceeding 3,800 megawatts; (2) The reactor fuel is in the form of sintered uranium... Nuclear Power Plants,” WASH-1238, December 1972, and Supp. 1 NUREG-75/038 April 1975. Both documents are...-water-cooled nuclear power reactor, and submitted after February 4, 1975, shall contain a statement...

Comparison of evolving photovoltaic and nuclear power systems for earth orbital applications

NASA Technical Reports Server (NTRS)

Rockey, D. E.; Jones, R. M.; Schulman, I.

1982-01-01

Photovoltaic and fission reactor orbital power systems are compared in terms of the end-to-end system power-to-mass ratios. Three PV systems are examined, i.e., a solid substrate with a cell array and a NiCd battery, a modified SEP array and an NiH2 battery, and a 62-micron Si cell array and a fuel cell. All arrays were modeled to be 13.5% efficient and to produce 25 kW dc. The SP-100 reactor consists of the heat source, radiation shield, heat pipes to transfer thermal energy from the reactor to thermoelectric elements, and a waste heat radiator. Consideration is given to system applications in orbits ranging from LEO to GEO, and to mission durations of 1, 5, and 10 yr. PV systems are concluded to be flight-proven, useful out of radiation belts, and best for low to moderate power levels. Limitations exist for operations where atmospheric drag may become a factor and due to the size of a large PV power supply. Space nuclear reactors will continue under development and uses at high power levels and in low altitude orbits are foreseen.

Nuclear power: the bargain we can't afford

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

Morgan, R.

1977-01-01

This is a handbook for citizens who wish to raise questions about the costs of atomic energy. It explains, step-by-step, why nuclear reactors have failed to produce low-cost electricity, and it tells citizens how they can use economic arguments to challenge nuclear expansion. Part One, The Costs of Nuclear Energy, contains 7 chapters--The Price of Power (electricity is big business); Mushrooming Capital Costs (nuclear construction costs are skyrocketing); Nuclear Lemons (reactors spend much of their time closed for repairs); The Faulty Fuel Cycle (turning uranium into electricity is not as simple as the utilities say); Hidden Costs (goverment subsidies obscuremore » the true costs of atomic energy); Ratepayer Roulette (nuclear problems translate into higher electric rates); and Alternatives to the Atom (coal-fired power and energy conservation can meet future energy needs more cheaply than nuclear energy). Part Two, Challenging Nuclear Power, contains 3 chapters--Regulators and Reactors (state utility commissions can eliminate the power companies' bias toward nuclear energy); Legislation, Licensing, and Lawsuits (nuclear critics can challenge reactor construction in numerous forums); and Winning the Battle (building an organization is a crucial step in fighting nuclear power). (MCW)« less

The effects of electric power industry restructuring on the safety of nuclear power plants in the United States

NASA Astrophysics Data System (ADS)

Butler, Thomas S.

Throughout the United States the electric utility industry is restructuring in response to federal legislation mandating deregulation. The electric utility industry has embarked upon an extraordinary experiment by restructuring in response to deregulation that has been advocated on the premise of improving economic efficiency by encouraging competition in as many sectors of the industry as possible. However, unlike the telephone, trucking, and airline industries, the potential effects of electric deregulation reach far beyond simple energy economics. This dissertation presents the potential safety risks involved with the deregulation of the electric power industry in the United States and abroad. The pressures of a competitive environment on utilities with nuclear power plants in their portfolio to lower operation and maintenance costs could squeeze them to resort to some risky cost-cutting measures. These include deferring maintenance, reducing training, downsizing staff, excessive reductions in refueling down time, and increasing the use of on-line maintenance. The results of this study indicate statistically significant differences at the .01 level between the safety of pressurized water reactor nuclear power plants and boiling water reactor nuclear power plants. Boiling water reactors exhibited significantly more problems than did pressurized water reactors.

New developments and prospects on COSI, the simulation software for fuel cycle analysis

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

Eschbach, R.; Meyer, M.; Coquelet-Pascal, C.

2013-07-01

COSI, software developed by the Nuclear Energy Direction of the CEA, is a code simulating a pool of nuclear power plants with its associated fuel cycle facilities. This code has been designed to study various short, medium and long term options for the introduction of various types of nuclear reactors and for the use of associated nuclear materials. In the frame of the French Act for waste management, scenario studies are carried out with COSI, to compare different options of evolution of the French reactor fleet and options of partitioning and transmutation of plutonium and minor actinides. Those studies aimmore » in particular at evaluating the sustainability of Sodium cooled Fast Reactors (SFR) deployment and the possibility to transmute minor actinides. The COSI6 version is a completely renewed software released in 2006. COSI6 is now coupled with the last version of CESAR (CESAR5.3 based on JEFF3.1.1 nuclear data) allowing the calculations on irradiated fuel with 200 fission products and 100 heavy nuclides. A new release is planned in 2013, including in particular the coupling with a recommended database of reactors. An exercise of validation of COSI6, carried out on the French PWR historic nuclear fleet, has been performed. During this exercise quantities like cumulative natural uranium consumption, or cumulative depleted uranium, or UOX/MOX spent fuel storage, or stocks of reprocessed uranium, or plutonium content in fresh MOX fuel, or the annual production of high level waste, have been computed by COSI6 and compared to industrial data. The results have allowed us to validate the essential phases of the fuel cycle computation, and reinforces the credibility of the results provided by the code.« less

Gas-core reactor power transient analysis

NASA Technical Reports Server (NTRS)

Kascak, A. F.

1972-01-01

The gas core reactor is a proposed device which features high temperatures. It has applications in high specific impulse space missions, and possibly in low thermal pollution MHD power plants. The nuclear fuel is a ball of uranium plasma radiating thermal photons as opposed to gamma rays. This thermal energy is picked up before it reaches the solid cavity liner by an inflowing seeded propellant stream and convected out through a rocket nozzle. A wall-burnout condition will exist if there is not enough flow of propellant to convect the energy back into the cavity. A reactor must therefore operate with a certain amount of excess propellant flow. Due to the thermal inertia of the flowing propellant, the reactor can undergo power transients in excess of the steady-state wall burnout power for short periods of time. The objective of this study was to determine how long the wall burnout power could be exceeded without burning out the cavity liner. The model used in the heat-transfer calculation was one-dimensional, and thermal radiation was assumed to be a diffusion process.

Nuclear power in the 21st century: Challenges and possibilities.

PubMed

Horvath, Akos; Rachlew, Elisabeth

2016-01-01

The current situation and possible future developments for nuclear power--including fission and fusion processes--is presented. The fission nuclear power continues to be an essential part of the low-carbon electricity generation in the world for decades to come. There are breakthrough possibilities in the development of new generation nuclear reactors where the life-time of the nuclear waste can be reduced to some hundreds of years instead of the present time-scales of hundred thousand of years. Research on the fourth generation reactors is needed for the realisation of this development. For the fast nuclear reactors, a substantial research and development effort is required in many fields--from material sciences to safety demonstration--to attain the envisaged goals. Fusion provides a long-term vision for an efficient energy production. The fusion option for a nuclear reactor for efficient production of electricity has been set out in a focussed European programme including the international project of ITER after which a fusion electricity DEMO reactor is envisaged.

Space rocket engine on the base of the reactor-pumped laser for the interplanetary flights and earth orbital applications

NASA Astrophysics Data System (ADS)

Gulevich, Andrey V.; Dyachenko, Peter P.; Kukharchuk, Oleg F.; Zrodnikov, Anatoly V.

2000-01-01

In this report the concept of vehicle-based reactor-laser engine for long time interplanetary and interorbital (LEO to GEO) flights is proposed. Reactor-pumped lasers offer the perspective way to create on the base of modern nuclear and lasers technologies the low mass and high energy density, repetitively pulsed vehicle-based laser of average power 100 kW. Nowadays the efficiency of nuclear-to-optical energy conversion reached the value of 2-3%. The demo model of reactor-pumped laser facility is under construction in Institute for Physics and Power Engineering (Obninsk, Russia). It enable us to hope that using high power laser on board of the vehicle could make the effective space laser engine possible. Such engine may provide the high specific impulse ~1000-2000 s with the thrust up to 10-100 n. Some calculation results of the characteristics of vehicle-based reactor-laser thermal engine concept are also presented. .

76 FR 79229 - Advisory Committee on Reactor Safeguards; Notice of Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2011-12-21

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards; Notice of Meeting In... Advisory Committee on Reactor Safeguards (ACRS) will hold a meeting on January 19-20, 2012, 11545 Rockville... Cooling Systems for Light- Water Nuclear Power Reactors'' (Open)--The Committee will hear presentations by...

10 CFR 50.36 - Technical specifications.

Code of Federal Regulations, 2013 CFR

2013-01-01

... nuclear reactors are limits upon important process variables that are found to be necessary to reasonably... Commission terminates the license for the reactor, except for nuclear power reactors licensed under § 50.21(b... for nuclear reactors are settings for automatic protective devices related to those variables having...

10 CFR 50.36 - Technical specifications.

Code of Federal Regulations, 2014 CFR

2014-01-01

... nuclear reactors are limits upon important process variables that are found to be necessary to reasonably... Commission terminates the license for the reactor, except for nuclear power reactors licensed under § 50.21(b... for nuclear reactors are settings for automatic protective devices related to those variables having...

10 CFR 50.36 - Technical specifications.

Code of Federal Regulations, 2012 CFR

2012-01-01

... nuclear reactors are limits upon important process variables that are found to be necessary to reasonably... Commission terminates the license for the reactor, except for nuclear power reactors licensed under § 50.21(b... for nuclear reactors are settings for automatic protective devices related to those variables having...

63. REACTOR CHAMBER (BASF) FROM NORTH SHOWING NEUTRON SHIELD TANK ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

63. REACTOR CHAMBER (BASF) FROM NORTH SHOWING NEUTRON SHIELD TANK AND REACTOR PIPING (LOCATION RRR) - Shippingport Atomic Power Station, On Ohio River, 25 miles Northwest of Pittsburgh, Shippingport, Beaver County, PA

Comparative evaluation of solar, fission, fusion, and fossil energy resources, part 3

NASA Technical Reports Server (NTRS)

Clement, J. D.; Reupke, W. A.

1974-01-01

The role of nuclear fission reactors in becoming an important power source in the world is discussed. The supply of fissile nuclear fuel will be severely depleted by the year 2000. With breeder reactors the world supply of uranium could last thousands of years. However, breeder reactors have problems of a large radioactive inventory and an accident potential which could present an unacceptable hazard. Although breeder reactors afford a possible solution to the energy shortage, their ultimate role will depend on demonstrated safety and acceptable risks and environmental effects. Fusion power would also be a long range, essentially permanent, solution to the world's energy problem. Fusion appears to compare favorably with breeders in safety and environmental effects. Research comparing a controlled fusion reactor with the breeder reactor in solving our long range energy needs is discussed.

Control console replacement at the WPI Reactor

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

Not Available

1992-01-01

With partial funding from the Department of Energy (DOE) University Reactor Instrumentation Upgrade Program (DOE Grant No. DE-FG02-90ER12982), the original control console at the Worcester Polytechnic Institute (WPI) Reactor has been replaced with a modern system. The new console maintains the original design bases and functionality while utilizing current technology. An advanced remote monitoring system has been added to augment the educational capabilities of the reactor. Designed and built by General Electric in 1959, the open pool nuclear training reactor at WPI was one of the first such facilities in the nation located on a university campus. Devoted to undergraduatemore » use, the reactor and its related facilities have been since used to train two generations of nuclear engineers and scientists for the nuclear industry. The reactor power level was upgraded from 1 to 10 kill in 1969, and its operating license was renewed for 20 years in 1983. In 1988, the reactor was converted to low enriched uranium. The low power output of the reactor and ergonomic facility design make it an ideal tool for undergraduate nuclear engineering education and other training.« 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.

Power flattening on modified CANDLE small long life gas-cooled fast reactor

NASA Astrophysics Data System (ADS)

Monado, Fiber; Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Ariani, Menik; Sekimoto, Hiroshi

2014-09-01

Gas-cooled Fast Reactor (GFR) is one of the candidates of next generation Nuclear Power Plants (NPPs) that expected to be operated commercially after 2030. In this research conceptual design study of long life 350 MWt GFR with natural uranium metallic fuel as fuel cycle input has been performed. Modified CANDLE burn-up strategy with first and second regions located near the last region (type B) has been applied. This reactor can be operated for 10 years without refuelling and fuel shuffling. Power peaking reduction is conducted by arranging the core radial direction into three regions with respectively uses fuel volume fraction 62.5%, 64% and 67.5%. The average power density in the modified core is about 82 Watt/cc and the power peaking factor decreased from 4.03 to 3.43.

10 CFR 52.131 - Scope of subpart.

Code of Federal Regulations, 2010 CFR

2010-01-01

... REGULATORY COMMISSION (CONTINUED) LICENSES, CERTIFICATIONS, AND APPROVALS FOR NUCLEAR POWER PLANTS Standard... review, and referral to the Advisory Committee on Reactor Safeguards of standard designs for a nuclear power reactor of the type described in § 50.22 of this chapter or major portions thereof. ...

Design of a tokamak fusion reactor first wall armor against neutral beam impingement

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

Myers, R.A.

1977-12-01

The maximum temperatures and thermal stresses are calculated for various first wall design proposals, using both analytical solutions and the TRUMP and SAP IV Computer Codes. Beam parameters, such as pulse time, cycle time, and beam power, are varied. It is found that uncooled plates should be adequate for near-term devices, while cooled protection will be necessary for fusion power reactors. Graphite and tungsten are selected for analysis because of their desirable characteristics. Graphite allows for higher heat fluxes compared to tungsten for similar pulse times. Anticipated erosion (due to surface effects) and plasma impurity fraction are estimated. Neutron irradiationmore » damage is also discussed. Neutron irradiation damage (rather than erosion, fatigue, or creep) is estimated to be the lifetime-limiting factor on the lifetime of the component in fusion power reactors. It is found that the use of tungsten in fusion power reactors, when directly exposed to the plasma, will cause serious plasma impurity problems; graphite should not present such an impurity problem.« less

Split-core heat-pipe reactors for out-of-pile thermionic power systems.

NASA Technical Reports Server (NTRS)

Niederauer, G.; Lantz, E.; Breitweiser, R.

1971-01-01

Description of the concept of splitting a heat-pipe reactor for out-of-core thermionics into two identical halves and using the resulting center gap for reactivity control. Short Li-W reactor heat pipes penetrate the axial reflectors and form a heat exchanger with long heat pipes which wind through the shield to the thermionic diodes. With one reactor half anchored to the shield, the other is attached to a long arm with a pivot behind the shield and swings through a small arc for reactivity control. A safety shim prevents large reactivity inputs, and a fueled control arm drive shaft acts as a power stabilizer. Reactors fueled with U-235C and with U-233C have been studied.-

The Economics of IRIS

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

Miller, K.; Paramonov, D.

2002-07-01

IRIS (International Reactor Innovative and Secure) is a small to medium advanced light water cooled modular reactor being developed by an international consortium led by Westinghouse/BNFL. This reactor design is specifically aimed at utilities looking to install new (or replacement) nuclear capacity to match market demands, or at developing countries for their distributed power needs. To determine the optimal configuration for IRIS, analysis was undertaken to establish Generation Costs ($/MWh) and Internal Rate of Return (IRR %) to the Utility at alternative power ratings. This was then combined with global market projections for electricity demand out to 2030, segmented intomore » key geographical regions. Finally this information is brought together to form insights, conclusions and recommendations regarding the optimal design. The resultant analysis reveals a single module sized at 335 MWe, with a construction period of 3 years and a 60-year plant life. Individual modules can be installed in a staggered fashion (3 equivalent to 1005 MWe) or built in pairs (2 sets of twin units' equivalent to 1340 MWe). Uncertainty in Market Clearing Price for electricity, Annual Operating Costs and Construction Costs primarily influence lifetime Net Present Values (NPV) and hence IRR % for Utilities. Generation Costs in addition are also influenced by Fuel Costs, Plant Output, Plant Availability and Plant Capacity Factor. Therefore for a site based on 3 single modules, located in North America, Generations Costs of 28.5 $/MWh are required to achieve an IRR of 20%, a level which enables IRIS to compete with all other forms of electricity production. Plant size is critical to commercial success. Sustained (lifetime) high factors for Plant Output, Availability and Capacity Factor are required to achieve a competitive advantage. Modularity offers Utilities the option to match their investments with market conditions, adding additional capacity as and when the circumstances are right. Construction schedule needs to be controlled. There is a clear trade-off between reducing financing charges and optimising revenue streams. (authors)« less

Calculated power distribution of a thermionic, beryllium oxide reflected, fast-spectrum reactor

NASA Technical Reports Server (NTRS)

Mayo, W.; Lantz, E.

1973-01-01

A procedure is developed and used to calculate the detailed power distribution in the fuel elements next to a beryllium oxide reflector of a fast-spectrum, thermionic reactor. The results of the calculations show that, although the average power density in these outer fuel elements is not far from the core average, the power density at the very edge of the fuel closest to the beryllium oxide is about 1.8 times the core avearge.

Power-Conversion Concept Designed for the Jupiter Icy Moons Orbiter

NASA Technical Reports Server (NTRS)

Mason, Lee S.

2004-01-01

The Jupiter Icy Moons Orbiter (JIMO) is a bold new mission being developed by NASA's Office of Space Science under Project Prometheus. JIMO is examining the potential of nuclear electric propulsion (NEP) technology to efficiently deliver scientific payloads to three of Jupiter's moons: Callisto, Ganymede, and Europa. A critical element of the NEP spacecraft is the space reactor power system (SRPS), consisting of the nuclear reactor, power conversion, heat rejection, and power management and distribution (PMAD).

PBF Cooling Tower. View from highbay roof of Reactor Building ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

PBF Cooling Tower. View from high-bay roof of Reactor Building (PER-620). Camera faces northwest. East louvered face has been installed. Inlet pipes protrude from fan deck. Two redwood vents under construction at top. Note piping, control, and power lines at sub-grade level in trench leading to Reactor Building. Photographer: Kirsh. Date: June 6, 1969. INEEL negative no. 69-3466 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

PREMOR: a point reactor exposure model computer code for survey analysis of power plant performance

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

Vondy, D.R.

1979-10-01

The PREMOR computer code was written to exploit a simple, two-group point nuclear reactor power plant model for survey analysis. Up to thirteen actinides, fourteen fission products, and one lumped absorber nuclide density are followed over a reactor history. Successive feed batches are accounted for with provision for from one to twenty batches resident. The effect of exposure of each of the batches to the same neutron flux is determined.

Navy Nuclear-Powered Surface Ships: Background, Issues, and Options for Congress

DTIC Science & Technology

2010-03-29

246 of H.R. 2647 would require DOD to submit to the congressional defense committees a study on the use of thorium -liquid fueled nuclear reactors ...Congressional Research Service 19 SEC. 246. STUDY ON THORIUM -LIQUID FUELED REACTORS FOR NAVAL FORCES. (a) Study Required- The Secretary of Defense and...the Chairman of the Joint Chiefs of Staff shall jointly carry out a study on the use of thorium -liquid fueled nuclear reactors for naval power

Navy Nuclear-Powered Surface Ships: Background, Issues, and Options for Congress

DTIC Science & Technology

2009-12-23

congressional defense committees a study on the use of thorium -liquid fueled nuclear reactors for Navy surface ships. The text of Section 246 is as follows...carry out a study on the use of thorium -liquid fueled nuclear reactors for naval power needs pursuant to section 1012, of the National Defense...force— (1) compare and contrast thorium -liquid fueled reactor concept to the 2005 Quick Look, 2006 Navy Alternative Propulsion Study, and the navy CG

A Comparison of Fission Power System Options for Lunar and Mars Surface Applications

NASA Technical Reports Server (NTRS)

Mason, Lee S.

2006-01-01

This paper presents a comparison of reactor and power conversion design options for 50 kWe class lunar and Mars surface power applications with scaling from 25 to 200 kWe. Design concepts and integration approaches are provided for three reactor-converter combinations: gas-cooled Brayton, liquid-metal Stirling, and liquid-metal thermoelectric. The study examines the mass and performance of low temperature, stainless steel based reactors and higher temperature refractory reactors. The preferred system implementation approach uses crew-assisted assembly and in-situ radiation shielding via installation of the reactor in an excavated hole. As an alternative, self-deployable system concepts that use earth-delivered, on-board radiation shielding are evaluated. The analyses indicate that among the 50 kWe stainless steel reactor options, the liquid-metal Stirling system provides the lowest mass at about 5300 kg followed by the gas-cooled Brayton at 5700 kg and the liquid-metal thermoelectric at 8400 kg. The use of a higher temperature, refractory reactor favors the gas-cooled Brayton option with a system mass of about 4200 kg as compared to the Stirling and thermoelectric options at 4700 and 5600 kg, respectively. The self-deployed concepts with on-board shielding result in a factor of two system mass increase as compared to the in-situ shielded concepts.

Evaluation of power density on the bioethanol production using mesoscale oscillatory baffled reactor and stirred tank reactor

NASA Astrophysics Data System (ADS)

Yussof, H. W.; Bahri, S. S.; Mazlan, N. A.

2018-03-01

A recent development in oscillatory baffled reactor technology is down-scaling the reactor, so that it can be used for production of small-scale bioproduct. In the present study, a mesoscale oscillatory baffled reactor (MOBR) with central baffle system was developed. The reactor performance of the MOBR was compared with conventional stirred tank reactor (STR) to evaluate the performance of bioethanol fermentation using Saccharomyces cerevisiae. Evaluation was made at similar power density of 24.21, 57.38, 112.35 and 193.67 Wm-3 by varying frequency (f), amplitude (xo) and agitation speed (rpm). It was found that the MOBR improved the mixing intensity resulted in lower glucose concentration (0.988 gL-1) and higher bioethanol concentration (38.98 gL-1) after 12 hours fermentation at power density of 193.67 Wm-3. Based on the results, the bioethanol yield obtained using MOBR was 39% higher than the maximum achieved in STR. Bioethanol production using MOBR proved to be feasible as it is not only able to compete with conventional STR but also offers advantages of straight-forward scale-up, whereas it is complicated and difficult in STR. Overall, MOBR offers great prospective over the conventional STR.

Delayed Neutrons Effect on Power Reactor with Variation of Fluid Fuel Velocity at MSR Fuji-12

NASA Astrophysics Data System (ADS)

Kuncoro Aji, Indarta; Pramuditya, Syeilendra; Novitrian; Irwanto, Dwi; Waris, Abdul

2017-01-01

As the nuclear reactor operate with liquid fuel, controlling velocity of the fuel flow on the Molten salt reactor very influence on the neutron kinetics in that reactor system. The effect of the pace fuel changes to the populations number of neutrons and power density on vertical direction (1 dimension) from the first until fifth year reactor operating had been analyzed on this research. This research had been conducted on MSR Fuji-12 with a two meters core high, and LiF-BeF2-ThF4-233UF4 as fuel composition respectively 71.78%-16%-11.86%-0.36%. Data of reactivity, neutron flux, and the macroscopic fission cross section obtained from ouput of SRAC (neutronic calculation code has been developed by JAEA, with JENDL-4.0 as data library on the SRAC calculation) was being used for the calculation process of this research. The calculation process of this research had been performed numerically by SOR (successive over relaxation) and finite difference methode, as well as using C programing language. From the calculation, regarding to the value of power density resulting from delayed neutrons, concluded that 20 m/s is the optimum fuel flow velocity in all the years reactor had operated. Where the increases number of power are inversely proportional with the fuel flow speed.

Using the sound of nuclear energy

DOE PAGES

Garrett, Steven; Smith, James; Smith, Robert; ...

2016-08-01

The generation of sound by heat has been documented as an “acoustical curiosity” since a Buddhist monk reported the loud tone generated by a ceremonial rice-cooker in his diary, in 1568. Over the last four decades, significant progress has been made in understanding “thermoacoustic processes,” enabling the design of thermoacoustic engines and refrigerators. Motivated by the Fukushima nuclear reactor disaster, we have developed and tested a thermoacoustic engine that exploits the energy-rich conditions in the core of a nuclear reactor to provide core condition information to the operators without a need for external electrical power. The heat engine is self-poweredmore » and can wirelessly transmit the temperature and reactor power level by generation of a pure tone which can be detected outside the reactor. We report here the first use of a fission-powered thermoacoustic engine capable of serving as a performance and safety sensor in the core of a research reactor and present data from the hydrophones in the coolant (far from the core) and an accelerometer attached to a structure outside the reactor. These measurements confirmed that the frequency of the sound produced indicates the reactor’s coolant temperature and that the amplitude (above an onset threshold) is related to the reactor’s operating power level. Furthermore, these signals can be detected even in the presence of substantial background noise generated by the reactor’s fluid pumps.« less

Using the sound of nuclear energy

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

Garrett, Steven; Smith, James; Smith, Robert

The generation of sound by heat has been documented as an “acoustical curiosity” since a Buddhist monk reported the loud tone generated by a ceremonial rice-cooker in his diary, in 1568. Over the last four decades, significant progress has been made in understanding “thermoacoustic processes,” enabling the design of thermoacoustic engines and refrigerators. Motivated by the Fukushima nuclear reactor disaster, we have developed and tested a thermoacoustic engine that exploits the energy-rich conditions in the core of a nuclear reactor to provide core condition information to the operators without a need for external electrical power. The heat engine is self-poweredmore » and can wirelessly transmit the temperature and reactor power level by generation of a pure tone which can be detected outside the reactor. We report here the first use of a fission-powered thermoacoustic engine capable of serving as a performance and safety sensor in the core of a research reactor and present data from the hydrophones in the coolant (far from the core) and an accelerometer attached to a structure outside the reactor. These measurements confirmed that the frequency of the sound produced indicates the reactor’s coolant temperature and that the amplitude (above an onset threshold) is related to the reactor’s operating power level. Furthermore, these signals can be detected even in the presence of substantial background noise generated by the reactor’s fluid pumps.« less

Conceptual design study of the moderate size superconducting spherical tokamak power plant

NASA Astrophysics Data System (ADS)

Gi, Keii; Ono, Yasushi; Nakamura, Makoto; Someya, Youji; Utoh, Hiroyasu; Tobita, Kenji; Ono, Masayuki

2015-06-01

A new conceptual design of the superconducting spherical tokamak (ST) power plant was proposed as an attractive choice for tokamak fusion reactors. We reassessed a possibility of the ST as a power plant using the conservative reactor engineering constraints often used for the conventional tokamak reactor design. An extensive parameters scan which covers all ranges of feasible superconducting ST reactors was completed, and five constraints which include already achieved plasma magnetohydrodynamic (MHD) and confinement parameters in ST experiments were established for the purpose of choosing the optimum operation point. Based on comparison with the estimated future energy costs of electricity (COEs) in Japan, cost-effective ST reactors can be designed if their COEs are smaller than 120 mills kW-1 h-1 (2013). We selected the optimized design point: A = 2.0 and Rp = 5.4 m after considering the maintenance scheme and TF ripple. A self-consistent free-boundary MHD equilibrium and poloidal field coil configuration of the ST reactor were designed by modifying the neutral beam injection system and plasma profiles. The MHD stability of the equilibrium was analysed and a ramp-up scenario was considered for ensuring the new ST design. The optimized moderate-size ST power plant conceptual design realizes realistic plasma and fusion engineering parameters keeping its economic competitiveness against existing energy sources in Japan.

The Satellite Nuclear Power Station - An option for future power generation.

NASA Technical Reports Server (NTRS)

Williams, J. R.; Clement, J. D.

1973-01-01

A new concept in nuclear power generation is being explored which essentially eliminates major objections to nuclear power. The Satellite Nuclear Power Station, remotely operated in synchronous orbit, would transmit power safely to the ground by a microwave beam. Fuel reprocessing would take place in space and no radioactive materials would ever be returned to earth. Even the worst possible accident to such a plant should have negligible effect on the earth. An exploratory study of a satellite nuclear power station to provide 10,000 MWe to the earth has shown that the system could weigh about 20 million pounds and cost less than $1000/KWe. An advanced breeder reactor operating with an MHD power cycle could achieve an efficiency of about 50% with a 1100 K radiator temperature. If a hydrogen moderated gas core reactor is used, its breeding ratio of 1.10 would result in a fuel doubling time of a few years. A rotating fluidized bed or NERVA type reactor might also be used. The efficiency of power transmission from synchronous orbit would range from 70% to 80%.

Gamma thermometer based reactor core liquid level detector

DOEpatents

Burns, Thomas J.

1983-01-01

A system is provided which employs a modified gamma thermometer for determining the liquid coolant level within a nuclear reactor core. The gamma thermometer which normally is employed to monitor local core heat generation rate (reactor power), is modified by thermocouple junctions and leads to obtain an unambiguous indication of the presence or absence of coolant liquid at the gamma thermometer location. A signal processor generates a signal based on the thermometer surface heat transfer coefficient by comparing the signals from the thermocouples at the thermometer location. The generated signal is a direct indication of loss of coolant due to the change in surface heat transfer when coolant liquid drops below the thermometer location. The loss of coolant indication is independent of reactor power at the thermometer location. Further, the same thermometer may still be used for the normal power monitoring function.

10 CFR 100.10 - Factors to be considered when evaluating sites.

Code of Federal Regulations, 2011 CFR

2011-01-01

... Section 100.10 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) REACTOR SITE CRITERIA Evaluation Factors for Stationary Power Reactor Site Applications Before January 10, 1997 and for Testing Reactors § 100... include those relating both to the proposed reactor design and the characteristics peculiar to the site...

10 CFR 100.10 - Factors to be considered when evaluating sites.

Code of Federal Regulations, 2010 CFR

2010-01-01

... Section 100.10 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) REACTOR SITE CRITERIA Evaluation Factors for Stationary Power Reactor Site Applications Before January 10, 1997 and for Testing Reactors § 100... include those relating both to the proposed reactor design and the characteristics peculiar to the site...

10 CFR 52.1 - Definitions.

Code of Federal Regulations, 2014 CFR

2014-01-01

... authorization means the authorization provided by the Director of New Reactors or the Director of Nuclear... identical nuclear reactors (modules) and each module is a separate nuclear reactor capable of being operated... nuclear power reactor of the type described in 10 CFR 50.22. The approval may be for either the final...

10 CFR 52.1 - Definitions.

Code of Federal Regulations, 2011 CFR

2011-01-01

... authorization means the authorization provided by the Director of New Reactors or the Director of Nuclear... identical nuclear reactors (modules) and each module is a separate nuclear reactor capable of being operated... nuclear power reactor of the type described in 10 CFR 50.22. The approval may be for either the final...

10 CFR 52.1 - Definitions.

Code of Federal Regulations, 2013 CFR

2013-01-01

... authorization means the authorization provided by the Director of New Reactors or the Director of Nuclear... identical nuclear reactors (modules) and each module is a separate nuclear reactor capable of being operated... nuclear power reactor of the type described in 10 CFR 50.22. The approval may be for either the final...

10 CFR 52.1 - Definitions.

Code of Federal Regulations, 2012 CFR

2012-01-01

... authorization means the authorization provided by the Director of New Reactors or the Director of Nuclear... identical nuclear reactors (modules) and each module is a separate nuclear reactor capable of being operated... nuclear power reactor of the type described in 10 CFR 50.22. The approval may be for either the final...

10 CFR Appendix I to Part 50 - Numerical Guides for Design Objectives and Limiting Conditions for Operation To Meet the...

Code of Federal Regulations, 2013 CFR

2013-01-01

... in Light-Water-Cooled Nuclear Power Reactor Effluents I Appendix I to Part 50 Energy NUCLEAR... Criterion “As Low as is Reasonably Achievable” for Radioactive Material in Light-Water-Cooled Nuclear Power... light-water-cooled nuclear power reactors licensed under 10 CFR part 50 or part 52 of this chapter. The...

10 CFR Appendix I to Part 50 - Numerical Guides for Design Objectives and Limiting Conditions for Operation To Meet the...

Code of Federal Regulations, 2012 CFR

2012-01-01

... in Light-Water-Cooled Nuclear Power Reactor Effluents I Appendix I to Part 50 Energy NUCLEAR... Criterion “As Low as is Reasonably Achievable” for Radioactive Material in Light-Water-Cooled Nuclear Power... light-water-cooled nuclear power reactors licensed under 10 CFR part 50 or part 52 of this chapter. The...

10 CFR Appendix I to Part 50 - Numerical Guides for Design Objectives and Limiting Conditions for Operation To Meet the...

Code of Federal Regulations, 2014 CFR

2014-01-01

... in Light-Water-Cooled Nuclear Power Reactor Effluents I Appendix I to Part 50 Energy NUCLEAR... Criterion “As Low as is Reasonably Achievable” for Radioactive Material in Light-Water-Cooled Nuclear Power... light-water-cooled nuclear power reactors licensed under 10 CFR part 50 or part 52 of this chapter. The...

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.

Decommissioning of German Research Reactors Under the Governance of the Federal Ministry of Education and Research - 12154

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

Weigl, M.

2012-07-01

Since 1956, nuclear research and development (R and D) in Germany has been supported by the Federal Government. The goal was to help German industry to become competitive in all fields of nuclear technology. National research centers were established and demonstration plants were built. In the meantime, all these facilities were shut down and are now in a state of decommissioning and dismantling (D and D). Meanwhile, Germany is one of the leading countries in the world in the field of D and D. Two big demonstration plants, the Niederaichbach Nuclear Power Plant (KKN) a heavy-water cooled pressure tube reactormore » with carbon-dioxide cooling and the Karlstein Superheated Steam Reactor (HDR) a boiling light water reactor with a thermal power of 100 MW, are totally dismantled and 'green field' is reached. Another big project was finished in 2008. The Forschungs-Reaktor Juelich 1 (FRJ1), a research reactor with a thermal power of 10 MW was completely dismantled and in September 2008 an oak tree was planted on a green field at the site, where the FRJ1 was standing before. This is another example for German success in the field of D and D. Within these projects a lot of new solutions and innovative techniques were tested, which were developed at German universities and in small and medium sized companies mostly funded by the Federal Ministry of Education and Research (BMBF). Some examples are underwater-cutting technologies like plasma arc cutting and contact arc metal cutting. This clearly shows that research on the field of D and D is important for the future. Moreover, these research activities are important to save the know-how in nuclear engineering in Germany and will enable enterprises to compete on the increasing market of D and D services. The author assumes that an efficient decommissioning of nuclear installations will help stabilize the credibility of nuclear energy. Some critics of nuclear energy are insisting that a return to 'green field sites' is not possible. The successful completion of two big D and D projects (HDR and KKN), which reached green field conditions, are showing quite the contrary. Moreover, research on D and D technologies offers the possibility to educate students on a field of nuclear technology, which will be very important in the future. In these days D and D companies are seeking for a lot of young engineers and this will not change in the coming years. (authors)« less

10 CFR 2.1105 - Definitions.

Code of Federal Regulations, 2011 CFR

2011-01-01

... Hearing Procedures for Expansion of Spent Nuclear Fuel Storage Capacity at Civilian Nuclear Power Reactors § 2.1105 Definitions. As used in this part: (a) Civilian nuclear power reactor means a civilian... 10 Energy 1 2011-01-01 2011-01-01 false Definitions. 2.1105 Section 2.1105 Energy NUCLEAR...

10 CFR 2.1105 - Definitions.

Code of Federal Regulations, 2012 CFR

2012-01-01

... Hearing Procedures for Expansion of Spent Nuclear Fuel Storage Capacity at Civilian Nuclear Power Reactors § 2.1105 Definitions. As used in this part: (a) Civilian nuclear power reactor means a civilian... 10 Energy 1 2012-01-01 2012-01-01 false Definitions. 2.1105 Section 2.1105 Energy NUCLEAR...

Annular gel reactor for chemical pattern formation

DOEpatents

Nosticzius, Zoltan; Horsthemke, Werner; McCormick, William D.; Swinney, Harry L.; Tam, Wing Y.

1990-01-01

The present invention is directed to an annular gel reactor suitable for the production and observation of spatiotemporal patterns created during a chemical reaction. The apparatus comprises a vessel having at least a first and second chamber separated one from the other by an annular polymer gel layer (or other fine porous medium) which is inert to the materials to be reacted but capable of allowing diffusion of the chemicals into it.

Investigation of the effects of radiolytic-gas bubbles on the long-term operation of solution reactors for medical-isotope production

NASA Astrophysics Data System (ADS)

Souto Mantecon, Francisco Javier

One of the most common and important medical radioisotopes is 99Mo, which is currently produced using the target irradiation technology in heterogeneous nuclear reactors. The medical isotope 99Mo can also be produced from uranium fission using aqueous homogeneous solution reactors. In solution reactors, 99Mo is generated directly in the fuel solution, resulting in potential advantages when compared with the target irradiation process in heterogeneous reactors, such as lower reactor power, less waste heat, and reduction by a factor of about 100 in the generation of spent fuel. The commercial production of medical isotopes in solution reactors requires steady-state operation at about 200 kW. At this power regime, the formation of radiolytic-gas bubbles creates a void volume in the fuel solution that introduces a negative coefficient of reactivity, resulting in power reduction and instabilities that may impede reactor operation for medical-isotope production. A model has been developed considering that reactivity effects are due to the increase in the fuel-solution temperature and the formation of radiolytic-gas bubbles. The model has been validated against experimental results from the Los Alamos National Laboratory uranyl fluoride Solution High-Energy Burst Assembly (SHEBA), and the SILENE uranyl nitrate solution reactor, commissioned at the Commissariat a l'Energie Atomique, in Valduc, France. The model shows the feasibility of solution reactors for the commercial production of medical isotopes and reveals some of the important parameters to consider in their design, including the fuel-solution type, 235U enrichment, uranium concentration, reactor vessel geometry, and neutron reflectors surrounding the reactor vessel. The work presented herein indicates that steady-state operation at 200 kW can be achieved with a solution reactor consisting of 120 L of uranyl nitrate solution enriched up to 20% with 235U and a uranium concentration of 145 kg/m3 in a graphite-reflected cylindrical geometry.

Closed Brayton cycle power conversion systems for nuclear reactors :

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

Wright, Steven A.; Lipinski, Ronald J.; Vernon, Milton E.

2006-04-01

This report describes the results of a Sandia National Laboratories internally funded research program to study the coupling of nuclear reactors to gas dynamic Brayton power conversion systems. The research focused on developing integrated dynamic system models, fabricating a 10-30 kWe closed loop Brayton cycle, and validating these models by operating the Brayton test-loop. The work tasks were performed in three major areas. First, the system equations and dynamic models for reactors and Closed Brayton Cycle (CBC) systems were developed and implemented in SIMULINKTM. Within this effort, both steady state and dynamic system models for all the components (turbines, compressors,more » reactors, ducting, alternators, heat exchangers, and space based radiators) were developed and assembled into complete systems for gas cooled reactors, liquid metal reactors, and electrically heated simulators. Various control modules that use proportional-integral-differential (PID) feedback loops for the reactor and the power-conversion shaft speed were also developed and implemented. The simulation code is called RPCSIM (Reactor Power and Control Simulator). In the second task an open cycle commercially available Capstone C30 micro-turbine power generator was modified to provide a small inexpensive closed Brayton cycle test loop called the Sandia Brayton test-Loop (SBL-30). The Capstone gas-turbine unit housing was modified to permit the attachment of an electrical heater and a water cooled chiller to form a closed loop. The Capstone turbine, compressor, and alternator were used without modification. The Capstone systems nominal operating point is 1150 K turbine inlet temperature at 96,000 rpm. The annular recuperator and portions of the Capstone control system (inverter) and starter system also were reused. The rotational speed of the turbo-machinery is controlled by adjusting the alternator load by using the electrical grid as the load bank. The SBL-30 test loop was operated at the manufacturers site (Barber-Nichols Inc.) and installed and operated at Sandia. A sufficiently detailed description of the loop is provided in this report along with the design characteristics of the turbo-alternator-compressor set to allow other researchers to compare their results with those measured in the Sandia test-loop. The third task consisted of a validation effort. In this task the test loop was operated and compared with the modeled results to develop a more complete understanding of this electrically heated closed power generation system and to validate the model. The measured and predicted system temperatures and pressures are in good agreement, indicating that the model is a reasonable representation of the test loop. Typical deviations between the model and the hardware results are less than 10%. Additional tests were performed to assess the capability of the Brayton engine to continue to remove decay heat after the reactor/heater is shutdown, to develop safe and effective control strategies, and to access the effectiveness of gas inventory control as an alternative means to provide load following. In one test the heater power was turned off to simulate a rapid reactor shutdown, and the turbomachinery was driven solely by the sensible heat stored in the heater for over 71 minutes without external power input. This is an important safety feature for CBC systems as it means that the closed Brayton loop will keep cooling the reactor without the need for auxiliary power (other than that needed to circulate the waste heat rejection coolant) provided the heat sink is available.« less

Summary of aerospace and nuclear engineering activities

NASA Technical Reports Server (NTRS)

1988-01-01

The Texas A&M Nuclear and Aerospace engineering departments have worked on five different projects for the NASA/USRA Advanced Design Program during the 1987/88 year. The aerospace department worked on two types of lunar tunnelers that would create habitable space. The first design used a heated cone to melt the lunar regolith, and the second used a conventional drill to bore its way through the crust. Both used a dump truck to get rid of waste heat from the reactor as well as excess regolith from the tunneling operation. The nuclear engineering department worked on three separate projects. The NEPTUNE system is a manned, outer-planetary explorer designed with Jupiter exploration as the baseline mission. The lifetime requirement for both reactor and power-conversion systems was twenty years. The second project undertaken for the power supply was a Mars Sample Return Mission power supply. This was designed to produce 2 kW of electrical power for seven years. The design consisted of a General Purpose Heat Source (GPHS) utilizing a Stirling engine as the power conversion unit. A mass optimization was performed to aid in overall design. The last design was a reactor to provide power for propulsion to Mars and power on the surface. The requirements of 300 kW of electrical power output and a mass of less than 10,000 Rg were set. This allowed the reactor and power conversion unit to fit within the Space Shuttle cargo bay.

Soviet space nuclear reactor incidents - Perception versus reality

NASA Technical Reports Server (NTRS)

Bennett, Gary L.

1992-01-01

Since the Soviet Union reportedly began flying nuclear power sources in 1965 it has had four publicly known accidents involving space reactors, two publicly known accidents involving radioisotope power sources and one close call with a space reactor (Cosmos 1900). The reactor accidents, particularly Cosmos 954 and Cosmos 1402, indicated that the Soviets had adopted burnup as their reentry philosophy which is consistent with the U.S. philosophy from the 1960s and 1970s. While quantitative risk analyses have shown that the Soviet accidents have not posed a serious risk to the world's population, concerns still remain about Soviet space nuclear safety practices.

Materials technology for an advanced space power nuclear reactor concept: Program summary

NASA Technical Reports Server (NTRS)

Gluyas, R. E.; Watson, G. K.

1975-01-01

The results of a materials technology program for a long-life (50,000 hr), high-temperature (950 C coolant outlet), lithium-cooled, nuclear space power reactor concept are reviewed and discussed. Fabrication methods and compatibility and property data were developed for candidate materials for fuel pins and, to a lesser extent, for potential control systems, reflectors, reactor vessel and piping, and other reactor structural materials. The effects of selected materials variables on fuel pin irradiation performance were determined. The most promising materials for fuel pins were found to be 85 percent dense uranium mononitride (UN) fuel clad with tungsten-lined T-111 (Ta-8W-2Hf).

Analysis of a boron-carbide-drum-controlled critical reactor experiment

NASA Technical Reports Server (NTRS)

Mayo, W. T.

1972-01-01

In order to validate methods and cross sections used in the neutronic design of compact fast-spectrum reactors for generating electric power in space, an analysis of a boron-carbide-drum-controlled critical reactor was made. For this reactor the transport analysis gave generally satisfactory results. The calculated multiplication factor for the most detailed calculation was only 0.7-percent Delta k too high. Calculated reactivity worth of the control drums was $11.61 compared to measurements of $11.58 by the inverse kinetics methods and $11.98 by the inverse counting method. Calculated radial and axial power distributions were in good agreement with experiment.

Extensions of the MCNP5 and TRIPOLI4 Monte Carlo Codes for Transient Reactor Analysis

NASA Astrophysics Data System (ADS)

Hoogenboom, J. Eduard; Sjenitzer, Bart L.

2014-06-01

To simulate reactor transients for safety analysis with the Monte Carlo method the generation and decay of delayed neutron precursors is implemented in the MCNP5 and TRIPOLI4 general purpose Monte Carlo codes. Important new variance reduction techniques like forced decay of precursors in each time interval and the branchless collision method are included to obtain reasonable statistics for the power production per time interval. For simulation of practical reactor transients also the feedback effect from the thermal-hydraulics must be included. This requires coupling of the Monte Carlo code with a thermal-hydraulics (TH) code, providing the temperature distribution in the reactor, which affects the neutron transport via the cross section data. The TH code also provides the coolant density distribution in the reactor, directly influencing the neutron transport. Different techniques for this coupling are discussed. As a demonstration a 3x3 mini fuel assembly with a moving control rod is considered for MCNP5 and a mini core existing of 3x3 PWR fuel assemblies with control rods and burnable poisons for TRIPOLI4. Results are shown for reactor transients due to control rod movement or withdrawal. The TRIPOLI4 transient calculation is started at low power and includes thermal-hydraulic feedback. The power rises about 10 decades and finally stabilises the reactor power at a much higher level than initial. The examples demonstrate that the modified Monte Carlo codes are capable of performing correct transient calculations, taking into account all geometrical and cross section detail.

Significance of breeding in fast nuclear reactors

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

Raza, S.M.; Abidi, S.B.M.

1983-12-01

Only breeder reactors--nuclear power plants that produce more fuel than they consume--are capable in principle of extracting the maximum amount of fission energy contained in uranium ore, thus offering a practical long-term solution to uranium supply problems. Uranium would then constitute a virtually inexhaustible fuel reserve for the world's future energy needs. The ultimate argument for breeding is to conserve the energy resources available to mankind. A long-term role for nuclear power with fast reactors is proven to be economically viable, environmentally acceptable and capable of wide scale exploitation in many countries. In this paper, various suggestions pertaining to themore » fuel fabrication route, fuel cycle economics, studies of the physics of fast nuclear reactors and of engineering design simplifications are presented. Fast reactors contain no moderator and inherently require enriched fuel. In general, the main aim is to suggest an improvement in the understanding of the safety and control characteristics of fast breeder power reactors. Development work is also being devoted to new carbide and nitride fuels, which are likely to exhibit breeding characteristics superior to those of the oxides of plutonium and uranium.« less

Degradation of Acid Orange 7 Dye in Two Hybrid Plasma Discharge Reactors

NASA Astrophysics Data System (ADS)

Shen, Yongjun; Lei, Lecheng; Zhang, Xingwang; Ding, Jiandong

2014-11-01

To get an optimized pulsed electrical plasma discharge reactor and to increase the energy utilization efficiency in the removal of pollutants, two hybrid plasma discharge reactors were designed and optimized. The reactors were compared via the discharge characteristics, energy transfer efficiency, the yields of the active species and the energy utilization in dye wastewater degradation. The results showed that under the same AC input power, the characteristics of the discharge waveform of the point-to-plate reactor were better. Under the same AC input power, the two reactors both had almost the same peak voltage of 22 kV. The peak current of the point-to-plate reactor was 146 A, while that of the wire-to-cylinder reactor was only 48.8 A. The peak powers of the point-to-plate reactor and the wire-to-cylinder reactor were 1.38 MW and 1.01 MW, respectively. The energy per pulse of the point-to-plate reactor was 0.2221 J, which was about 29.4% higher than that of the wire-to-cylinder reactor (0.1716 J). To remove 50% Acid Orange 7 (AO7), the energy utilizations of the point-to-plate reactor and the wire-to-cylinder reactor were 1.02 × 10-9 mol/L and 0.61 × 10-9 mol/L, respectively. In the point-to-plate reactor, the concentration of hydrogen peroxide in pure water was 3.6 mmol/L after 40 min of discharge, which was higher than that of the wire-to-cylinder reactor (2.5 mmol/L). The concentration of liquid phase ozone in the point-to-plate reactor (5.7 × 10-2 mmol/L) was about 26.7% higher than that in the wire-to-cylinder reactor (4.5 × 10-2 mmol/L). The analysis results of the variance showed that the type of reactor and reaction time had significant impacts on the yields of the hydrogen peroxide and ozone. The main degradation intermediates of AO7 identified by gas chromatography and mass spectrometry (GCMS) were acetic acid, maleic anhydride, p-benzoquinone, phenol, benzoic acid, phthalic anhydride, coumarin and 2-naphthol. Proposed degradation pathways were elucidated in light of the analyzed degradation products.

Function of university reactors in operator licensing training for nuclear utilities

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

Wicks, F.

1985-11-01

The director of the Division of the US Nuclear Regulatory Commission in generic letter 84-10, dated April 26, 1984, spoke the requirement that applicants for senior reactor operator licenses for power reactors shall have performed then reactor startups. Simulator startups were not acknowledged. Startups performed on a university reactor are acceptable. The content and results of a five-day program combining instruction and experiments with the Rensselaer reactor are summarized.

A fission-fusion hybrid reactor in steady-state L-mode tokamak configuration with natural uranium

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

Reed, Mark; Parker, Ronald R.; Forget, Benoit

2012-06-19

This work develops a conceptual design for a fusion-fission hybrid reactor operating in steady-state L-mode tokamak configuration with a subcritical natural or depleted uranium pebble bed blanket. A liquid lithium-lead alloy breeds enough tritium to replenish that consumed by the D-T fusion reaction. The fission blanket augments the fusion power such that the fusion core itself need not have a high power gain, thus allowing for fully non-inductive (steady-state) low confinement mode (L-mode) operation at relatively small physical dimensions. A neutron transport Monte Carlo code models the natural uranium fission blanket. Maximizing the fission power gain while breeding sufficient tritiummore » allows for the selection of an optimal set of blanket parameters, which yields a maximum prudent fission power gain of approximately 7. A 0-D tokamak model suffices to analyze approximate tokamak operating conditions. This fission blanket would allow the fusion component of a hybrid reactor with the same dimensions as ITER to operate in steady-state L-mode very comfortably with a fusion power gain of 6.7 and a thermal fusion power of 2.1 GW. Taking this further can determine the approximate minimum scale for a steady-state L-mode tokamak hybrid reactor, which is a major radius of 5.2 m and an aspect ratio of 2.8. This minimum scale device operates barely within the steady-state L-mode realm with a thermal fusion power of 1.7 GW. Basic thermal hydraulic analysis demonstrates that pressurized helium could cool the pebble bed fission blanket with a flow rate below 10 m/s. The Brayton cycle thermal efficiency is 41%. This reactor, dubbed the Steady-state L-mode non-Enriched Uranium Tokamak Hybrid (SLEUTH), with its very fast neutron spectrum, could be superior to pure fission reactors in terms of breeding fissile fuel and transmuting deleterious fission products. It would likely function best as a prolific plutonium breeder, and the plutonium it produces could actually be more proliferation-resistant than that bred by conventional fast reactors. Furthermore, it can maintain constant total hybrid power output as burnup proceeds by varying the neutron source strength.« less

Graphite for the nuclear industry

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

Burchell, T.D.; Fuller, E.L.; Romanoski, G.R.

Graphite finds applications in both fission and fusion reactors. Fission reactors harness the energy liberated when heavy elements, such as uranium or plutonium, fragment or fission''. Reactors of this type have existed for nearly 50 years. The first nuclear fission reactor, Chicago Pile No. 1, was constructed of graphite under a football stand at Stagg Field, University of Chicago. Fusion energy devices will produce power by utilizing the energy produced when isotopes of the element hydrogen are fused together to form helium, the same reaction that powers our sun. The role of graphite is very different in these two reactormore » systems. Here we summarize the function of the graphite in fission and fusion reactors, detailing the reasons for their selection and discussing some of the challenges associated with their application in nuclear fission and fusion reactors. 10 refs., 15 figs., 1 tab.« less

Prospects for development of an innovative water-cooled nuclear reactor for supercritical parameters of coolant

NASA Astrophysics Data System (ADS)

Kalyakin, S. G.; Kirillov, P. L.; Baranaev, Yu. D.; Glebov, A. P.; Bogoslovskaya, G. P.; Nikitenko, M. P.; Makhin, V. M.; Churkin, A. N.

2014-08-01

The state of nuclear power engineering as of February 1, 2014 and the accomplished elaborations of a supercritical-pressure water-cooled reactor are briefly reviewed, and the prospects of this new project are discussed based on this review. The new project rests on the experience gained from the development and operation of stationary water-cooled reactor plants, including VVERs, PWRs, BWRs, and RBMKs (their combined service life totals more than 15 000 reactor-years), and long-term experience gained around the world with operation of thermal power plants the turbines of which are driven by steam with supercritical and ultrasupercritical parameters. The advantages of such reactor are pointed out together with the scientific-technical problems that need to be solved during further development of such installations. The knowledge gained for the last decade makes it possible to refine the concept and to commence the work on designing an experimental small-capacity reactor.