Kinetic modeling of the photocatalytic degradation of clofibric acid in a slurry reactor.

PubMed

Manassero, Agustina; Satuf, María Lucila; Alfano, Orlando Mario

2015-01-01

A kinetic study of the photocatalytic degradation of the pharmaceutical clofibric acid is presented. Experiments were carried out under UV radiation employing titanium dioxide in water suspension. The main reaction intermediates were identified and quantified. Intrinsic expressions to represent the kinetics of clofibric acid and the main intermediates were derived. The modeling of the radiation field in the reactor was carried out by Monte Carlo simulation. Experimental runs were performed by varying the catalyst concentration and the incident radiation. Kinetic parameters were estimated from the experiments by applying a non-linear regression procedure. Good agreement was obtained between model predictions and experimental data, with an error of 5.9 % in the estimations of the primary pollutant concentration.

Cooling Performance Analysis of ThePrimary Cooling System ReactorTRIGA-2000Bandung

NASA Astrophysics Data System (ADS)

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

2018-02-01

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

Immobilization: A Revolution in Traditional Brewing

NASA Astrophysics Data System (ADS)

Virkajärvi, Ilkka; Linko, Matti

In nature many micro-organisms tend to bind to solid surfaces. This tendency has long been utilized in a number of processes, for example in producing vinegar and acetic acid in bioreactors filled with wood shavings. Acetobacteria are attached to the surface of these shavings. In modern technical language: they are immobilized. Also yeast cells can be immobilized. In the brewing industry this has been the basis for maintaining efficient, continuous fermentation in bioreactors with very high yeast concentrations. The most dramatic change in brewing over recent years has been the replacement of traditional lagering of several weeks by a continuous process in which the residence time is only about 2h. Continuous primary fermentation is used on a commercial scale in New Zealand. In this process, instead of a carrier, yeast is retained in reactors by returning it partly after separation. In many pilot scale experiments the primary fermentation is shortened from about 1week to 1-2days using immobilized yeast reactors. When using certain genetically modified yeast strains no secondary fermentation is needed, and the total fermentation time in immobilized yeast reactors can therefore be shortened to only 2days.

Simulation of German PKL refill/reflood experiment K9A using RELAP4/MOD7. [PWR

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

Hsu, M.T.; Davis, C.B.; Behling, S.R.

This paper describes a RELAP4/MOD7 simulation of West Germany's Kraftwerk Union (KWU) Primary Coolant Loop (PKL) refill/reflood experiment K9A. RELAP4/MOD7, a best-estimate computer program for the calculation of thermal and hydraulic phenomena in a nuclear reactor or related system, is the latest version in the RELAP4 code development series. This study was the first major simulation using RELAP4/MOD7 since its release by the Idaho National Engineering Laboratory (INEL). The PKL facility is a reduced scale (1:134) representation of a typical West German four-loop 1300 MW pressurized water reactor (PWR). A prototypical scale of the total volume to power ratio wasmore » maintained. The test facility was designed specifically for an experiment simulating the refill/reflood phase of a Loss-of-Coolant Accident (LOCA).« less

The Experimental Breeder Reactor II seismic probabilistic risk assessment

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

Roglans, J; Hill, D J

1994-02-01

The Experimental Breeder Reactor II (EBR-II) is a US Department of Energy (DOE) Category A research reactor located at Argonne National Laboratory (ANL)-West in Idaho. EBR-II is a 62.5 MW-thermal Liquid Metal Reactor (LMR) that started operation in 1964 and it is currently being used as a testbed in the Integral Fast Reactor (IFR) Program. ANL has completed a Level 1 Probabilistic Risk Assessment (PRA) for EBR-II. The Level 1 PRA for internal events and most external events was completed in June 1991. The seismic PRA for EBR-H has recently been completed. The EBR-II reactor building contains the reactor, themore » primary system, and the decay heat removal systems. The reactor vessel, which contains the core, and the primary system, consisting of two primary pumps and an intermediate heat exchanger, are immersed in the sodium-filled primary tank, which is suspended by six hangers from a beam support structure. Three systems or functions in EBR-II were identified as the most significant from the standpoint of risk of seismic-induced fuel damage: (1) the reactor shutdown system, (2) the structural integrity of the passive decay heat removal systems, and (3) the integrity of major structures, like the primary tank containing the reactor that could threaten both the reactivity control and decay heat removal functions. As part of the seismic PRA, efforts were concentrated in studying these three functions or systems. The passive safety response of EBR-II reactor -- both passive reactivity shutdown and passive decay heat removal, demonstrated in a series of tests in 1986 -- was explicitly accounted for in the seismic PRA as it had been included in the internal events assessment.« less

NuLat: A Novel Design for a Reactor Anti-Neutrino Detector

NASA Astrophysics Data System (ADS)

Rountree, S. Derek; NuLat Collaboration

2015-04-01

NuLat is a proposed very-short baseline (3-10m) reactor electron antineutrino (anti-νe) experiment that will probe the current best fit for light sterile neutrino mixing, the 5 MeV excess seen in current short baseline reactor experiments, and serve as a portable surface detector for cooperative (~ 30m baseline) surface monitoring of reactors. The NuLat detector will use an optically segmented 3D Raghavan optical lattice (ROL) detector that channels light via total internal reflection from a scintillation event down the 3 primary axes to the detector faces. The high degree of segmentation allows for each voxel's energy to be determined independently of other voxels, thus providing high temporal and spatial resolution and energy reconstruction independent of position. NuLat detects anti-νe via inverse beta decay (IBD), which produces a positron and a neutron. Most of the time, the positron deposits its kinetic energy into a single voxel allowing superior derivation of the incident anti-νe's energy. The final state neutron is captured via (n, α) on 6 Li or 10 B after a characteristic delay time giving a coincidence tag. This talk will discuss the physics reach of NuLat using a solid loaded scintillator, and the timeline of the NuLat reactor anti-νe program. This research has been funded in part by the National Science Foundation on Award Numbers 1001394 and 1001078.

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

DOEpatents

Hunsbedt, Anstein; Boardman, Charles E.

1993-01-01

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

Multiple Irradiation Capsule Experiment (MICE)-3B Irradiation Test of Space Fuel Specimens in the Advanced Test Reactor (ATR) - Close Out Documentation for Naval Reactors (NR) Information

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

M. Chen; CM Regan; D. Noe

2006-01-09

Few data exist for UO{sub 2} or UN within the notional design space for the Prometheus-1 reactor (low fission rate, high temperature, long duration). As such, basic testing is required to validate predictions (and in some cases determine) performance aspects of these fuels. Therefore, the MICE-3B test of UO{sub 2} pellets was designed to provide data on gas release, unrestrained swelling, and restrained swelling at the upper range of fission rates expected for a space reactor. These data would be compared with model predictions and used to determine adequacy of a space reactor design basis relative to fission gas releasemore » and swelling of UO{sub 2} fuel and to assess potential pellet-clad interactions. A primary goal of an irradiation test for UN fuel was to assess performance issues currently associated with this fuel type such as gas release, swelling and transient performance. Information learned from this effort may have enabled use of UN fuel for future applications.« less

DynMo: Dynamic Simulation Model for Space Reactor Power Systems

NASA Astrophysics Data System (ADS)

El-Genk, Mohamed; Tournier, Jean-Michel

2005-02-01

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

Effects of combustion temperature on PCDD/Fs formation in laboratory-scale fluidized-bed incineration.

PubMed

Hatanaka, T; Imagawa, T; Kitajima, A; Takeuchi, M

2001-12-15

Combustion experiments in a laboratory-scale fluidized-bed reactor were performed to elucidate the effects of combustion temperature on PCDD/Fs formation during incineration of model wastes with poly(vinyl chloride) or sodium chloride as a chlorine source and copper chloride as a catalyst. Each temperature of primary and secondary combustion zones in the reactor was set independently to 700, 800, and 900 degrees C using external electric heaters. The PCDD/Fs concentration is reduced as the temperature of the secondary combustion zone increases. It is effective to keep the temperature of the secondary combustion zone high enough to reduce their release during the waste incineration. On the other hand, as the temperature of the primary combustion zone rises, the PCDD/Fs concentration also increases. Lower temperature of the primary combustion zone results in less PCDD/Fs concentration in these experimental conditions. This result is probably related to the devolatilization rate of the solid waste in the primary combustion zone. The temperature decrease slows the devolatilization rate and promotes mixing of oxygen and volatile matters from the solid waste. This contributes to completing combustion reactions, resulting in reducing the PCDD/Fs concentration.

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

77 FR 42771 - License Renewal for the Dow Chemical TRIGA Research Reactor

Federal Register 2010, 2011, 2012, 2013, 2014

2012-07-20

... Chemical Company in Midland, MI and is a part of the Analytical Sciences Laboratory. The reactor is housed...-Radiological Impacts The Dow TRIGA Research Reactor core is cooled by a light water primary system consisting... provided by the volume of primary coolant allows several hours of full-power operation without any...

Measurements Methods for the analysis of Nuclear Reactors Thermal Hydraulic in Water Scaled Facilities

NASA Astrophysics Data System (ADS)

Spaccapaniccia, C.; Planquart, P.; Buchlin, J. M. AB(; ), AC(; )

2018-01-01

The Belgian nuclear research institute (SCK•CEN) is developing MYRRHA. MYRRHA is a flexible fast spectrum research reactor, conceived as an accelerator driven system (ADS). The configuration of the primary loop is pool-type: the primary coolant and all the primary system components (core and heat exchangers) are contained within the reactor vessel, while the secondary fluid is circulating in the heat exchangers. The primary coolant is Lead Bismuth Eutectic (LBE). The recent nuclear accident of Fukushima in 2011 changed the requirements for the design of new reactors, which should include the possibility to remove the residual decay heat through passive primary and secondary systems, i.e. natural convection (NC). After the reactor shut down, in the unlucky event of propeller failures, the primary and secondary loops should be able to remove the decay heat in passive way (Natural Convection). The present study analyses the flow and the temperature distribution in the upper plenum by applying laser imaging techniques in a laboratory scaled water model. A parametric study is proposed to study stratification mitigation strategies by varying the geometry of the buffer tank simulating the upper plenum.

IAEA coordinated research project on thermal-hydraulics of Supercritical Water-Cooled Reactors (SCWRs)

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

Yamada, K.; Aksan, S. N.

The Supercritical Water-Cooled Reactor (SCWR) is an innovative water-cooled reactor concept, which uses supercritical pressure water as reactor coolant. It has been attracting interest of many researchers in various countries mainly due to its benefits of high thermal efficiency and simple primary systems, resulting in low capital cost. The IAEA started in 2008 a Coordinated Research Project (CRP) on Thermal-Hydraulics of SCWRs as a forum to foster the exchange of technical information and international collaboration in research and development. This paper summarizes the activities and current status of the CRP, as well as major progress achieved to date. At present,more » 15 institutions closely collaborate in several tasks. Some organizations have been conducting thermal-hydraulics experiments and analysing the data, and others have been participating in code-to-test and/or code-to-code benchmark exercises. The expected outputs of the CRP are also discussed. Finally, the paper introduces several IAEA activities relating to or arising from the CRP. (authors)« less

Estimated inventory of radionuclides in former Soviet Union naval reactors dumped in the Kara Sea

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

Mount, M.E.; Sheaffer, M.K.; Abbott, D.T.

1993-07-01

Radionuclide inventories have been estimated for the reactor cores, reactor components, and primary system corrosion products in the former Soviet Union naval reactors dumped at the Abrosimov Inlet, Tsivolka Inlet, Stepovoy Inlet, Techeniye Inlet, and Novaya Zemlya Depression sites in the Kara Sea between 1965 and 1988. For the time of disposal, the inventories are estimated at 69 to 111 kCi of actinides plus daughters and 3,053 to 7,472 kCi of fission products in the reactor cores, 917 to 1,127 kCi of activation products in the reactor components, and 1.4 to 1.6 kCi of activation products in the primary systemmore » corrosion products. At the present time, the inventories are estimated to have decreased to 23 to 38 kCi of actinides plus daughters and 674 to 708 kCi of fission products in the reactor cores, 124 to 126 kCi of activation products in the reactor components, and 0.16 to 0.17 kCi of activation products in the primary system corrosion products. Twenty years from now, the inventories are projected to be 11 to 18 kCi of actinides plus daughters and 415 to 437 kCi of fission products in the reactor cores, 63.5 to 64 kCi of activation products in the reactor components, and 0.014 to 0.015 kCi of activation products in the primary system corrosion products. All actinide activities are estimated to be within a factor of two.« less

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

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

Colley, J.R.

1962-12-01

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

The Use Of High Speed Photography In Reactor Safety Studies At The Atomic Energy Establishment, Winfrith

NASA Astrophysics Data System (ADS)

Maddison, R. J.

1985-02-01

The investigation of certain areas of nuclear reactor safety involves the study of high speed phenomena with timescales ranging from microseconds to a few hundreds of milliseconds. Examples which have been extensively studied at Winfrith are firstly, the thermal interaction of molten fuel and reactor coolant which can generate high pressures on the 100 msec timescale, and which involves phenomena such as vapour film collapse which takes place on the microsecond timescale. Secondly, there is the response of reactor structures to such pressures, and finally there is the response of structural materials such as metals and concrete to the impulsive loading arising from the impact of heavy, high velocity missiles. A wide range of experimental techniques is used in these studies, many of which have been developed specially for this type of work which ranges from small laboratory scale to large field scale experiments. There are two important features which characterise many of these experiments:- i) a long period of meticulous preparation of very heavily instrumented, short duration experiments and; ii) the destructive nature of the experiments. Various forms of High Speed photography are included in the inventory of experimental techniques. These include the use of single and double exposure, short duration, spark photography; the use of an Image Convertor Camera (IMACON 790); and a number of rotating prism cine cameras. High Speed Photography is used both in a primary experimental role in the studies, and in a supportive role for other instrumentation. Because of the sometimes violent nature of these experiments, cameras are often heavily protected and operated remotely; lighting systems are sometimes destroyed. This has led to the development of unconventional techniques for camera operation and subject lighting. This paper will describe some of the experiments and the way in which High Speed Photography has been applied as an essential experimental tool. It will be illustrated with cine film taken during the experiments.

CRBR pump water test experience

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

Cook, M.E.; Huber, K.A.

1983-01-01

The hydraulic design features and water testing of the hydraulic scale model and prototype pump of the sodium pumps used in the primary and intermediate sodium loops of the Clinch River Breeder Reactor Plant (CRBRP) are described. The Hydraulic Scale Model tests are performed and the results of these tests are discussed. The Prototype Pump tests are performed and the results of these tests are discussed.

An impact source localization technique for a nuclear power plant by using sensors of different types.

PubMed

Choi, Young-Chul; Park, Jin-Ho; Choi, Kyoung-Sik

2011-01-01

In a nuclear power plant, a loose part monitoring system (LPMS) provides information on the location and the mass of a loosened or detached metal impacted onto the inner surface of the primary pressure boundary. Typically, accelerometers are mounted on the surface of a reactor vessel to localize the impact location caused by the impact of metallic substances on the reactor system. However, in some cases, the number of accelerometers is not sufficient to estimate the impact location precisely. In such a case, one of useful methods is to utilize other types of sensor that can measure the vibration of the reactor structure. For example, acoustic emission (AE) sensors are installed on the reactor structure to detect leakage or cracks on the primary pressure boundary. However, accelerometers and AE sensors have a different frequency range. The frequency of interest of AE sensors is higher than that of accelerometers. In this paper, we propose a method of impact source localization by using both accelerometer signals and AE signals, simultaneously. The main concept of impact location estimation is based on the arrival time difference of the impact stress wave between different sensor locations. However, it is difficult to find the arrival time difference between sensors, because the primary frequency ranges of accelerometers and AE sensors are different. To overcome the problem, we used phase delays of an envelope of impact signals. This is because the impact signals from the accelerometer and the AE sensor are similar in the whole shape (envelope). To verify the proposed method, we have performed experiments for a reactor mock-up model and a real nuclear power plant. The experimental results demonstrate that we can enhance the reliability and precision of the impact source localization. Therefore, if the proposed method is applied to a nuclear power plant, we can obtain the effect of additional installed sensors. Crown Copyright © 2010. Published by Elsevier Ltd. All rights reserved.

78 FR 28896 - Design Limits and Loading Combinations for Metal Primary Reactor Containment System Components

Federal Register 2010, 2011, 2012, 2013, 2014

2013-05-16

... NUCLEAR REGULATORY COMMISSION [NRC-2013-0095] Design Limits and Loading Combinations for Metal... Regulatory Guide (RG) 1.57, ``Design Limits and Loading Combinations for Metal Primary Reactor Containment... the NRC staff considers acceptable for design limits and loading combinations for metal primary...

Safety and core design of large liquid-metal cooled fast breeder reactors

NASA Astrophysics Data System (ADS)

Qvist, Staffan Alexander

In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cycle to power the entire world for several centuries to come. Breed & burn (B&B) type fast reactor cores can unlock the energy potential of readily available fertile material such as depleted uranium without the need for chemical reprocessing. Using B&B technology, nuclear waste generation, uranium mining needs and proliferation concerns can be greatly reduced, and after a transitional period, enrichment facilities may no longer be needed. In this dissertation, new passively operating safety systems for fast reactors cores are presented. New analysis and optimization methods for B&B core design have been developed, along with a comprehensive computer code that couples neutronics, thermal-hydraulics and structural mechanics and enables a completely automated and optimized fast reactor core design process. In addition, an experiment that expands the knowledge-base of corrosion issues of lead-based coolants in nuclear reactors was designed and built. The motivation behind the work presented in this thesis is to help facilitate the widespread adoption of safe and efficient fast reactor technology.

Fluidized bed silicon deposition from silane

NASA Technical Reports Server (NTRS)

Hsu, George C. (Inventor); Levin, Harry (Inventor); Hogle, Richard A. (Inventor); Praturi, Ananda (Inventor); Lutwack, Ralph (Inventor)

1982-01-01

A process and apparatus for thermally decomposing silicon containing gas for deposition on fluidized nucleating silicon seed particles is disclosed. Silicon seed particles are produced in a secondary fluidized reactor by thermal decomposition of a silicon containing gas. The thermally produced silicon seed particles are then introduced into a primary fluidized bed reactor to form a fluidized bed. Silicon containing gas is introduced into the primary reactor where it is thermally decomposed and deposited on the fluidized silicon seed particles. Silicon seed particles having the desired amount of thermally decomposed silicon product thereon are removed from the primary fluidized reactor as ultra pure silicon product. An apparatus for carrying out this process is also disclosed.

Fluidized bed silicon deposition from silane

NASA Technical Reports Server (NTRS)

Hsu, George (Inventor); Levin, Harry (Inventor); Hogle, Richard A. (Inventor); Praturi, Ananda (Inventor); Lutwack, Ralph (Inventor)

1984-01-01

A process and apparatus for thermally decomposing silicon containing gas for deposition on fluidized nucleating silicon seed particles is disclosed. Silicon seed particles are produced in a secondary fluidized reactor by thermal decomposition of a silicon containing gas. The thermally produced silicon seed particles are then introduced into a primary fluidized bed reactor to form a fludized bed. Silicon containing gas is introduced into the primary reactor where it is thermally decomposed and deposited on the fluidized silicon seed particles. Silicon seed particles having the desired amount of thermally decomposed silicon product thereon are removed from the primary fluidized reactor as ultra pure silicon product. An apparatus for carrying out this process is also disclosed.

POWER REACTOR

DOEpatents

Zinn, W.H.

1958-07-01

A fast nuclear reactor system ls described for producing power and radioactive isotopes. The reactor core is of the heterogeneous, fluid sealed type comprised of vertically arranged elongated tubular fuel elements having vertical coolant passages. The active portion is surrounded by a neutron reflector and a shield. The system includes pumps and heat exchangers for the primary and secondary coolant circuits. The core, primary coolant pump and primary heat exchanger are disposed within an irapenforate tank which is filled with the primary coolant, in this case a liquid metal such as Na or NaK, to completely submerge these elements. The tank is completely surrounded by a thick walled concrete shield. This reactor system utilizes enriched uranium or plutonium as the fissionable material, uranium or thorium as a diluent and thorium or uranium containing less than 0 7% of the U/sup 235/ isotope as a fertile material.

Estimated inventory of radionuclides in Former Soviet Union Naval Reactors dumped in the Kara Sea and their associated health risk

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

Mount, M.E.; Layton, D.W.; Schwertz, N.L.

1993-05-01

Radionuclide inventories have bin estimated for the reactor cores, reactor components, and primary system corrosion products in the former Soviet Union naval reactors dumped at the Abrosimov Inlet, Tsivolka Inlet, Stepovoy Inlet, Techeniye Inlet, and Novaya Zemlya Depression sites in the Kara Sea between 1965 and 1988. For the time of disposal, the inventories are estimated at 17 to 66 kCi of actinides plus daughters and 1695 to 4782 kCi of fission products in the reactor cores, 917 to 1127 kCi of activation products in the reactor components, and 1.4 to 1.6 kCi of activation products in the primary systemmore » corrosion products. At the present time, the inventories are estimated to have decreased to 6 to 24 kCi of actinides plus daughters and 492 to 540 kCi of fission products in the reactor cores, 124 to 126 kCi of activation products in the reactor components, and 0.16 to 0.17 kCi of activation products in the primary system corrosion products. All actinide activities are estimated to be within a factor of two.« less

78 FR 58575 - Review of Experiments for Research Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-09-24

... NUCLEAR REGULATORY COMMISSION [NRC-2013-0219] Review of Experiments for Research Reactors AGENCY... Commission (NRC) is withdrawing Regulatory Guide (RG) 2.4, ``Review of Experiments for Research Reactors... withdrawing RG 2.4, ``Review of Experiments for Research Reactors,'' (ADAMS Accession No. ML003740131) because...

IEA-R1 Nuclear Research Reactor: 58 Years of Operating Experience and Utilization for Research, Teaching and Radioisotopes Production

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

Cardenas, Jose Patricio Nahuel; Filho, Tufic Madi; Saxena, Rajendra

IEA-R1 research reactor at the Instituto de Pesquisas Energeticas e Nucleares (Nuclear and Energy Research Institute) IPEN, Sao Paulo, Brazil is the largest power research reactor in Brazil, with a maximum power rating of 5 MWth. It is being used for basic and applied research in the nuclear and neutron related sciences, for the production of radioisotopes for medical and industrial applications, and for providing services of neutron activation analysis, real time neutron radiography, and neutron transmutation doping of silicon. IEA-R1 is a swimming pool reactor, with light water as the coolant and moderator, and graphite and beryllium as reflectors.more » The reactor was commissioned on September 16, 1957 and achieved its first criticality. It is currently operating at 4.5 MWth with a 60-hour cycle per week. In the early sixties, IPEN produced {sup 131}I, {sup 32}P, {sup 198}Au, {sup 24}Na, {sup 35}S, {sup 51}Cr and labeled compounds for medical use. During the past several years, a concerted effort has been made in order to upgrade the reactor power to 5 MWth through refurbishment and modernization programs. One of the reasons for this decision was to produce {sup 99}Mo at IPEN. The reactor cycle will be gradually increased to 120 hours per week continuous operation. It is anticipated that these programs will assure the safe and sustainable operation of the IEA-R1 reactor for several more years, to produce important primary radioisotopes {sup 99}Mo, {sup 125}I, {sup 131}I, {sup 153}Sm and {sup 192}Ir. Currently, all aspects of dealing with fuel element fabrication, fuel transportation, isotope processing, and spent fuel storage are handled by IPEN at the site. The reactor modernization program is slated for completion by 2015. This paper describes 58 years of operating experience and utilization of the IEA-R1 research reactor for research, teaching and radioisotopes production. (authors)« less

Cooling system for a nuclear reactor

DOEpatents

Amtmann, Hans H.

1982-01-01

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

Efficiency of a pilot-scale integrated sludge thickening and digestion reactor in treating low-organic excess sludge.

PubMed

He, Qiang; Li, Jiang; Liu, Hongxia; Tang, Chuandong; de Koning, Jaap; Spanjers, Henri

2012-06-01

The sludge production from medium- and small-scale wastewater treatment plants in the Three Gorges Reservoir Region is low and non-stable; especially, the organic content in this sludge is low (near 40% of VS/TS). An integrated thickening and digestion (ISTD) reactor was developed to treat this low-organic excess sludge. After a flow test and start-up experiment of the reactor, a running experiment was used to investigate the excess sludge treatment efficiency under five different excess sludge inflows: 200, 300, 400, 500 and 400 L/d (a mixture of excess sludge and primary sludge in a volume ratio of 9:1). This trial was carried out in the wastewater treatment plant in Chongqing, which covers 80% of the Three Gorges Reservoir Region, under the following conditions: (1) sludge was heated to 38-40 degrees C using an electrical heater to maintain anaerobic mesophilic digestion; (2) the biogas produced was recirculated to mix raw sludge with anaerobic sludge in the reactor under the flow rate of 12.5 L/min. There were three main results. Firstly, the flow pattern of the inner reactor was almost completely mixed under the air flow of 12.0 L/min using clear water. Secondly, under all the different sludge inflows, the water content in the outlet sludge was below 93%. Thirdly, the organic content in the outlet sludge was decreased from 37% to 30% and from 24% to 20%, whose removal ratio was in relation to the organic content of the inlet sludge. The excess sludge treatment capacity of the ISTD reactor was according to the organic content in the excess sludge.

Comparisons of RELAP5-3D Analyses to Experimental Data from the Natural Convection Shutdown Heat Removal Test Facility

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

Bucknor, Matthew; Hu, Rui; Lisowski, Darius

2016-04-17

The Reactor Cavity Cooling System (RCCS) is an important passive safety system being incorporated into the overall safety strategy for high temperature advanced reactor concepts such as the High Temperature Gas- Cooled Reactors (HTGR). The Natural Convection Shutdown Heat Removal Test Facility (NSTF) at Argonne National Laboratory (Argonne) reflects a 1/2-scale model of the primary features of one conceptual air-cooled RCCS design. The project conducts ex-vessel, passive heat removal experiments in support of Department of Energy Office of Nuclear Energy’s Advanced Reactor Technology (ART) program, while also generating data for code validation purposes. While experiments are being conducted at themore » NSTF to evaluate the feasibility of the passive RCCS, parallel modeling and simulation efforts are ongoing to support the design, fabrication, and operation of these natural convection systems. Both system-level and high fidelity computational fluid dynamics (CFD) analyses were performed to gain a complete understanding of the complex flow and heat transfer phenomena in natural convection systems. This paper provides a summary of the RELAP5-3D NSTF model development efforts and provides comparisons between simulation results and experimental data from the NSTF. Overall, the simulation results compared favorably to the experimental data, however, further analyses need to be conducted to investigate any identified differences.« less

Reflux cooling experiments on the NCSU scaled PWR facility

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

Doster, J.M.; Giavedoni, E.

1993-01-01

Under loss of forced circulation, coupled with the loss or reduction in primary side coolant inventory, horizontal stratified flows can develop in the hot and cold legs of pressurized water reactors (PWRs). Vapor produced in the reactor vessel is transported through the hot leg to the steam generator tubes where it condenses and flows back to the reactor vessel. Within the steam generator tubes, the flow regimes may range from countercurrent annular flow to single-phase convection. As a result, a number of heat transfer mechanisms are possible, depending on the loop configuration, total heat transfer rate, and the steam flowmore » rate within the tubes. These include (but are not limited to) two-phase natural circulation, where the condensate flows concurrent to the vapor stream and is transported to the cold leg so that the entire reactor coolant loop is active, and reflux cooling, where the condensate flows back down the interior of the coolant tubes countercurrent to the vapor stream and is returned to the reactor vessel through the hot leg. While operating in the reflux cooling mode, the cold leg can effectively be inactive. Heat transfer can be further influenced by noncondensables in the vapor stream, which accumulate within the upper regions of the steam generator tube bundle. In addition to reducing the steam generator's effective heat transfer area, under these conditions operation under natural circulation may not be possible, and reflux cooling may be the only viable heat transfer mechanism. The scaled PWR (SPWR) facility in the nuclear engineering department at North Carolina State Univ. (NCSU) is being used to study the effectiveness of two-phase natural circulation and reflux cooling under conditions associated with loss of forced circulation, midloop coolant levels, and noncondensables in the primary coolant system.« less

RELAP5 Analysis of the Hybrid Loop-Pool Design for Sodium Cooled Fast Reactors

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

Hongbin Zhang; Haihua Zhao; Cliff Davis

2008-06-01

An innovative hybrid loop-pool design for sodium cooled fast reactors (SFR-Hybrid) has been recently proposed. This design takes advantage of the inherent safety of a pool design and the compactness of a loop design to improve economics and safety of SFRs. In the hybrid loop-pool design, primary loops are formed by connecting the reactor outlet plenum (hot pool), intermediate heat exchangers (IHX), primary pumps and the reactor inlet plenum with pipes. The primary loops are immersed in the cold pool (buffer pool). Passive safety systems -- modular Pool Reactor Auxiliary Cooling Systems (PRACS) – are added to transfer decay heatmore » from the primary system to the buffer pool during loss of forced circulation (LOFC) transients. The primary systems and the buffer pool are thermally coupled by the PRACS, which is composed of PRACS heat exchangers (PHX), fluidic diodes and connecting pipes. Fluidic diodes are simple, passive devices that provide large flow resistance in one direction and small flow resistance in reverse direction. Direct reactor auxiliary cooling system (DRACS) heat exchangers (DHX) are immersed in the cold pool to transfer decay heat to the environment by natural circulation. To prove the design concepts, especially how the passive safety systems behave during transients such as LOFC with scram, a RELAP5-3D model for the hybrid loop-pool design was developed. The simulations were done for both steady-state and transient conditions. This paper presents the details of RELAP5-3D analysis as well as the calculated thermal response during LOFC with scram. The 250 MW thermal power conventional pool type design of GNEP’s Advanced Burner Test Reactor (ABTR) developed by Argonne National Laboratory was used as the reference reactor core and primary loop design. The reactor inlet temperature is 355 °C and the outlet temperature is 510 °C. The core design is the same as that for ABTR. The steady state buffer pool temperature is the same as the reactor inlet temperature. The peak cladding, hot pool, cold pool and reactor inlet temperatures were calculated during LOFC. The results indicate that there are two phases during LOFC transient – the initial thermal equilibration phase and the long term decay heat removal phase. The initial thermal equilibration phase occurs over a few hundred seconds, as the system adjusts from forced circulation to natural circulation flow. Subsequently, during long-term heat removal phase all temperatures evolve very slowly due to the large thermal inertia of the primary and buffer pool systems. The results clearly show that passive safety PRACS can effectively transfer decay heat from the primary system to the buffer pool by natural circulation. The DRACS system in turn can effectively transfer the decay heat to the environment.« less

Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

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

Brown, Nicholas R.; Mueller, Donald E.; Patton, Bruce W.

2016-08-31

Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 7LiF-BeF 2) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. Themore » objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.« 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

Pressurized fluidized bed reactor

DOEpatents

Isaksson, J.

1996-03-19

A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine. 1 fig.

Pressurized fluidized bed reactor

DOEpatents

Isaksson, Juhani

1996-01-01

A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine.

Evolution of the Reactor Antineutrino Flux and Spectrum at Daya Bay

DOE PAGES

An, F. P.; Balantekin, A. B.; Band, H. R.; ...

2017-06-19

Here, the Daya Bay experiment has observed correlations between reactor core fuel evolution and changes in the reactor antineutrino flux and energy spectrum. Four antineutrino detectors in two experimental halls were used to identify 2.2 million inverse beta decays (IBDs) over 1230 days spanning multiple fuel cycles for each of six 2.9 GW th reactor cores at the Daya Bay and Ling Ao nuclear power plants. Using detector data spanning effective 239Pu fission fractions F 239 from 0.25 to 0.35, Daya Bay measures an average IBD yield ¯σf of (5.90±0.13)×10 –43 cm 2/fission and a fuel-dependent variation in the IBDmore » yield, dσ f/dF 239, of (–1.86±0.18)×10 –43 cm 2/fission. This observation rejects the hypothesis of a constant antineutrino flux as a function of the 239Pu fission fraction at 10 standard deviations. The variation in IBD yield is found to be energy dependent, rejecting the hypothesis of a constant antineutrino energy spectrum at 5.1 standard deviations. While measurements of the evolution in the IBD spectrum show general agreement with predictions from recent reactor models, the measured evolution in total IBD yield disagrees with recent predictions at 3.1σ. This discrepancy indicates that an overall deficit in the measured flux with respect to predictions does not result from equal fractional deficits from the primary fission isotopes 235U, 239Pu, 238U, and 241Pu. Based on measured IBD yield variations, yields of (6.17±0.17) and (4.27±0.26)×10 –43 cm 2/fission have been determined for the two dominant fission parent isotopes 235U and 239Pu. A 7.8% discrepancy between the observed and predicted 235U yields suggests that this isotope may be the primary contributor to the reactor antineutrino anomaly.« less

Evolution of the Reactor Antineutrino Flux and Spectrum at Daya Bay.

PubMed

An, F P; Balantekin, A B; Band, H R; Bishai, M; Blyth, S; Cao, D; Cao, G F; Cao, J; Chan, Y L; Chang, J F; Chang, Y; Chen, H S; Chen, Q Y; Chen, S M; Chen, Y X; Chen, Y; Cheng, J; Cheng, Z K; Cherwinka, J J; Chu, M C; Chukanov, A; Cummings, J P; Ding, Y Y; Diwan, M V; Dolgareva, M; Dove, J; Dwyer, D A; Edwards, W R; Gill, R; Gonchar, M; Gong, G H; Gong, H; Grassi, M; Gu, W Q; Guo, L; Guo, X H; Guo, Y H; Guo, Z; Hackenburg, R W; Hans, S; He, M; Heeger, K M; Heng, Y K; Higuera, A; Hsiung, Y B; Hu, B Z; Hu, T; Huang, E C; Huang, H X; Huang, X T; Huang, Y B; Huber, P; Huo, W; Hussain, G; Jaffe, D E; Jen, K L; Ji, X P; Ji, X L; Jiao, J B; Johnson, R A; Jones, D; Kang, L; Kettell, S H; Khan, A; Kohn, S; Kramer, M; Kwan, K K; Kwok, M W; Langford, T J; Lau, K; Lebanowski, L; Lee, J; Lee, J H C; Lei, R T; Leitner, R; Leung, J K C; Li, C; Li, D J; Li, F; Li, G S; Li, Q J; Li, S; Li, S C; Li, W D; Li, X N; Li, X Q; Li, Y F; Li, Z B; Liang, H; Lin, C J; Lin, G L; Lin, S; Lin, S K; Lin, Y-C; Ling, J J; Link, J M; Littenberg, L; Littlejohn, B R; Liu, J L; Liu, J C; Loh, C W; Lu, C; Lu, H Q; Lu, J S; Luk, K B; Ma, X Y; Ma, X B; Ma, Y Q; Malyshkin, Y; Martinez Caicedo, D A; McDonald, K T; McKeown, R D; Mitchell, I; Nakajima, Y; Napolitano, J; Naumov, D; Naumova, E; Ngai, H Y; Ochoa-Ricoux, J P; Olshevskiy, A; Pan, H-R; Park, J; Patton, S; Pec, V; Peng, J C; Pinsky, L; Pun, C S J; Qi, F Z; Qi, M; Qian, X; Qiu, R M; Raper, N; Ren, J; Rosero, R; Roskovec, B; Ruan, X C; Steiner, H; Stoler, P; Sun, J L; Tang, W; Taychenachev, D; Treskov, K; Tsang, K V; Tull, C E; Viaux, N; Viren, B; Vorobel, V; Wang, C H; Wang, M; Wang, N Y; Wang, R G; Wang, W; Wang, X; Wang, Y F; Wang, Z; Wang, Z; Wang, Z M; Wei, H Y; Wen, L J; Whisnant, K; White, C G; Whitehead, L; Wise, T; Wong, H L H; Wong, S C F; Worcester, E; Wu, C-H; Wu, Q; Wu, W J; Xia, D M; Xia, J K; Xing, Z Z; Xu, J L; Xu, Y; Xue, T; Yang, C G; Yang, H; Yang, L; Yang, M S; Yang, M T; Yang, Y Z; Ye, M; Ye, Z; Yeh, M; Young, B L; Yu, Z Y; Zeng, S; Zhan, L; Zhang, C; Zhang, C C; Zhang, H H; Zhang, J W; Zhang, Q M; Zhang, R; Zhang, X T; Zhang, Y M; Zhang, Y X; Zhang, Y M; Zhang, Z J; Zhang, Z Y; Zhang, Z P; Zhao, J; Zhou, L; Zhuang, H L; Zou, J H

2017-06-23

The Daya Bay experiment has observed correlations between reactor core fuel evolution and changes in the reactor antineutrino flux and energy spectrum. Four antineutrino detectors in two experimental halls were used to identify 2.2 million inverse beta decays (IBDs) over 1230 days spanning multiple fuel cycles for each of six 2.9 GW_{th} reactor cores at the Daya Bay and Ling Ao nuclear power plants. Using detector data spanning effective ^{239}Pu fission fractions F_{239} from 0.25 to 0.35, Daya Bay measures an average IBD yield σ[over ¯]_{f} of (5.90±0.13)×10^{-43}  cm^{2}/fission and a fuel-dependent variation in the IBD yield, dσ_{f}/dF_{239}, of (-1.86±0.18)×10^{-43}  cm^{2}/fission. This observation rejects the hypothesis of a constant antineutrino flux as a function of the ^{239}Pu fission fraction at 10 standard deviations. The variation in IBD yield is found to be energy dependent, rejecting the hypothesis of a constant antineutrino energy spectrum at 5.1 standard deviations. While measurements of the evolution in the IBD spectrum show general agreement with predictions from recent reactor models, the measured evolution in total IBD yield disagrees with recent predictions at 3.1σ. This discrepancy indicates that an overall deficit in the measured flux with respect to predictions does not result from equal fractional deficits from the primary fission isotopes ^{235}U, ^{239}Pu, ^{238}U, and ^{241}Pu. Based on measured IBD yield variations, yields of (6.17±0.17) and (4.27±0.26)×10^{-43}  cm^{2}/fission have been determined for the two dominant fission parent isotopes ^{235}U and ^{239}Pu. A 7.8% discrepancy between the observed and predicted ^{235}U yields suggests that this isotope may be the primary contributor to the reactor antineutrino anomaly.

Evolution of the Reactor Antineutrino Flux and Spectrum at Daya Bay

NASA Astrophysics Data System (ADS)

An, F. P.; Balantekin, A. B.; Band, H. R.; Bishai, M.; Blyth, S.; Cao, D.; Cao, G. F.; Cao, J.; Chan, Y. L.; Chang, J. F.; Chang, Y.; Chen, H. S.; Chen, Q. Y.; Chen, S. M.; Chen, Y. X.; Chen, Y.; Cheng, J.; Cheng, Z. K.; Cherwinka, J. J.; Chu, M. C.; Chukanov, A.; Cummings, J. P.; Ding, Y. Y.; Diwan, M. V.; Dolgareva, M.; Dove, J.; Dwyer, D. A.; Edwards, W. R.; Gill, R.; Gonchar, M.; Gong, G. H.; Gong, H.; Grassi, M.; Gu, W. Q.; Guo, L.; Guo, X. H.; Guo, Y. H.; Guo, Z.; Hackenburg, R. W.; Hans, S.; He, M.; Heeger, K. M.; Heng, Y. K.; Higuera, A.; Hsiung, Y. B.; Hu, B. Z.; Hu, T.; Huang, E. C.; Huang, H. X.; Huang, X. T.; Huang, Y. B.; Huber, P.; Huo, W.; Hussain, G.; Jaffe, D. E.; Jen, K. L.; Ji, X. P.; Ji, X. L.; Jiao, J. B.; Johnson, R. A.; Jones, D.; Kang, L.; Kettell, S. H.; Khan, A.; Kohn, S.; Kramer, M.; Kwan, K. K.; Kwok, M. W.; Langford, T. J.; Lau, K.; Lebanowski, L.; Lee, J.; Lee, J. H. C.; Lei, R. T.; Leitner, R.; Leung, J. K. C.; Li, C.; Li, D. J.; Li, F.; Li, G. S.; Li, Q. J.; Li, S.; Li, S. C.; Li, W. D.; Li, X. N.; Li, X. Q.; Li, Y. F.; Li, Z. B.; Liang, H.; Lin, C. J.; Lin, G. L.; Lin, S.; Lin, S. K.; Lin, Y.-C.; Ling, J. J.; Link, J. M.; Littenberg, L.; Littlejohn, B. R.; Liu, J. L.; Liu, J. C.; Loh, C. W.; Lu, C.; Lu, H. Q.; Lu, J. S.; Luk, K. B.; Ma, X. Y.; Ma, X. B.; Ma, Y. Q.; Malyshkin, Y.; Martinez Caicedo, D. A.; McDonald, K. T.; McKeown, R. D.; Mitchell, I.; Nakajima, Y.; Napolitano, J.; Naumov, D.; Naumova, E.; Ngai, H. Y.; Ochoa-Ricoux, J. P.; Olshevskiy, A.; Pan, H.-R.; Park, J.; Patton, S.; Pec, V.; Peng, J. C.; Pinsky, L.; Pun, C. S. J.; Qi, F. Z.; Qi, M.; Qian, X.; Qiu, R. M.; Raper, N.; Ren, J.; Rosero, R.; Roskovec, B.; Ruan, X. C.; Steiner, H.; Stoler, P.; Sun, J. L.; Tang, W.; Taychenachev, D.; Treskov, K.; Tsang, K. V.; Tull, C. E.; Viaux, N.; Viren, B.; Vorobel, V.; Wang, C. H.; Wang, M.; Wang, N. Y.; Wang, R. G.; Wang, W.; Wang, X.; Wang, Y. F.; Wang, Z.; Wang, Z.; Wang, Z. M.; Wei, H. Y.; Wen, L. J.; Whisnant, K.; White, C. G.; Whitehead, L.; Wise, T.; Wong, H. L. H.; Wong, S. C. F.; Worcester, E.; Wu, C.-H.; Wu, Q.; Wu, W. J.; Xia, D. M.; Xia, J. K.; Xing, Z. Z.; Xu, J. L.; Xu, Y.; Xue, T.; Yang, C. G.; Yang, H.; Yang, L.; Yang, M. S.; Yang, M. T.; Yang, Y. Z.; Ye, M.; Ye, Z.; Yeh, M.; Young, B. L.; Yu, Z. Y.; Zeng, S.; Zhan, L.; Zhang, C.; Zhang, C. C.; Zhang, H. H.; Zhang, J. W.; Zhang, Q. M.; Zhang, R.; Zhang, X. T.; Zhang, Y. M.; Zhang, Y. X.; Zhang, Y. M.; Zhang, Z. J.; Zhang, Z. Y.; Zhang, Z. P.; Zhao, J.; Zhou, L.; Zhuang, H. L.; Zou, J. H.; Daya Bay Collaboration

2017-06-01

The Daya Bay experiment has observed correlations between reactor core fuel evolution and changes in the reactor antineutrino flux and energy spectrum. Four antineutrino detectors in two experimental halls were used to identify 2.2 million inverse beta decays (IBDs) over 1230 days spanning multiple fuel cycles for each of six 2.9 G Wth reactor cores at the Daya Bay and Ling Ao nuclear power plants. Using detector data spanning effective 239Pu fission fractions F239 from 0.25 to 0.35, Daya Bay measures an average IBD yield σ¯f of (5.90 ±0.13 )×10-43 cm2/fission and a fuel-dependent variation in the IBD yield, d σf/d F239, of (-1.86 ±0.18 )×10-43 cm2/fission . This observation rejects the hypothesis of a constant antineutrino flux as a function of the 239Pu fission fraction at 10 standard deviations. The variation in IBD yield is found to be energy dependent, rejecting the hypothesis of a constant antineutrino energy spectrum at 5.1 standard deviations. While measurements of the evolution in the IBD spectrum show general agreement with predictions from recent reactor models, the measured evolution in total IBD yield disagrees with recent predictions at 3.1 σ . This discrepancy indicates that an overall deficit in the measured flux with respect to predictions does not result from equal fractional deficits from the primary fission isotopes 235U, 239Pu, 238U, and 241Pu. Based on measured IBD yield variations, yields of (6.17 ±0.17 ) and (4.27 ±0.26 )×10-43 cm2 /fission have been determined for the two dominant fission parent isotopes 235U and 239Pu. A 7.8% discrepancy between the observed and predicted 235U yields suggests that this isotope may be the primary contributor to the reactor antineutrino anomaly.

Evolution of the Reactor Antineutrino Flux and Spectrum at Daya Bay

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

An, F. P.; Balantekin, A. B.; Band, H. R.

Here, the Daya Bay experiment has observed correlations between reactor core fuel evolution and changes in the reactor antineutrino flux and energy spectrum. Four antineutrino detectors in two experimental halls were used to identify 2.2 million inverse beta decays (IBDs) over 1230 days spanning multiple fuel cycles for each of six 2.9 GW th reactor cores at the Daya Bay and Ling Ao nuclear power plants. Using detector data spanning effective 239Pu fission fractions F 239 from 0.25 to 0.35, Daya Bay measures an average IBD yield ¯σf of (5.90±0.13)×10 –43 cm 2/fission and a fuel-dependent variation in the IBDmore » yield, dσ f/dF 239, of (–1.86±0.18)×10 –43 cm 2/fission. This observation rejects the hypothesis of a constant antineutrino flux as a function of the 239Pu fission fraction at 10 standard deviations. The variation in IBD yield is found to be energy dependent, rejecting the hypothesis of a constant antineutrino energy spectrum at 5.1 standard deviations. While measurements of the evolution in the IBD spectrum show general agreement with predictions from recent reactor models, the measured evolution in total IBD yield disagrees with recent predictions at 3.1σ. This discrepancy indicates that an overall deficit in the measured flux with respect to predictions does not result from equal fractional deficits from the primary fission isotopes 235U, 239Pu, 238U, and 241Pu. Based on measured IBD yield variations, yields of (6.17±0.17) and (4.27±0.26)×10 –43 cm 2/fission have been determined for the two dominant fission parent isotopes 235U and 239Pu. A 7.8% discrepancy between the observed and predicted 235U yields suggests that this isotope may be the primary contributor to the reactor antineutrino anomaly.« less

A Semi-Batch Reactor Experiment for the Undergraduate Laboratory

ERIC Educational Resources Information Center

Derevjanik, Mario; Badri, Solmaz; Barat, Robert

2011-01-01

This experiment and analysis offer an economic yet challenging semi-batch reactor experience. Household bleach is pumped at a controlled rate into a batch reactor containing pharmaceutical hydrogen peroxide solution. Batch temperature, product molecular oxygen, and the overall change in solution conductivity are metered. The reactor simulation…

A comparison of implantation-driven permeation characteristics of fusion reactor structural materials

NASA Astrophysics Data System (ADS)

Longhurst, G. R.; Anderl, R. A.; Struttmann, D. A.

1986-11-01

Implantation-driven permeation experiments have been conducted on samples of the ferritic steel HT-9, the austenitic Primary Candidate Alloy (PCA) and the vanadium alloy V-15Cr-5Ti using D 3+ ions under conditions that simulate charge-exchange neutral loading on a fusion reactor first wall. The steels all exhibited an initially intense permeation "spike" followed by an exponential decrease to low steady-state values. That spike was not evident in the V-15Cr-5Ti experiments. Steady-state permeation was highest in the vanadium alloy and lowest in the austenitic steel. Though permeation rates in the HT-9 were lower than those in V-15Cr-5Ti, permeation transients were much faster in HT-9 than in other materials tested. Sputtering of the steel surface resulted in enhanced reemission, whereas in the vanadium tests, recombination and diffusivity both appeared to diminish as the deuterium concentration rose. We conclude that for conditions comparable to those of these experiments, tritium retention and permeation loss in first wall structures made of steels will be less than in structures made of V-15Cr-5Ti.

10 CFR Appendix A to Part 110 - Illustrative List of Nuclear Reactor Equipment Under NRC Export Licensing Authority

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 2 2014-01-01 2014-01-01 false Illustrative List of Nuclear Reactor Equipment Under NRC... List of Nuclear Reactor Equipment Under NRC Export Licensing Authority Note: A nuclear reactor... core of a nuclear reactor and capable of withstanding the operating pressure of the primary coolant. (2...

Passive cooling system for top entry liquid metal cooled nuclear reactors

DOEpatents

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

1992-01-01

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

The reactor antineutrino anomaly and low energy threshold neutrino experiments

NASA Astrophysics Data System (ADS)

Cañas, B. C.; Garcés, E. A.; Miranda, O. G.; Parada, A.

2018-01-01

Short distance reactor antineutrino experiments measure an antineutrino spectrum a few percent lower than expected from theoretical predictions. In this work we study the potential of low energy threshold reactor experiments in the context of a light sterile neutrino signal. We discuss the perspectives of the recently detected coherent elastic neutrino-nucleus scattering in future reactor antineutrino experiments. We find that the expectations to improve the current constraints on the mixing with sterile neutrinos are promising. We also analyze the measurements of antineutrino scattering off electrons from short distance reactor experiments. In this case, the statistics is not competitive with inverse beta decay experiments, although future experiments might play a role when compare it with the Gallium anomaly.

10 CFR Appendix A to Part 110 - Illustrative List of Nuclear Reactor Equipment Under NRC Export Licensing Authority

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 2 2012-01-01 2012-01-01 false Illustrative List of Nuclear Reactor Equipment Under NRC... List of Nuclear Reactor Equipment Under NRC Export Licensing Authority Note—A nuclear reactor basically... nuclear reactor and capable of withstanding the operating pressure of the primary coolant. (2) On-line (e...

10 CFR Appendix A to Part 110 - Illustrative List of Nuclear Reactor Equipment Under NRC Export Licensing Authority

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 2 2013-01-01 2013-01-01 false Illustrative List of Nuclear Reactor Equipment Under NRC... List of Nuclear Reactor Equipment Under NRC Export Licensing Authority Note—A nuclear reactor basically... nuclear reactor and capable of withstanding the operating pressure of the primary coolant. (2) On-line (e...

Pressurized fluidized bed reactor and a method of operating the same

DOEpatents

Isaksson, J.

1996-02-20

A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine. 1 fig.

Pressurized fluidized bed reactor and a method of operating the same

DOEpatents

Isaksson, Juhani

1996-01-01

A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine.

Bulk Shielding Facility quarterly report, April, May and June 1984

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

Corbett, B.L.; Lance, E.D.

1984-12-01

The BSR operated at an average power level of 1310 kW for 3.8% of the time during April, May, and June. Water-quality control in both the reactor primary and secondary cooling systems was satisfactory. The PCA was used in training startups and was operated on five occasions for the NBS and HEDL recheck of a previous experiment run on the LWR pressure vessel surveillance dosimetry improvement program.

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.

Overview of Experiments for Physics of Fast Reactors from the International Handbooks of Evaluated Criticality Safety Benchmark Experiments and Evaluated Reactor Physics Benchmark Experiments

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

Bess, J. D.; Briggs, J. B.; Gulliford, J.

Overview of Experiments to Study the Physics of Fast Reactors Represented in the International Directories of Critical and Reactor Experiments John D. Bess Idaho National Laboratory Jim Gulliford, Tatiana Ivanova Nuclear Energy Agency of the Organisation for Economic Cooperation and Development E.V.Rozhikhin, M.Yu.Sem?nov, A.M.Tsibulya Institute of Physics and Power Engineering The study the physics of fast reactors traditionally used the experiments presented in the manual labor of the Working Group on Evaluation of sections CSEWG (ENDF-202) issued by the Brookhaven National Laboratory in 1974. This handbook presents simplified homogeneous model experiments with relevant experimental data, as amended. The Nuclear Energymore » Agency of the Organization for Economic Cooperation and Development coordinates the activities of two international projects on the collection, evaluation and documentation of experimental data - the International Project on the assessment of critical experiments (1994) and the International Project on the assessment of reactor experiments (since 2005). The result of the activities of these projects are replenished every year, an international directory of critical (ICSBEP Handbook) and reactor (IRPhEP Handbook) experiments. The handbooks present detailed models of experiments with minimal amendments. Such models are of particular interest in terms of the settlements modern programs. The directories contain a large number of experiments which are suitable for the study of physics of fast reactors. Many of these experiments were performed at specialized critical stands, such as BFS (Russia), ZPR and ZPPR (USA), the ZEBRA (UK) and the experimental reactor JOYO (Japan), FFTF (USA). Other experiments, such as compact metal assembly, is also of interest in terms of the physics of fast reactors, they have been carried out on the universal critical stands in Russian institutes (VNIITF and VNIIEF) and the US (LANL, LLNL, and others.). Also worth mentioning is the critical experiments with fast reactor fuel rods in water, interesting in terms of justification of nuclear safety during transportation and storage of fresh and spent fuel. These reports provide a detailed review of the experiment, designate the area of their application and include results of calculations on modern systems of constants in comparison with the estimated experimental data.« less

Experience with chemical system decontamination by the CORD process and electrochemical decontamination of pipe ends

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

Wille, H.; Bertholdt, H.O.; Operschall, H.

Efforts to reduce occupational radiation exposure during inspection and repair work in nuclear power plants turns steadily increasing attention to the decontamination of systems and components. Due to the advanced age of nuclear power plants resulting in increasing dose rates, the decontamination of components, or rather of complete systems, or loops to protect operating and inspection personnel becomes demanding. Besides, decontaminating complete primary loops is in many cases less difficult than cleaning large components. Based on experience gained in nuclear power plants, an outline of two different decontamination methods performed recently are given. For the decontamination of complete systems ormore » loops, Kraftwerk Union AG has developed CORD, a low-concentration process. For the decontamination performance of a subsystem, such as the steam generator (SG) channel heads of a pressurized water reactor or the recirculation loops of a boiling water reactor the automated mobile decontamination appliance is used. The electrochemical decontamination process is primarily applicable for the treatment of specially limited surface areas.« less

Key metrics for HFIR HEU and LEU models

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

Ilas, Germina; Betzler, Benjamin R.; Chandler, David

This report compares key metrics for two fuel design models of the High Flux Isotope Reactor (HFIR). The first model represents the highly enriched uranium (HEU) fuel currently in use at HFIR, and the second model considers a low-enriched uranium (LEU) interim design fuel. Except for the fuel region, the two models are consistent, and both include an experiment loading that is representative of HFIR's current operation. The considered key metrics are the neutron flux at the cold source moderator vessel, the mass of 252Cf produced in the flux trap target region as function of cycle time, the fast neutronmore » flux at locations of interest for material irradiation experiments, and the reactor cycle length. These key metrics are a small subset of the overall HFIR performance and safety metrics. They were defined as a means of capturing data essential for HFIR's primary missions, for use in optimization studies assessing the impact of HFIR's conversion from HEU fuel to different types of LEU fuel designs.« less

Neutrino Physics at Drexel

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

Lane, Charles; Dolinski, Michelle; Neilson, Russell

Our primary goal is to improve the understanding of the properties and interactions of neutrinos. We are pursuing this by means of the DUNE long-baseline and PROSPECT short-baseline neutrino experiments. For DUNE, a neutrino beam from Fermilab will be detected at the SURF facility in South Dakota, with the aim of determining the neutrino mass hierarchy (the mass ordering of neutrino flavors), and a measurement or limit on CP-violation via neutrinos. Our near-term experimental goal is to improve the characterization of the neutrino beam by measurements of muons produced as a byproduct of neutrino beam generation, to quantify the beammore » composition and flux. The short-range neutrino program has the aim of using the HFIR reactor at Oak Ridge as a neutrino source, with a detector placed nearby to find if there are short-distance oscillations to sterile neutrino flavors, and to resolve the 'reactor neutrino spectral anomaly' which has shown up as an unexplained 'bump' in the neutrino energy spectrum in recent experiments.« less

High temperature UF6 RF plasma experiments applicable to uranium plasma core reactors

NASA Technical Reports Server (NTRS)

Roman, W. C.

1979-01-01

An investigation was conducted using a 1.2 MW RF induction heater facility to aid in developing the technology necessary for designing a self critical fissioning uranium plasma core reactor. Pure, high temperature uranium hexafluoride (UF6) was injected into an argon fluid mechanically confined, steady state, RF heated plasma while employing different exhaust systems and diagnostic techniques to simulate and investigate some potential characteristics of uranium plasma core nuclear reactors. The development of techniques and equipment for fluid mechanical confinement of RF heated uranium plasmas with a high density of uranium vapor within the plasma, while simultaneously minimizing deposition of uranium and uranium compounds on the test chamber peripheral wall, endwall surfaces, and primary exhaust ducts, is discussed. The material tests and handling techniques suitable for use with high temperature, high pressure, gaseous UF6 are described and the development of complementary diagnostic instrumentation and measurement techniques to characterize the uranium plasma, effluent exhaust gases, and residue deposited on the test chamber and exhaust system components is reported.

NEUTRONIC REACTOR OPERATIONAL METHOD AND CORE SYSTEM

DOEpatents

Winters, C.E.; Graham, C.B.; Culver, J.S.; Wilson, R.H.

1960-07-19

Homogeneous neutronic reactor systems are described wherein an aqueous fuel solution is continuously circulated through a spherical core tank. The pumped fuel solution-is injected tangentially into the hollow spherical interior, thereby maintaining vigorous rotation of the solution within the tank in the form of a vortex; gaseous radiolytic decomposition products concentrate within the axial vortex cavity. The evolved gas is continuously discharged through a gas- outlet port registering with an extremity of the vortex cavity. and the solution stream is discharged through an annular liquid outlet port concentrically encircling the gas outlet by virtue of which the vortex and its cavity are maintained precisely axially aligned with the gas outlet. A primary heat exchanger extracts useful heat from the hot effluent fuel solution before its recirculation into the core tank. Hollow cylinders and other alternative core- tank configurations defining geometric volumes of revolution about a principal axis are also covered. AEC's Homogeneous Reactor Experiment No. 1 is a preferred embodiment.

DE-NE0008277_PROTEUS final technical report 2018

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

Enqvist, Andreas

This project details re-evaluations of experiments of gas-cooled fast reactor (GCFR) core designs performed in the 1970s at the PROTEUS reactor and create a series of International Reactor Physics Experiment Evaluation Project (IRPhEP) benchmarks. Currently there are no gas-cooled fast reactor (GCFR) experiments available in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). These experiments are excellent candidates for reanalysis and development of multiple benchmarks because these experiments provide high-quality integral nuclear data relevant to the validation and refinement of thorium, neptunium, uranium, plutonium, iron, and graphite cross sections. It would be cost prohibitive to reproduce suchmore » a comprehensive suite of experimental data to support any future GCFR endeavors.« less

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.

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

DOEpatents

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

1994-05-03

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

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

DOEpatents

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

1994-01-01

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

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.

Magnetic Flux Compression Reactor Concepts for Spacecraft Propulsion and Power (MSFC Center Director's Discretionary Fund; Project No. 99-24). Part 1

NASA Technical Reports Server (NTRS)

Litchford, R. J.; Robertson, G. A.; Hawk, C. W.; Turner, M. W.; Koelfgen, S.; Litchford, Ron J. (Technical Monitor)

2001-01-01

This technical publication (TP) examines performance and design issues associated with magnetic flux compression reactor concepts for nuclear/chemical pulse propulsion and power. Assuming that low-yield microfusion detonations or chemical detonations using high-energy density matter can eventually be realized in practice, various magnetic flux compression concepts are conceivable. In particular, reactors in which a magnetic field would be compressed between an expanding detonation-driven plasma cloud and a stationary structure formed from a high-temperature superconductor are envisioned. Primary interest is accomplishing two important functions: (1) Collimation and reflection of a hot diamagnetic plasma for direct thrust production, and (2) electric power generation for fusion standoff drivers and/or dense plasma formation. In this TP, performance potential is examined, major technical uncertainties related to this concept accessed, and a simple performance model for a radial-mode reactor developed. Flux trapping effectiveness is analyzed using a skin layer methodology, which accounts for magnetic diffusion losses into the plasma armature and the stationary stator. The results of laboratory-scale experiments on magnetic diffusion in bulk-processed type II superconductors are also presented.

The effect of mixing on fermentation of primary solids, glycerol, and biodiesel waste.

PubMed

Ghasemi, Marzieh; Randall, Andrew A

2018-03-01

In this study, the effect of mixing on volatile fatty acid (VFA) production and composition was investigated through running five identical bench-scale reactors that were filled with primary solid and dosed with either pure glycerol or biodiesel waste. Experimental results revealed that there was an inverse correlation between the mixing intensity and the VFA production. The total VFA production in the un-mixed reactor was 9,787 ± 3,601 mg COD/L, whereas in the reactor mixed at 100 rpm this dropped to 3,927 ± 1,175 mg COD/L, while both types of reactor were dosed with pure glycerol at the beginning of each cycle to reach the initial concentration of 1,000 mg/L (1,217 mg COD/L). Propionic acid was the dominant VFA in all the reactors except the reactor mixed at 30 rpm. It is hypothesized that low mixing facilitated hydrogen transfer between obligate hydrogen producing acetogens (OHPA) and hydrogen consuming acidogens in these non-methanogenic reactors. Also, in a narrower range of mixing (0 or 7 rpm), the total VFA production in biodiesel waste-fed reactors was considerably higher than that of pure glycerol-fed reactors.

Development of fission-products transport model in severe-accident scenarios for Scdap/Relap5

NASA Astrophysics Data System (ADS)

Honaiser, Eduardo Henrique Rangel

The understanding and estimation of the release of fission products during a severe accident became one of the priorities of the nuclear community after 1980, with the events of the Three-mile Island unit 2 (TMI-2), in 1979, and Chernobyl accidents, in 1986. Since this time, theoretical developments and experiments have shown that the primary circuit systems of light water reactors (LWR) have the potential to attenuate the release of fission products, a fact that had been neglected before. An advanced tool, compatible with nuclear thermal-hydraulics integral codes, is developed to predict the retention and physical evolution of the fission products in the primary circuit of LWRs, without considering the chemistry effects. The tool embodies the state-of-the-art models for the involved phenomena as well as develops new models. The capabilities acquired after the implementation of this tool in the Scdap/Relap5 code can be used to increase the accuracy of probability safety assessment (PSA) level 2, enhance the reactor accident management procedures and design new emergency safety features.

Fast reactor power plant design having heat pipe heat exchanger

DOEpatents

Huebotter, P.R.; McLennan, G.A.

1984-08-30

The invention relates to a pool-type fission reactor power plant design having a reactor vessel containing a primary coolant (such as liquid sodium), and a steam expansion device powered by a pressurized water/steam coolant system. Heat pipe means are disposed between the primary and water coolants to complete the heat transfer therebetween. The heat pipes are vertically oriented, penetrating the reactor deck and being directly submerged in the primary coolant. A U-tube or line passes through each heat pipe, extended over most of the length of the heat pipe and having its walls spaced from but closely proximate to and generally facing the surrounding walls of the heat pipe. The water/steam coolant loop includes each U-tube and the steam expansion device. A heat transfer medium (such as mercury) fills each of the heat pipes. The thermal energy from the primary coolant is transferred to the water coolant by isothermal evaporation-condensation of the heat transfer medium between the heat pipe and U-tube walls, the heat transfer medium moving within the heat pipe primarily transversely between these walls.

Fast reactor power plant design having heat pipe heat exchanger

DOEpatents

Huebotter, Paul R.; McLennan, George A.

1985-01-01

The invention relates to a pool-type fission reactor power plant design having a reactor vessel containing a primary coolant (such as liquid sodium), and a steam expansion device powered by a pressurized water/steam coolant system. Heat pipe means are disposed between the primary and water coolants to complete the heat transfer therebetween. The heat pipes are vertically oriented, penetrating the reactor deck and being directly submerged in the primary coolant. A U-tube or line passes through each heat pipe, extended over most of the length of the heat pipe and having its walls spaced from but closely proximate to and generally facing the surrounding walls of the heat pipe. The water/steam coolant loop includes each U-tube and the steam expansion device. A heat transfer medium (such as mercury) fills each of the heat pipes. The thermal energy from the primary coolant is transferred to the water coolant by isothermal evaporation-condensation of the heat transfer medium between the heat pipe and U-tube walls, the heat transfer medium moving within the heat pipe primarily transversely between these walls.

Preconceptual design of a fluoride high temperature salt-cooled engineering demonstration reactor: Motivation and overview

DOE PAGES

Qualls, A. Louis; Betzler, Benjamin R.; Brown, Nicholas R.; ...

2016-12-21

Engineering demonstration reactors are nuclear reactors built to establish proof of concept for technology options that have never been built. Examples of engineering demonstration reactors include Peach Bottom 1 for high temperature gas-cooled reactors (HTGRs) and Experimental Breeder Reactor-II (EBR-II) for sodium-cooled fast reactors. Historically, engineering demonstrations have played a vital role in advancing the technology readiness level of reactor technologies. Our paper details a preconceptual design for a fluoride salt-cooled engineering demonstration reactor. The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would usemore » tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 7LiF-BeF2) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is a necessary intermediate step to enable near-term commercial FHRs. The design philosophy of the FHR DR was focused on safety, near-term deployment, and flexibility. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated as an engineering demonstration with minimal risk and cost. These technologies include TRISO particle fuel, replaceable core structures, and consistent structural material selection for core structures and the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Important capabilities to be demonstrated by building and operating the FHR DR include fabrication and operation of high temperature reactors; heat exchanger performance (including passive decay heat removal); pump performance; and reactivity control; salt chemistry control to maximize vessel life; tritium management; core design methodologies; salt procurement, handling, maintenance and ultimate disposal. It is recognized that non-nuclear separate and integral test efforts (e.g., heated salt loops or loops using simulant fluids) are necessary to develop the technologies that will be demonstrated in the FHR DR.« less

Preconceptual design of a fluoride high temperature salt-cooled engineering demonstration reactor: Motivation and overview

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

Qualls, A. Louis; Betzler, Benjamin R.; Brown, Nicholas R.

Engineering demonstration reactors are nuclear reactors built to establish proof of concept for technology options that have never been built. Examples of engineering demonstration reactors include Peach Bottom 1 for high temperature gas-cooled reactors (HTGRs) and Experimental Breeder Reactor-II (EBR-II) for sodium-cooled fast reactors. Historically, engineering demonstrations have played a vital role in advancing the technology readiness level of reactor technologies. Our paper details a preconceptual design for a fluoride salt-cooled engineering demonstration reactor. The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would usemore » tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 7LiF-BeF2) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is a necessary intermediate step to enable near-term commercial FHRs. The design philosophy of the FHR DR was focused on safety, near-term deployment, and flexibility. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated as an engineering demonstration with minimal risk and cost. These technologies include TRISO particle fuel, replaceable core structures, and consistent structural material selection for core structures and the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Important capabilities to be demonstrated by building and operating the FHR DR include fabrication and operation of high temperature reactors; heat exchanger performance (including passive decay heat removal); pump performance; and reactivity control; salt chemistry control to maximize vessel life; tritium management; core design methodologies; salt procurement, handling, maintenance and ultimate disposal. It is recognized that non-nuclear separate and integral test efforts (e.g., heated salt loops or loops using simulant fluids) are necessary to develop the technologies that will be demonstrated in the FHR DR.« less

Primary system fission product release and transport: A state-of-the-art report to the committee on the safety of nuclear installations

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

Wright, A.L.

This report presents a summary of the status of research activities associated with fission product behavior (release and transport) under severe accident conditions within the primary systems of water-moderated and water-cooled nuclear reactors. For each of the areas of fission product release and fission product transport, the report summarizes relevant information on important phenomena, major experiments performed, relevant computer models and codes, comparisons of computer code calculations with experimental results, and general conclusions on the overall state of the art. Finally, the report provides an assessment of the overall importance and knowledge of primary system release and transport phenomena andmore » presents major conclusions on the state of the art.« less

Summary of the Workshop on Molten Salt Reactor Technologies Commemorating the 50th Anniversary of the Startup of the Molten Salt Reactor Experiment

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

Betzler, Benjamin R; Mays, Gary T

2016-01-01

A workshop on Molten Salt Reactor (MSR) technologies commemorating the 50th anniversary of the Molten Salt Reactor Experiment (MSRE) was held at Oak Ridge National Laboratory on October 15 16, 2015. The MSRE represented a pioneering experiment that demonstrated an advanced reactor technology: the molten salt eutectic-fueled reactor. A multinational group of more than 130 individuals representing a diverse set of stakeholders gathered to discuss the historical, current, and future technical challenges and paths to deployment of MSR technology. This paper provides a summary of the key messages from this workshop.

PBF Reactor Building (PER620). After lowering reactor vessel onto blocks, ...

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

PBF Reactor Building (PER-620). After lowering reactor vessel onto blocks, it is rolled on logs into PBF. Metal framework under vessel is handling device. Various penetrations in reactor bottom were for instrumentation, poison injection, drains. Large one, below center "manhole" was for primary coolant. Photographer: Larry Page. Date: February 13, 1970. INEEL negative no. 70-736 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

Nuclear Reactor Safety--The APS Submits its Report

ERIC Educational Resources Information Center

Physics Today, 1975

1975-01-01

Presents the summary section of the American Physical Society (APS) report on the safety features of the light-water reactor, reviews the design, construction, and operation of a reactor and outlines the primary engineered safety features. Summarizes the major recommendations of the study group. (GS)

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

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

Takaaki Sakai; Yasuhiro Enuma; Takashi Iwasaki

2002-07-01

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

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

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

Sakai, Takaaki; Enuma, Yasuhiro; Iwasaki, Takashi

2004-03-15

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

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.

Design principles of a simple and safe 200-MW(thermal) nuclear district heating plant

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

Goetzmann, C.; Bittermann, D.; Gobel, A.

Kraftwerk Union AG has almost completed the development of a dedicated 200-MW(thermal) nuclear district heating plant to provide environmentally clean energy at a predictably low cost. The concept can easily be adapted to meet power requirements within the 100- to 500-MW(thermal) range. This technology is the product of the experience gained with large pressurized water reactor and boiling water reactor power plants, with respect to both plant and fuel performance. The major development task is that of achieving sufficiently low capital cost by tailoring components and systems designed for large plants to the specific requirements of district heating. These requirementsmore » are small absolute power, low temperatures and pressures, and modest load following, all of which result in the characteristics that are summarized. A fully integrated primary system with natural circulation permits a very compact reactor building containing all safety-related systems and components. Plant safety is essentially guaranteed by inherent features. The reactor containment is tightly fitted around the reactor pressure vessel in such a way that, in the event of any postulated coolant leak, the core cannot become uncovered, even temporarily. Shutdown is assured by gravity drop of the control rods mounted above the core. Decay heat is removed from the core by means of natural circulation via dedicated intermediate circuits of external aircoolers.« less

Calculation to experiment comparison of SPND signals in various nuclear reactor environments

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

Barbot, Loic; Radulovic, Vladimir; Fourmentel, Damien

2015-07-01

In the perspective of irradiation experiments in the future Jules Horowitz Reactor (JHR), the Instrumentation Sensors and Dosimetry Laboratory of CEA Cadarache (France) is developing a numerical tool for SPND design, simulation and operation. In the frame of the SPND numerical tool qualification, dedicated experiments have been performed both in the Slovenian TRIGA Mark II reactor (JSI) and very recently in the French CEA Saclay OSIRIS reactor, as well as a test of two detectors in the core of the Polish MARIA reactor (NCBJ). A full description of experimental set-ups and neutron-gamma calculations schemes are provided in the first partmore » of the paper. Calculation to experiment comparison of the various SPNDs in the different reactors is thoroughly described and discussed in the second part. Presented comparisons show promising final results. (authors)« less

Overview of large scale experiments performed within the LBB project in the Czech Republic

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

Kadecka, P.; Lauerova, D.

1997-04-01

During several recent years NRI Rez has been performing the LBB analyses of safety significant primary circuit pipings of NPPs in Czech and Slovak Republics. The analyses covered the NPPs with reactors WWER 440 Type 230 and 213 and WWER 1000 Type 320. Within the relevant LBB projects undertaken with the aim to prove the fulfilling of the requirements of LBB, a series of large scale experiments were performed. The goal of these experiments was to verify the properties of the components selected, and to prove the quality and/or conservatism of assessments used in the LBB-analyses. In this poster, amore » brief overview of experiments performed in Czech Republic under guidance of NRI Rez is presented.« less

CRITICAL EXPERIMENT TANK (CET) REACTOR HAZARDS SUMMARY

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

Becar, N.J.; Kunze, J.F.; Pincock, G..D.

1961-03-31

The Critical Experiment Tank (CET) reactor assembly, the associated systems, and the Low Power Test Facility in which the reactor is to be operated are described. An evaluation and summary of the hazards associated with the operation of the CET reactor in the LPTF at the ldsho Test Station are also presented. (auth)

Heat exchanger with auxiliary cooling system

DOEpatents

Coleman, John H.

1980-01-01

A heat exchanger with an auxiliary cooling system capable of cooling a nuclear reactor should the normal cooling mechanism become inoperable. A cooling coil is disposed around vertical heat transfer tubes that carry secondary coolant therethrough and is located in a downward flow of primary coolant that passes in heat transfer relationship with both the cooling coil and the vertical heat transfer tubes. A third coolant is pumped through the cooling coil which absorbs heat from the primary coolant which increases the downward flow of the primary coolant thereby increasing the natural circulation of the primary coolant through the nuclear reactor.

Cometabolic degradation of ethyl mercaptan by phenol-utilizing Ralstonia eutropha in suspended growth and gas-recycling trickle-bed reactor.

PubMed

Sedighi, Mahsa; Zamir, Seyed Morteza; Vahabzadeh, Farzaneh

2016-01-01

The degradability of ethyl mercaptan (EM), by phenol-utilizing cells of Ralstonia eutropha, in both suspended and immobilized culture systems, was investigated in the present study. Free-cells experiments conducted at EM concentrations ranging from 1.25 to 14.42 mg/l, showed almost complete removal of EM at concentrations below 10.08 mg/l, which is much higher than the maximum biodegradable EM concentration obtained in experiments that did not utilize phenol as the primary substrate, i.e. 2.5 mg/l. The first-order kinetic rate constant (kSKS) for EM biodegradation by the phenol-utilizing cells (1.7 l/g biomass/h) was about 10 times higher than by cells without phenol utilization. Immobilized-cells experiments performed in a gas recycling trickle-bed reactor packed with kissiris particles at EM concentrations ranging from 1.6 to 36.9 mg/l, showed complete removal at all tested concentrations in a much shorter time, compared with free cells. The first-order kinetic rate constant (rmaxKs) for EM utilization was 0.04 l/h for the immobilized system compared to 0.06 for the suspended-growth culture, due to external mass transfer diffusion. Diffusion limitation was decreased by increasing the recycling-liquid flow rate from 25 to 65 ml/min. The removed EM was almost completely mineralized according to TOC and sulfate measurements. Shut down and starvation experiments revealed that the reactor could effectively handle the starving conditions and was reliable for full-scale application. Copyright © 2015 Elsevier Ltd. All rights reserved.

Passive cooling safety system for liquid metal cooled nuclear reactors

DOEpatents

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

1991-01-01

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

Control rod drive for reactor shutdown

DOEpatents

McKeehan, Ernest R.; Shawver, Bruce M.; Schiro, Donald J.; Taft, William E.

1976-01-20

A means for rapidly shutting down or scramming a nuclear reactor, such as a liquid metal-cooled fast breeder reactor, and serves as a backup to the primary shutdown system. The control rod drive consists basically of an in-core assembly, a drive shaft and seal assembly, and a control drive mechanism. The control rod is driven into the core region of the reactor by gravity and hydraulic pressure forces supplied by the reactor coolant, thus assuring that common mode failures will not interfere with or prohibit scramming the reactor when necessary.

Indirect passive cooling system for liquid metal cooled nuclear reactors

DOEpatents

Hunsbedt, Anstein; Boardman, Charles E.

1990-01-01

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

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

Toward a Mechanistic Source Term in Advanced Reactors: Characterization of Radionuclide Transport and Retention in a Sodium Cooled Fast Reactor

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

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

A vital component of the U.S. reactor licensing process is an integrated safety analysis in which a source term representing the release of radionuclides during normal operation and accident sequences is analyzed. Historically, source term analyses have utilized bounding, deterministic assumptions regarding radionuclide release. However, advancements in technical capabilities and the knowledge state have enabled the development of more realistic and best-estimate retention and release models such that a mechanistic source term assessment can be expected to be a required component of future licensing of advanced reactors. Recently, as part of a Regulatory Technology Development Plan effort for sodium cooledmore » fast reactors (SFRs), Argonne National Laboratory has investigated the current state of knowledge of potential source terms in an SFR via an extensive review of previous domestic experiments, accidents, and operation. As part of this work, the significant sources and transport processes of radionuclides in an SFR have been identified and characterized. This effort examines all stages of release and source term evolution, beginning with release from the fuel pin and ending with retention in containment. Radionuclide sources considered in this effort include releases originating both in-vessel (e.g. in-core fuel, primary sodium, cover gas cleanup system, etc.) and ex-vessel (e.g. spent fuel storage, handling, and movement). Releases resulting from a primary sodium fire are also considered as a potential source. For each release group, dominant transport phenomena are identified and qualitatively discussed. The key product of this effort was the development of concise, inclusive diagrams that illustrate the release and retention mechanisms at a high level, where unique schematics have been developed for in-vessel, ex-vessel and sodium fire releases. This review effort has also found that despite the substantial range of phenomena affecting radionuclide release, the current state of knowledge is extensive, and in most areas may be sufficient. Several knowledge gaps were identified, such as uncertainty in release from molten fuel and availability of thermodynamic data for lanthanides and actinides in liquid sodium. However, the overall findings suggest that high retention rates can be expected within the fuel and primary sodium for all radionuclides other than noble gases.« less

A bioassay experience and lessons learned on the internal contamination of (131)I during a maintenance period in a Korean nuclear power plant.

PubMed

Kim, Hee Geun; Kong, Tae Young

2012-08-01

During a maintenance period at a Korean nuclear power plant, internal exposure of radiation workers occurred by the inhalation of (131)I that was released into the reactor building from a primary system opening due to defective fuels. The internal activity in radiation workers contaminated by (131)I was immediately measured using a whole body counter (WBC). A whole body counting was performed again a few days later, considering the factors of equilibrium in the body. The intake and the committed effective dose were estimated based on the WBC results. The intake was also calculated by hand, based on both the entrance records to the reactor building, and the counted results of the air concentration for (131)I were compared with the whole body counting results.

Physics of reactor safety. Quarterly report, January--March 1977. [LMFBR

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

None

1977-06-01

This report summarizes work done on reactor safety, Monte Carlo analysis of safety-related critical assembly experiments, and planning of DEMI safety-related critical experiments. Work on reactor core thermal-hydraulics is also included.

The advantages and disadvantages of using the TREAT reactor for nuclear laser experiments

NASA Astrophysics Data System (ADS)

Dickson, P. W.; Snyder, A. M.; Imel, G. R.; McConnell, R. J.

The Transient Reactor Test Facility (TREAT) is a large air-cooled test facility located at the Idaho National Engineering Laboratory. Two of the major design features of TREAT, its large size and its being an air-cooled reactor, provide clues to both its advantages and disadvantages for supporting nuclear laser experiments. Its large size, which is dictated by the dilute uranium/graphite fuel, permits accommodation of geometrically large experiments. However, TREAT's large size also results in relatively long transients so that the energy deposited in an experiment is large relative to the peak power available from the reactor. TREAT's air-cooling mode of operation allows its configuration to be changed fairly readily. Due to air cooling, the reactor cools down slowly, permitting only one full power transient a day, which can be a disadvantage in some experimental programs. The reactor is capable of both steady-state or transient operation.

40 CFR 60.705 - Reporting and recordkeeping requirements.

Code of Federal Regulations, 2011 CFR

2011-07-01

... Volatile Organic Compound Emissions From Synthetic Organic Chemical Manufacturing Industry (SOCMI) Reactor... used or where the reactor process vent stream is introduced as the primary fuel to any size boiler or... equipment or reactors; (2) Any recalculation of the TRE index value performed pursuant to § 60.704(f); and...

40 CFR 60.705 - Reporting and recordkeeping requirements.

Code of Federal Regulations, 2010 CFR

2010-07-01

... Volatile Organic Compound Emissions From Synthetic Organic Chemical Manufacturing Industry (SOCMI) Reactor... used or where the reactor process vent stream is introduced as the primary fuel to any size boiler or... equipment or reactors; (2) Any recalculation of the TRE index value performed pursuant to § 60.704(f); and...

Operating manual for the High Flux Isotope Reactor. Volume I. Description of the facility

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

Not Available

1982-09-01

This volume contains a comprehensive description of the High Flux Isotope Reactor Facility. Its primary purpose is to supplement the detailed operating procedures, providing the reactor operators with background information on the various HFIR systems. The detailed operating procdures are presented in another report.

Non-equilibrium radiation nuclear reactor

NASA Technical Reports Server (NTRS)

Thom, K.; Schneider, R. T. (Inventor)

1978-01-01

An externally moderated thermal nuclear reactor is disclosed which is designed to provide output power in the form of electromagnetic radiation. The reactor is a gaseous fueled nuclear cavity reactor device which can operate over wide ranges of temperature and pressure, and which includes the capability of processing and recycling waste products such as long-lived transuranium actinides. The primary output of the device may be in the form of coherent radiation, so that the reactor may be utilized as a self-critical nuclear pumped laser.

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

Kupca, L.; Beno, P.

A very brief summary is provided of a primary circuit piping material properties analysis. The analysis was performed for the Bohunice V-1 reactor and the Kola-1 and -2 reactors. Assessment was performed on Bohunice V-1 archive materials and primary piping material cut from the Kola units after 100,000 hours of operation. Main research program tasks included analysis of mechanical properties, corrosion stability, and microstructural properties. Analysis results are not provided.

NUCLEAR POWER PLANTS

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

Bell, F.R.

1962-12-01

A power plant is described that comprises a nuclear reactor and a heat exchanger which is included in primary and secondary circuits. Fluid in the primary circuit extracts heat from the reactor and transfers it in the heat exchanger to the fluid in the secondary circuit which transmits energy to one or more utilization points. Means are provided for detecting, isolating, and removing radioactive fluid from the secondary circuit. (R.J.S.)

ETR HEAT EXCHANGER BUILDING, TRA644. A PRIMARY COOLANT PUMP AND ...

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

ETR HEAT EXCHANGER BUILDING, TRA-644. A PRIMARY COOLANT PUMP AND 24-INCH CHECK VALVE ARE MOUNTED IN A SHIELDED CUBICLE. NOTE CONNECTION AT RIGHT THROUGH SHIELD WALL TO PUMP MOTOR ON OTHER SIDE. INL NEGATIVE NO. 56-4177. Jack L. Anderson, Photographer, 12/21/1956 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

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.

New Reactor Physics Benchmark Data in the March 2012 Edition of the IRPhEP Handbook

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

John D. Bess; J. Blair Briggs; Jim Gulliford

2012-11-01

The International Reactor Physics Experiment Evaluation Project (IRPhEP) was established to preserve integral reactor physics experimental data, including separate or special effects data for nuclear energy and technology applications. Numerous experiments that have been performed worldwide, represent a large investment of infrastructure, expertise, and cost, and are valuable resources of data for present and future research. These valuable assets provide the basis for recording, development, and validation of methods. If the experimental data are lost, the high cost to repeat many of these measurements may be prohibitive. The purpose of the IRPhEP is to provide an extensively peer-reviewed set ofmore » reactor physics-related integral data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next-generation reactors and establish the safety basis for operation of these reactors. Contributors from around the world collaborate in the evaluation and review of selected benchmark experiments for inclusion in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook) [1]. Several new evaluations have been prepared for inclusion in the March 2012 edition of the IRPhEP Handbook.« less

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

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.

IET. Typical detail during Snaptran reactor experiments. Shielding bricks protect ...

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

IET. Typical detail during Snaptran reactor experiments. Shielding bricks protect ion chamber beneath reactor on dolly. Photographer: Page Comiskey. Date: August 11, 1965. INEEL negative no. 65-4039 - Idaho National Engineering Laboratory, Test Area North, Scoville, Butte County, ID

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

Brown, Nicholas R.; Powers, Jeffrey J.; Mueller, Don

In September 2016, reactor physics measurements were conducted at Research Centre Rez (RC Rez) using the FLiBe (2 7LiF + BeF 2) salt from the Molten Salt Reactor Experiment (MSRE) in the LR-0 low power nuclear reactor. These experiments were intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems using FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL), in collaboration with RC Rez, performed sensitivity/uncertainty (S/U) analyses of these experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy researchmore » and development. The objectives of these analyses were (1) to identify potential sources of bias in fluoride salt-cooled and salt-fueled reactor simulations resulting from cross section uncertainties, and (2) to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a final report on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. In the future, these S/U analyses could be used to inform the design of additional FLiBe-based experiments using the salt from MSRE. The key finding of this work is that, for both solid and liquid fueled fluoride salt reactors, radiative capture in 7Li is the most significant contributor to potential bias in neutronics calculations within the FLiBe salt.« less

Future Reactor Neutrino Experiments (RRNOLD)1

NASA Astrophysics Data System (ADS)

Jaffe, David E.

The prospects for future reactor neutrino experiments that would use tens of kilotons of liquid scintillator with a ∼ 50 km baseline are discussed. These experiments are generically dubbed "RRNOLD" for Radical Reactor Neutrino Oscillation Liquid scintillator Detector experiment. Such experiments are designed to resolve the neutrino mass hierarchy and make sub-percent measurements sin2θ12, Δm232 and Δm122 . RRNOLD would also be sensitive to neutrinos from other sources and have notable sensitivity to proton decay.

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

Wichman, K.; Tsao, J.; Mayfield, M.

The regulatory application of leak before break (LBB) for operating and advanced reactors in the U.S. is described. The U.S. Nuclear Regulatory Commission (NRC) has approved the application of LBB for six piping systems in operating reactors: reactor coolant system primary loop piping, pressurizer surge, safety injection accumulator, residual heat removal, safety injection, and reactor coolant loop bypass. The LBB concept has also been applied in the design of advanced light water reactors. LBB applications, and regulatory considerations, for pressurized water reactors and advanced light water reactors are summarized in this paper. Technology development for LBB performed by the NRCmore » and the International Piping Integrity Research Group is also briefly summarized.« less

Solution of heat removal from nuclear reactors by natural convection

NASA Astrophysics Data System (ADS)

Zitek, Pavel; Valenta, Vaclav

2014-03-01

This paper summarizes the basis for the solution of heat removal by natural convection from both conventional nuclear reactors and reactors with fuel flowing coolant (such as reactors with molten fluoride salts MSR).The possibility of intensification of heat removal through gas lift is focused on. It might be used in an MSR (Molten Salt Reactor) for cleaning the salt mixture of degassed fission products and therefore eliminating problems with iodine pitting. Heat removal by natural convection and its intensification increases significantly the safety of nuclear reactors. Simultaneously the heat removal also solves problems with lifetime of pumps in the primary circuit of high-temperature reactors.

Method for removing cesium from a nuclear reactor coolant

DOEpatents

Colburn, Richard P.

1986-01-01

A method of and system for removing cesium from a liquid metal reactor coolant including a carbon packing trap in the primary coolant system for absorbing a major portion of the radioactive cesium from the coolant flowing therethrough at a reduced temperature. A regeneration subloop system having a secondary carbon packing trap is selectively connected to the primary system for isolating the main trap therefrom and connecting it to the regeneration system. Increasing the temperature of the sodium flowing through the primary trap diffuses a portion of the cesium

Neutrino scattering and the reactor antineutrino anomaly

NASA Astrophysics Data System (ADS)

Garcés, Estela; Cañas, Blanca; Miranda, Omar; Parada, Alexander

2017-12-01

Low energy threshold reactor experiments have the potential to give insight into the light sterile neutrino signal provided by the reactor antineutrino anomaly and the gallium anomaly. In this work we analyze short baseline reactor experiments that detect by elastic neutrino electron scattering in the context of a light sterile neutrino signal. We also analyze the sensitivity of experimental proposals of coherent elastic neutrino nucleus scattering (CENNS) detectors in order to exclude or confirm the sterile neutrino signal with reactor antineutrinos.

Updated Global Analysis of Neutrino Oscillations in the Presence of eV-Scale Sterile Neutrinos

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

Dentler, Mona; Hernández-Cabezudo, Alvaro; Kopp, Joachim

We discuss the possibility to explain the anomalies in short-baseline neutrino oscillation experiments in terms of sterile neutrinos. We work in a 3+1 framework and pay special attention to recent new data from reactor experiments, IceCube and MINOS+. We find that results from the DANSS and NEOS reactor experiments support the sterile neutrino explanation of the reactor anomaly, based on an analysis that relies solely on the relative comparison of measured reactor spectra. Global data from themore » $$\

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.

Study of Compatibility of Stainless Steel Weld Joints with Liquid Sodium-Potassium Coolants for Fission Surface Power Reactors for Lunar and Space Applications

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

Grossbeck, Martin; Qualls, Louis

To make a manned mission to the surface of the moon or to Mars with any significant residence time, the power requirements will make a nuclear reactor the most feasible source of energy. To prepare for such a mission, NASA has teamed with the DOE to develop Fission Surface Power technology with the goal of developing viable options. The Fission Surface Power System (FSPS) recommended as the initial baseline design includes a liquid metal reactor and primary coolant system that transfers heat to two intermediate liquid metal heat transfer loops. Each intermediate loop transfers heat to two Stirling heat exchangersmore » that each power two Stirling converters. Both the primary and the intermediate loops will use sodium-potassium (NaK) as the liquid metal coolant, and the primary loop will operate at temperatures exceeding 600°C. The alloy selected for the heat exchangers and piping is AISI Type 316L stainless steel. The extensive experience with NaK in breeder reactor programs and with earlier space reactors for unmanned missions lends considerable confidence in using NaK as a coolant in contact with stainless steel alloys. However, the microstructure, chemical segregation, and stress state of a weld leads to the potential for corrosion and cracking. Such failures have been experienced in NaK systems that have operated for times less than the eight year goal for the FSPS. For this reason, it was necessary to evaluate candidate weld techniques and expose welds to high-temperature, flowing NaK in a closed, closely controlled system. The goal of this project was to determine the optimum weld configuration for a NaK system that will withstand service for eight years under FSPS conditions. Since the most difficult weld to make and to evaluate is the tube to tube sheet weld in the intermediate heat exchangers, it was the focus of this research. A pumped loop of flowing NaK was fabricated for exposure of candidate weld specimens at temperatures of 600°C, the expected temperature within the intermediate heat exchangers. Since metal transfer from a high-temperature region to a cooler region is a predominant mode of corrosion in liquid metal systems, specimens were placed at zones in the loop at the above temperature to evaluate the effects of both alloy component leaching and metal deposition. Microstructural analysis was performed to evaluate weld performance on control weld specimens. The research was coordinated with Oak Ridge National Laboratory (ORNL) where most of the weld samples were prepared. In addition, ORNL participated in the loop operation to assist in keeping the testing relevant to the project and to take advantage of the extensive experience in liquid metal research at ORNL.« less

Application of Radiation Chemistry to Some Selected Technological Issues Related to the Development of Nuclear Energy.

PubMed

Bobrowski, Krzysztof; Skotnicki, Konrad; Szreder, Tomasz

2016-10-01

The most important contributions of radiation chemistry to some selected technological issues related to water-cooled reactors, reprocessing of spent nuclear fuel and high-level radioactive wastes, and fuel evolution during final radioactive waste disposal are highlighted. Chemical reactions occurring at the operating temperatures and pressures of reactors and involving primary transients and stable products from water radiolysis are presented and discussed in terms of the kinetic parameters and radiation chemical yields. The knowledge of these parameters is essential since they serve as input data to the models of water radiolysis in the primary loop of light water reactors and super critical water reactors. Selected features of water radiolysis in heterogeneous systems, such as aqueous nanoparticle suspensions and slurries, ceramic oxides surfaces, nanoporous, and cement-based materials, are discussed. They are of particular concern in the primary cooling loops in nuclear reactors and long-term storage of nuclear waste in geological repositories. This also includes radiation-induced processes related to corrosion of cladding materials and copper-coated iron canisters, dissolution of spent nuclear fuel, and changes of bentonite clays properties. Radiation-induced processes affecting stability of solvents and solvent extraction ligands as well oxidation states of actinide metal ions during recycling of the spent nuclear fuel are also briefly summarized.

Reliability Analysis of RSG-GAS Primary Cooling System to Support Aging Management Program

NASA Astrophysics Data System (ADS)

Deswandri; Subekti, M.; Sunaryo, Geni Rina

2018-02-01

Multipurpose Research Reactor G.A. Siwabessy (RSG-GAS) which has been operating since 1987 is one of the main facilities on supporting research, development and application of nuclear energy programs in BATAN. Until now, the RSG-GAS research reactor has been successfully operated safely and securely. However, because it has been operating for nearly 30 years, the structures, systems and components (SSCs) from the reactor would have started experiencing an aging phase. The process of aging certainly causes a decrease in reliability and safe performances of the reactor, therefore the aging management program is needed to resolve the issues. One of the programs in the aging management is to evaluate the safety and reliability of the system and also screening the critical components to be managed.One method that can be used for such purposes is the Fault Tree Analysis (FTA). In this papers FTA method is used to screening the critical components in the RSG-GAS Primary Cooling System. The evaluation results showed that the primary isolation valves are the basic events which are dominant against the system failure.

139. ARAIII Index of drwaings of gascooled reactor experiment buildings. ...

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

139. ARA-III Index of drwaings of gas-cooled reactor experiment buildings. Aerojet-general 880-area/GCRE-100. Date: February 1958. Ineel index code no. 063-9999-80-013-102505. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

Schedule and status of irradiation experiments

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

Rowcliffe, A.F.; Grossbeck, M.L.; Robertson, J.P.

1998-09-01

The current status of reactor irradiation experiments is presented in tables summarizing the experimental objectives, conditions, and schedule. Currently, the program has one irradiation experiment in reactor and five experiments in the design or construction stages. Postirradiation examination and testing is in progress on ten experiments.

Review of Transient Testing of Fast Reactor Fuels in the Transient REActor Test Facility (TREAT)

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

Jensen, C.; Wachs, D.; Carmack, J.

The restart of the Transient REActor Test (TREAT) facility provides a unique opportunity to engage the fast reactor fuels community to reinitiate in-pile experimental safety studies. Historically, the TREAT facility played a critical role in characterizing the behavior of both metal and oxide fast reactor fuels under off-normal conditions, irradiating hundreds of fuel pins to support fast reactor fuel development programs. The resulting test data has provided validation for a multitude of fuel performance and severe accident analysis computer codes. This paper will provide a review of the historical database of TREAT experiments including experiment design, instrumentation, test objectives, andmore » salient findings. Additionally, the paper will provide an introduction to the current and future experiment plans of the U.S. transient testing program at TREAT.« less

TREAT Reactor Control and Protection System

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

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

1985-01-01

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

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

Feng Xie; Hong Li; Jianzhu Cao

A reform will be implemented in the helium purification system of the 10 MW High Temperature Gas-cooled Test Reactor (HTR-10) in China. The measurement of the gamma dose rates of facilities, including valves, pipes, dust filter, etc., in the purification system of the HTR-10, has been performed. The results indicated that most radiation nuclides are concentrated in the dust filter and facilities at the entrance of the helium purification system upstream of the dust filter. Other facilities have the same gamma dose rate level as the background. Based on the previous study and experiences in AVR, the measurement results canmore » be understood that the radioactive dust carried by the helium gas was filtered by the dust filter. It provides important insights for the decontamination and decommissioning of facilities in the primary loop, especially in the helium purification system of the HTR-10 as well as the High Temperature Reactor-Pebble bed Modules (HTR-PM). (authors)« less

Double Chooz and a history of reactor θ 13 experiments

DOE PAGES

Suekane, Fumihiko; Junqueira de Castro Bezerra, Thiago

2016-04-11

This is a contribution paper from the Double Chooz (DC) experiment to the special issue of Nuclear Physics B on the topics of neutrino oscillations, celebrating the recent Nobel prize to Profs. T. Kajita and A.B. McDonald. DC is a reactor neutrino experiment which measures the last neutrino mixing angle θ 13. In addition, the DC group presented an indication of disappearance of the reactor neutrinos at a baseline of similar to 1 km for the first time in 2011 and is improving the measurement of θ 13. DC is a pioneering experiment of this research field. In accordance withmore » the nature of this special issue, physics and history of the reactor-θ 13 experiments, as well as the Double Chooz experiment and its neutrino oscillation analyses, are reviewed.« less

Double Chooz and a history of reactor θ 13 experiments

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

Suekane, Fumihiko; Junqueira de Castro Bezerra, Thiago

This is a contribution paper from the Double Chooz (DC) experiment to the special issue of Nuclear Physics B on the topics of neutrino oscillations, celebrating the recent Nobel prize to Profs. T. Kajita and A.B. McDonald. DC is a reactor neutrino experiment which measures the last neutrino mixing angle θ 13. In addition, the DC group presented an indication of disappearance of the reactor neutrinos at a baseline of similar to 1 km for the first time in 2011 and is improving the measurement of θ 13. DC is a pioneering experiment of this research field. In accordance withmore » the nature of this special issue, physics and history of the reactor-θ 13 experiments, as well as the Double Chooz experiment and its neutrino oscillation analyses, are reviewed.« less

Latest progress from the Daya Bay reactor neutrino experiment

NASA Astrophysics Data System (ADS)

Wang, Zhe; Daya Bay Collaboration

2016-05-01

Recently the Daya Bay reactor neutrino experiment has presented several new results about neutrino and reactor physics after acquiring a large data sample and after gaining a more sophisticated understanding of the experiment. In this talk I will introduce the latest progress made by the experiment including a three-flavor neutrino oscillation analysis using neutron capture on gadolinium, which gave sin2 2θ 13 = 0.084 ± 0.005 and |Δm2 ee| = (2.42 ±0.11) × 10-3 eV2, an independent θ 13 measurement using neutron capture on hydrogen, a search for a light sterile neutrino, and a measurement of the reactor antineutrino flux and spectrum.

Reactor transient control in support of PFR/TREAT TUCOP experiments

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

Burrows, D.R.; Larsen, G.R.; Harrison, L.J.

1984-01-01

Unique energy deposition and experiment control requirements posed bythe PFR/TREAT series of transient undercooling/overpower (TUCOP) experiments resulted in equally unique TREAT reactor operations. New reactor control computer algorithms were written and used with the TREAT reactor control computer system to perform such functions as early power burst generation (based on test train flow conditions), burst generation produced by a step insertion of reactivity following a controlled power ramp, and shutdown (SCRAM) initiators based on both test train conditions and energy deposition. Specialized hardware was constructed to simulate test train inputs to the control computer system so that computer algorithms couldmore » be tested in real time without irradiating the experiment.« less

STATUS OF TRISO FUEL IRRADIATIONS IN THE ADVANCED TEST REACTOR SUPPORTING HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGNS

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

Davenport, Michael; Petti, D. A.; Palmer, Joe

2016-11-01

The United States Department of Energy’s Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experimentsmore » are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and completed in October 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and completed in April 2014. Since the purpose of this experiment was to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment was significantly different from the first two experiments, though the control and monitoring systems are very similar. The final experiment, AGR-5/6/7, is scheduled to begin irradiation in early summer 2017.« less

FOEHN: The critical experiment for the Franco-German High Flux Reactor

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

Scharmer, K.; Eckert, H. G.

1991-01-01

A critical experiment for the Franco-German High Flux Reactor was carried out in the French reactor EOLE (CEN Cadarache). The purpose of the experiment was to check the calculation methods in a realistic geometry and to measure effects that can only be calculated imprecisely (e.g. beam hole effects). The structure of the experiment and the measurement and calculation methods are described. A detailed comparison between theoretical and experimental results was performed. 30 refs., 105 figs.

Impact of VOC Composition and Reactor Conditions on the Aging of Biomass Cookstove Emission in an Oxidation Flow Reactor

EPA Science Inventory

Oxidation flow reactor (OFR) experiments in our lab have explored secondary organic aerosol (SOA) production during photochemical aging of emissions from cookstoves used by billions in developing countries. Previous experiments, conducted with red oak fuel under conditions of hig...

Impact of VOC Composition and Reactor Conditions on the Aging of Biomass Cookstove Emissions in an Oxidation Flow Reactor

EPA Science Inventory

Oxidation flow reactor (OFR) experiments in our lab have explored secondary organic aerosol (SOA) production during photochemical aging of emissions from cookstoves used by billions in developing countries. Previous experiments, conducted with red oak fuel under conditions of hig...

Discontinuous Mode Power Supply

NASA Technical Reports Server (NTRS)

Lagadinos, John; Poulos, Ethel

2012-01-01

A document discusses the changes made to a standard push-pull inverter circuit to avoid saturation effects in the main inverter power supply. Typically, in a standard push-pull arrangement, the unsymmetrical primary excitation causes variations in the volt second integral of each half of the excitation cycle that could lead to the establishment of DC flux density in the magnetic core, which could eventually cause saturation of the main inverter transformer. The relocation of the filter reactor normally placed across the output of the power supply solves this problem. The filter reactor was placed in series with the primary circuit of the main inverter transformer, and is presented as impedance against the sudden changes on the input current. The reactor averaged the input current in the primary circuit, avoiding saturation of the main inverter transformer. Since the implementation of the described change, the above problem has not reoccurred, and failures in the main power transistors have been avoided.

Neutrino Physics with Nuclear Reactors: An Overview

NASA Astrophysics Data System (ADS)

Ochoa-Ricoux, J. P.

Nuclear reactors provide an excellent environment for studying neutrinos and continue to play a critical role in unveiling the secrets of these elusive particles. A rich experimental program with reactor antineutrinos is currently ongoing, and leads the way in precision measurements of several oscillation parameters and in searching for new physics, such as the existence of light sterile neutrinos. Ongoing experiments have also been able to measure the flux and spectral shape of reactor antineutrinos with unprecedented statistics and as a function of core fuel evolution, uncovering anomalies that will lead to new physics and/or to an improved understanding of antineutrino emission from nuclear reactors. The future looks bright, with an aggressive program of next generation reactor neutrino experiments that will go after some of the biggest open questions in the field. This includes the JUNO experiment, the largest liquid scintillator detector ever constructed which will push the limits of this detection technology.

Advanced gray rod control assembly

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

Drudy, Keith J; Carlson, William R; Conner, Michael E

An advanced gray rod control assembly (GRCA) for a nuclear reactor. The GRCA provides controlled insertion of gray rod assemblies into the reactor, thereby controlling the rate of power produced by the reactor and providing reactivity control at full power. Each gray rod assembly includes an elongated tubular member, a primary neutron-absorber disposed within the tubular member said neutron-absorber comprising an absorber material, preferably tungsten, having a 2200 m/s neutron absorption microscopic capture cross-section of from 10 to 30 barns. An internal support tube can be positioned between the primary absorber and the tubular member as a secondary absorber tomore » enhance neutron absorption, absorber depletion, assembly weight, and assembly heat transfer characteristics.« less

Advanced Test Reactor Tour

ScienceCinema

Miley, Don

2017-12-21

The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored.

Research Program of a Super Fast Reactor

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

Oka, Yoshiaki; Ishiwatari, Yuki; Liu, Jie

2006-07-01

Research program of a supercritical-pressure light water cooled fast reactor (Super Fast Reactor) is funded by MEXT (Ministry of Education, Culture, Sports, Science and Technology) in December 2005 as one of the research programs of Japanese NERI (Nuclear Energy Research Initiative). It consists of three programs. (1) development of Super Fast Reactor concept; (2) thermal-hydraulic experiments; (3) material developments. The purpose of the concept development is to pursue the advantage of high power density of fast reactor over thermal reactors to achieve economic competitiveness of fast reactor for its deployment without waiting for exhausting uranium resources. Design goal is notmore » breeding, but maximizing reactor power by using plutonium from spent LWR fuel. MOX will be the fuel of the Super Fast Reactor. Thermal-hydraulic experiments will be conducted with HCFC22 (Hydro chlorofluorocarbons) heat transfer loop of Kyushu University and supercritical water loop at JAEA. Heat transfer data including effect of grid spacers will be taken. The critical flow and condensation of supercritical fluid will be studied. The materials research includes the development and testing of austenitic stainless steel cladding from the experience of PNC1520 for LMFBR. Material for thermal insulation will be tested. SCWR (Supercritical-Water Cooled Reactor) of GIF (Generation-4 International Forum) includes both thermal and fast reactors. The research of the Super Fast Reactor will enhance SCWR research and the data base. The research period will be until March 2010. (authors)« less

Buoyancy Driven Coolant Mixing Studies of Natural Circulation Flows at the ROCOM Test Facility Using ANSYS CFX

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

Hohne, Thomas; Kliem, Soren; Rohde, Ulrich

2006-07-01

Coolant mixing in the cold leg, downcomer and the lower plenum of pressurized water reactors is an important phenomenon mitigating the reactivity insertion into the core. Therefore, mixing of the de-borated slugs with the ambient coolant in the reactor pressure vessel was investigated at the four loop 1:5 scaled ROCOM mixing test facility. Thermal hydraulics analyses showed, that weakly borated condensate can accumulate in particular in the pump loop seal of those loops, which do not receive safety injection. After refilling of the primary circuit, natural circulation in the stagnant loops can re-establish simultaneously and the de-borated slugs are shiftedmore » towards the reactor pressure vessel (RPV). In the ROCOM experiments, the length of the flow ramp and the initial density difference between the slugs and the ambient coolant was varied. From the test matrix experiments with 0 resp. 2% density difference between the de-borated slugs and the ambient coolant were used to validate the CFD software ANSYS CFX. To model the effects of turbulence on the mean flow a higher order Reynolds stress turbulence model was employed and a mesh consisting of 6.4 million hybrid elements was utilized. Only the experiments and CFD calculations with modeled density differences show a stratification in the downcomer. Depending on the degree of density differences the less dense slugs flow around the core barrel at the top of the downcomer. At the opposite side the lower borated coolant is entrained by the colder safety injection water and transported to the core. The validation proves that ANSYS CFX is able to simulate appropriately the flow field and mixing effects of coolant with different densities. (authors)« less

Background studies for the MINER Coherent Neutrino Scattering reactor experiment

NASA Astrophysics Data System (ADS)

Agnolet, G.; Baker, W.; Barker, D.; Beck, R.; Carroll, T. J.; Cesar, J.; Cushman, P.; Dent, J. B.; De Rijck, S.; Dutta, B.; Flanagan, W.; Fritts, M.; Gao, Y.; Harris, H. R.; Hays, C. C.; Iyer, V.; Jastram, A.; Kadribasic, F.; Kennedy, A.; Kubik, A.; Lang, K.; Mahapatra, R.; Mandic, V.; Marianno, C.; Martin, R. D.; Mast, N.; McDeavitt, S.; Mirabolfathi, N.; Mohanty, B.; Nakajima, K.; Newhouse, J.; Newstead, J. L.; Ogawa, I.; Phan, D.; Proga, M.; Rajput, A.; Roberts, A.; Rogachev, G.; Salazar, R.; Sander, J.; Senapati, K.; Shimada, M.; Soubasis, B.; Strigari, L.; Tamagawa, Y.; Teizer, W.; Vermaak, J. I. C.; Villano, A. N.; Walker, J.; Webb, B.; Wetzel, Z.; Yadavalli, S. A.

2017-05-01

The proposed Mitchell Institute Neutrino Experiment at Reactor (MINER) experiment at the Nuclear Science Center at Texas A&M University will search for coherent elastic neutrino-nucleus scattering within close proximity (about 2 m) of a 1 MW TRIGA nuclear reactor core using low threshold, cryogenic germanium and silicon detectors. Given the Standard Model cross section of the scattering process and the proposed experimental proximity to the reactor, as many as 5-20 events/kg/day are expected. We discuss the status of preliminary measurements to characterize the main backgrounds for the proposed experiment. Both in situ measurements at the experimental site and simulations using the MCNP and GEANT4 codes are described. A strategy for monitoring backgrounds during data taking is briefly discussed.

Phenomenology and modeling of particulate corrosion product behavior in Hanford N Reactor primary coolant

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

Bechtold, D.B.

1983-12-31

The levels and composition of filterable corrosion products in the Hanford N Reactor Primary Loop are measurable by filtration. The suspended crud level has ranged from 0.0005 ppM to 6.482 ppM with a median 0.050 ppM. The composition approximates magnetite. The particle size distribution has been found in 31 cases to be uniformly a log normal distribution with a count median ranging from 1.10 to 2.31 microns with a median of 1.81 microns, and the geometric standard deviation ranging from 1.60 to 2.34 with a median of 1.84. An auto-correcting inline turbidimeter was found to respond to linearly to suspendedmore » crud levels over a range 0.05 to at least 6.5 ppM by direct comparison with filter sample weights. Cause of crud bursts in the primary loop were found to be power decreases. The crud transients associated with a reactor power drop, several reactor shutdowns, and several reactor startups could be modeled consistently with each other using a simple stirred-tank, first order exchange model of particulate between makeup, coolant, letdown, and loosely adherent crud on pipe walls. Over 3/10 of the average steady running particulate crud level could be accounted for by magnetically filterable particulate in the makeup feed. A simulation model of particulate transport has been coded in FORTRAN.« less

Thermos reactors

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

Labrousse, M.; Lerouge, B.; Dupuy, G.

1978-04-01

THERMOS is a water reactor designed to provide hot water up to 120/sup 0/C for district heating or for desalination applications. It is a 100-MW reactor based on proven technology: oxide fuel plate elements, integrated primary circuit, and reactor vessel located in the bottom of a pool. As in swimming pool reactors, the pool is used for biological shielding, emergency core cooling, and fission product filtering (in case of an accident). Before economics, safety is the main characteristic of the concept: no fuel failure admitted, core under water in any accidental configuration, inspection of every ''nuclear'' component, and double-wall containment.

[Research of input water ratio's impact on the quality of effluent water from hydrolysis reactor].

PubMed

Liang, Kang-Qiang; Xiong, Ya; Qi, Mao-Rong; Lin, Xiu-Jun; Zhu, Min; Song, Ying-Hao

2012-11-01

Based on high SS/BOD and low C/N ratio of waste water of municipal wastewater treatment plant, the structure of currently existing hydrolysis reactor was reformed to improve the influent quality. In order to strengthen the sludge hydrolysis and improve effluent water quality, two layers water distributors were set up so that the sludge hydrolysis zone was formed between the two layers distribution. For the purpose of the hydrolysis reactor not only plays the role of the primary sedimentation tank but also improves the effluent water biodegradability, input water ratios of the upper and lower water distributor in the experiment were changed to get the best input water ratio to guide the large-scale application of this sort hydrolysis reactor. Results show, four kinds of input water ratio have varying degrees COD and SS removal efficiency, however, input water ratio for 1 : 1 can substantially increase SCOD/COD ratio and VFA concentration of effluent water compared with the other three input water ratios. To improve the effluent biodegradability, input water ratio for 1 : 1 was chosen for the best input water ratio. That was the ratio of flow of upper distributor was 50%, and the ratio of the lower one was 50%, at this case it can reduce the processing burden of COD and SS for follow-up treatment, but also improve the biodegradability of the effluent.

THE EXPERIENCE IN THE UNITED STATES WITH REACTOR OPERATION AND REACTOR SAFEGUARDS

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

McCullough, C.R.

1958-10-31

Reactors are operating or planned at locations in the United States in cities, near cities, and at remote locations. There is a general pattern that the higher power reactors are not in, but fairly uear cities, and the testing reactors for more hazardous experiments are at remote locations. A great deal has been done on the theoretical and experimental study of importunt features of reactor design. The metal-water reaction is still a theoretical possibility but tests of fuel element burnout under conditions approaching reactor operation gave no reaction. It appears that nucleate boiling does not necessarily result in steam blanketingmore » and fuel melting. Much attention is being given to the calculation of core kinetics but it is being found that temperature, power, and void coefficients cannot be calculated with accuracy and experiments are required. Some surprises are found giving positive localized void coefficients. Possible oscillatory behavior of reactors is being given careful study. No dangerous oscillations have been found in operating reactors but osciliations hare appeared in experimeats. The design of control and safety systems varies wvith different constructors. The relation of control to the kinetic behavior of the reactor is being studied. The importance of sensing element locations in order to know actual local reactor power level is being recognized. The time constants of instrumentation as related to reactor kinetics are being studied. Pressure vessels for reactors are being designed and manufactured. Many of these are beyond any previous experience. The stress problem is being given careful study. The effect of radiation is being studied experimentally. The stress problems of piping and pressure vessels is a difficult design problem being met successfully in reactor plants. The proper organization and procedure for operation of reactors is being evolved for resourch, testing, and power reactors. The importance of written standards and instructions for both normal and abnormal operating conditions is recogmized. Corfinement of radioactive materials either by tight steel shells, tight buildings, or semi-tight structures vented through filters is considered necessary in the United States. A discussion will be given of specifications, construction, and testing of these structures. The need for emergency plans has been stressed by recent experiences in radioactive releases. The problems of such plans to cover all grades of accidents will be discussed. The theoretical consequences of releases of radioactive materials have been studied and these results will be compared with actual experience. The problem of exposures from normal and abnormal operetion of reactors is a problem of desiga and operation on one hand and the amount of damage to be expected on the other. The safeguard problem is closely related to the acceptable doses of radiouctivity which the ICRP recommend. The future of atomic energy depends upon adequate safeguards and economical design and operation. Accepted criteria are required to guide designers as to the proper balance of caution and boldness. (auth)« less

DESIGN AND HAZARDS SUMMARY REPORT, BOILING REACTOR EXPERIMENT V (BORAX V)

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

None

1961-05-01

Design data for BORAX V are presented along with results of hazards evaluation studies. Considcration of the hazards associated with the operation of BORAX V was based on the following conditions: For normal steady-state power and experimental operation, the reactor and plant are adequately shielded and ventilated to allow personnel to be safely stationed in the turbine building and on the main floor of the reactor building. The control building is located one- half mile distant from the reactor building. For special, hazardous experiments, personnel are withdrawn from the reactor area. (M.C.G.)

Overview of the 2014 Edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook)

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

John D. Bess; J. Blair Briggs; Jim Gulliford

2014-10-01

The International Reactor Physics Experiment Evaluation Project (IRPhEP) is a widely recognized world class program. The work of the IRPhEP is documented in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). Integral data from the IRPhEP Handbook is used by reactor safety and design, nuclear data, criticality safety, and analytical methods development specialists, worldwide, to perform necessary validations of their calculational techniques. The IRPhEP Handbook is among the most frequently quoted reference in the nuclear industry and is expected to be a valuable resource for future decades.

Modelling the activity of 129I in the primary coolant of a CANDU reactor

NASA Astrophysics Data System (ADS)

Lewis, Brent J.; Husain, Aamir

2003-01-01

A mathematical treatment has been developed to describe the activity levels of 129I as a function of time in the primary heat transport system during constant power operation and for a reactor shutdown situation. The model accounts for a release of fission-product iodine from defective fuel rods and tramp uranium contamination on in-core surfaces. The physical transport constants of the model are derived from a coolant activity analysis of the short-lived radioiodine species. An estimate of 3×10 -9 has been determined for the coolant activity ratio of 129I/ 131I in a CANDU Nuclear Generating Station (NGS), which is in reasonable agreement with that observed in the primary coolant and for plant test resin columns from pressurized and boiling water reactor plants. The model has been further applied to a CANDU NGS, by fitting it to the observed short-lived iodine and long-lived cesium data, to yield a coolant activity ratio of ˜2×10 -8 for 129I/ 137Cs. This ratio can be used to estimate the levels of 129I in reactor waste based on a measurement of the activity of 137Cs.

Startup of reactors for anoxic ammonium oxidation: experiences from the first full-scale anammox reactor in Rotterdam.

PubMed

van der Star, Wouter R L; Abma, Wiebe R; Blommers, Dennis; Mulder, Jan-Willem; Tokutomi, Takaaki; Strous, Marc; Picioreanu, Cristian; van Loosdrecht, Mark C M

2007-10-01

The first full-scale anammox reactor in the world was started in Rotterdam (NL). The reactor was scaled-up directly from laboratory-scale to full-scale and treats up to 750 kg-N/d. In the initial phase of the startup, anammox conversions could not be identified by traditional methods, but quantitative PCR proved to be a reliable indicator for growth of the anammox population, indicating an anammox doubling time of 10-12 days. The experience gained during this first startup in combination with the availability of seed sludge from this reactor, will lead to a faster startup of anammox reactors in the future. The anammox reactor type employed in Rotterdam was compared to other reactor types for the anammox process. Reactors with a high specific surface area like the granular sludge reactor employed in Rotterdam provide the highest volumetric loading rates. Mass transfer of nitrite into the biofilm is limiting the conversion of those reactor types that have a lower specific surface area. Now the first full-scale commercial anammox reactor is in operation, a consistent and descriptive nomenclature is suggested for reactors in which the anammox process is employed.

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

Mitchell K Meyer

Blister–threshold testing of fuel plates is a standard method through which the safety margin for operation of plate-type in research and test reactors is assessed. The blister-threshold temperature is indicative of the ability of fuel to operate at high temperatures for short periods of time (transient conditions) without failure. This method of testing was applied to the newly developed U-Mo monolithic fuel system. Blister annealing studies on the U-Mo monolithic fuel plates began in 2007, with the Reduced Enrichment for Research and Test Reactors (RERTR)-6 experiment, and they have continued as the U-Mo fuel system has evolved through the researchmore » and development process. Blister anneal threshold temperatures from early irradiation experiments (RERTR-6 through RERTR-10) ranged from 400 to 500°C. These temperatures were projected to be acceptable for NRC-licensed research reactors and the high-power Advanced Test Reactor (ATR) and the High Flux Isotope Reactor (HFIR) based on current safety-analysis reports (SARs). Initial blister testing results from the RERTR-12 experiment capsules X1 and X2 showed a decrease in the blister-threshold temperatures. Blister threshold temperatures from this experiment ranged from 300 to 400°C. Selected plates from the AFIP-4 experiment, which was fabricated using a process similar to that used to fabricate the RERTR-12 experiment, also underwent blister testing to determine whether results would be similar. The measured blister-threshold temperatures from the AFIP-4 plates fell within the same blister-threshold temperature range measured in the RERTR-12 plates. Investigation of the cause of this decrease in bister threshold temperature is being conducted under the guidance of Idaho National Laboratory PLN-4155, “Analysis of Low Blister Threshold Temperatures in the RERTR-12 and AFIP-4 Experiments,” and is driven by hypotheses. The main focus of the investigation is in the following areas: 1. Fabrication variables 2. Pre-irradiation characterization 3. Irradiation conditions 4. Post-irradiation examination 5. Additional blister testing 6. Mechanical modeling This report documents the preliminary results of this investigation. Several hypotheses can be dismissed as a result of this investigation. Two primary categories of causes remain. The most prominent theory, supported by the data, is that low blister-threshold temperature is the result of mechanical energy imparted on the samples during the fabrication process (hot and cold rolling) without adequate post processing (annealing). The mechanisms are not clearly understood and require further investigation, but can be divided into two categories: • Residual Stress • Undesirable interaction boundary and/or U-Mo microstructure change A secondary theory that cannot be dismissed with the information that is currently available is that a change in the test conditions has resulted in a statistically significant downward shift of measured blister temperature. This report outlines the results of the forensic investigations conducted to date. The data and conclusions presented in this report are preliminary. Definitive cause and effect relationships will be established by future experimental programs.« less

Rapid preparation of high electrochemical performance LiFePO4/C composite cathode material with an ultrasonic-intensified micro-impinging jetting reactor.

PubMed

Dong, Bin; Huang, Xiani; Yang, Xiaogang; Li, Guang; Xia, Lan; Chen, George

2017-11-01

A joint chemical reactor system referred to as an ultrasonic-intensified micro-impinging jetting reactor (UIJR), which possesses the feature of fast micro-mixing, was proposed and has been employed for rapid preparation of FePO 4 particles that are amalgamated by nanoscale primary crystals. As one of the important precursors for the fabrication of lithium iron phosphate cathode, the properties of FePO 4 nano particles significantly affect the performance of the lithium iron phosphate cathode. Thus, the effects of joint use of impinging stream and ultrasonic irradiation on the formation of mesoporous structure of FePO 4 nano precursor particles and the electrochemical properties of amalgamated LiFePO 4 /C have been investigated. Additionally, the effects of the reactant concentration (C=0.5, 1.0 and 1.5molL -1 ), and volumetric flow rate (V=17.15, 51.44, and 85.74mLmin -1 ) on synthesis of FePO 4 ·2H 2 O nucleus have been studied when the impinging jetting reactor (IJR) and UIJR are to operate in nonsubmerged mode. It was affirmed from the experiments that the FePO 4 nano precursor particles prepared using UIJR have well-formed mesoporous structures with the primary crystal size of 44.6nm, an average pore size of 15.2nm, and a specific surface area of 134.54m 2 g -1 when the reactant concentration and volumetric flow rate are 1.0molL -1 and 85.74mLmin -1 respectively. The amalgamated LiFePO 4 /C composites can deliver good electrochemical performance with discharge capacities of 156.7mAhg -1 at 0.1C, and exhibit 138.0mAhg -1 after 100 cycles at 0.5C, which is 95.3% of the initial discharge capacity. Copyright © 2017. Published by Elsevier B.V.

A study to compute integrated dpa for neutron and ion irradiation environments using SRIM-2013

NASA Astrophysics Data System (ADS)

Saha, Uttiyoarnab; Devan, K.; Ganesan, S.

2018-05-01

Displacements per atom (dpa), estimated based on the standard Norgett-Robinson-Torrens (NRT) model, is used for assessing radiation damage effects in fast reactor materials. A computer code CRaD has been indigenously developed towards establishing the infrastructure to perform improved radiation damage studies in Indian fast reactors. We propose a method for computing multigroup neutron NRT dpa cross sections based on SRIM-2013 simulations. In this method, for each neutron group, the recoil or primary knock-on atom (PKA) spectrum and its average energy are first estimated with CRaD code from ENDF/B-VII.1. This average PKA energy forms the input for SRIM simulation, wherein the recoil atom is taken as the incoming ion on the target. The NRT-dpa cross section of iron computed with "Quick" Kinchin-Pease (K-P) option of SRIM-2013 is found to agree within 10% with the standard NRT-dpa values, if damage energy from SRIM simulation is used. SRIM-2013 NRT-dpa cross sections applied to estimate the integrated dpa for Fe, Cr and Ni are in good agreement with established computer codes and data. A similar study carried out for polyatomic material, SiC, shows encouraging results. In this case, it is observed that the NRT approach with average lattice displacement energy of 25 eV coupled with the damage energies from the K-P option of SRIM-2013 gives reliable displacement cross sections and integrated dpa for various reactor spectra. The source term of neutron damage can be equivalently determined in the units of dpa by simulating self-ion bombardment. This shows that the information of primary recoils obtained from CRaD can be reliably applied to estimate the integrated dpa and damage assessment studies in accelerator-based self-ion irradiation experiments of structural materials. This study would help to advance the investigation of possible correlations between the damages induced by ions and reactor neutrons.

Status of advanced fuel candidates for Sodium Fast Reactor within the Generation IV International Forum

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

F. Delage; J. Carmack; C. B. Lee

2013-10-01

The main challenge for fuels for future Sodium Fast Reactor systems is the development and qualification of a nuclear fuel sub-assembly which meets the Generation IV International Forum goals. The Advanced Fuel project investigates high burn-up minor actinide bearing fuels as well as claddings and wrappers to withstand high neutron doses and temperatures. The R&D outcome of national and collaborative programs has been collected and shared between the AF project members in order to review the capability of sub-assembly material and fuel candidates, to identify the issues and select the viable options. Based on historical experience and knowledge, both oxidemore » and metal fuels emerge as primary options to meet the performance and the reliability goals of Generation IV SFR systems. There is a significant positive experience on carbide fuels but major issues remain to be overcome: strong in-pile swelling, atmosphere required for fabrication as well as Pu and Am losses. The irradiation performance database for nitride fuels is limited with longer term R&D activities still required. The promising core material candidates are Ferritic/Martensitic (F/M) and Oxide Dispersed Strengthened (ODS) steels.« less

The PROSPECT physics program

NASA Astrophysics Data System (ADS)

Ashenfelter, J.; Balantekin, A. B.; Band, H. R.; Barclay, G.; Bass, C. D.; Berish, D.; Bignell, L.; Bowden, N. S.; Bowes, A.; Brodsky, J. P.; Bryan, C. D.; Cherwinka, J. J.; Chu, R.; Classen, T.; Commeford, K.; Conant, A. J.; Davee, D.; Dean, D.; Deichert, G.; Diwan, M. V.; Dolinski, M. J.; Dolph, J.; DuVernois, M.; Erikson, A. S.; Febbraro, M. T.; Gaison, J. K.; Galindo-Uribarri, A.; Gilje, K.; Glenn, A.; Goddard, B. W.; Green, M.; Hackett, B. T.; Han, K.; Hans, S.; Heeger, K. M.; Heffron, B.; Insler, J.; Jaffe, D. E.; Jones, D.; Langford, T. J.; Littlejohn, B. R.; Martinez Caicedo, D. A.; Matta, J. T.; McKeown, R. D.; Mendenhall, M. P.; Mueller, P. E.; Mumm, H. P.; Napolitano, J.; Neilson, R.; Nikkel, J. A.; Norcini, D.; Pushin, D.; Qian, X.; Romero, E.; Rosero, R.; Seilhan, B. S.; Sharma, R.; Sheets, S.; Surukuchi, P. T.; Trinh, C.; Varner, R. L.; Viren, B.; Wang, W.; White, B.; White, C.; Wilhelmi, J.; Williams, C.; Wise, T.; Yao, H.; Yeh, M.; Yen, Y.-R.; Zangakis, G. Z.; Zhang, C.; Zhang, X.; PROSPECT Collaboration

2016-11-01

The precision reactor oscillation and spectrum experiment, PROSPECT, is designed to make a precise measurement of the antineutrino spectrum from a highly-enriched uranium reactor and probe eV-scale sterile neutrinos by searching for neutrino oscillations over a distance of several meters. PROSPECT is conceived as a 2-phase experiment utilizing segmented 6Li-doped liquid scintillator detectors for both efficient detection of reactor antineutrinos through the inverse beta decay reaction and excellent background discrimination. PROSPECT Phase I consists of a movable 3 ton antineutrino detector at distances of 7-12 m from the reactor core. It will probe the best-fit point of the {ν }e disappearance experiments at 4σ in 1 year and the favored region of the sterile neutrino parameter space at \\gt 3σ in 3 years. With a second antineutrino detector at 15-19 m from the reactor, Phase II of PROSPECT can probe the entire allowed parameter space below 10 eV2 at 5σ in 3 additional years. The measurement of the reactor antineutrino spectrum and the search for short-baseline oscillations with PROSPECT will test the origin of the spectral deviations observed in recent {θ }13 experiments, search for sterile neutrinos, and conclusively address the hypothesis of sterile neutrinos as an explanation of the reactor anomaly.

A Simple Experiment for Teaching Process Intensification by Static Mixing in Chemical Reaction Engineering

ERIC Educational Resources Information Center

Baz-Rodríguez, Sergio; Herrera-Soberanis, Natali; Rodríguez-Novelo, Miguel; Guillén-Francisc, Juana; Rocha-Uribe, José

2016-01-01

An experiment for teaching mixing intensification in reaction engineering is described. For this, a simple tubular reactor was constructed; helical static mixer elements were fabricated from stainless steel strips and inserted into the reactor. With and without the internals, the equipment operates as a static mixer reactor or a laminar flow…

Phyto-dewatering of sewage sludge using Panicum repens L.

PubMed

El-Gendy, A S; El-Kassas, H I; Razek, T M A; Abdel-Latif, H

2017-04-01

Experiments in the field environment have been conducted to study the growth of Panicum repens L., an aquatic plant, in the sewage sludge matrix. The experiments were also carried out to investigate the ability of this plant to dewater sewage sludge to increase the capacity of conventional drying beds. In addition, the ability of Panicum repens L. to reduce the sludge contents of certain elements (copper (Cu), Iron (Fe), Sodium (Na), lead (Pb), and Zinc (Zn)) was also investigated. All experiments were carried out in batch reactors. Different plant coverage densities were tested (0.00 to 27.3 kg/m 2 ). The liquid sewage sludge was collected from a wastewater treatment plant in Helwan city, Cairo Governorate, Egypt. The collected sludge represents a mixture of the primary sludge and waste activated sludge before discharging into drying beds.

Recent results on implantation and permeation into fusion reactor materials

NASA Astrophysics Data System (ADS)

Anderl, R. A.; Holland, D. F.; Longhurst, G. R.; Struttman, D. A.

This paper reports on implantation-driven permeation experiments that have been made for primary candidate alloy (PCA) and the ferritic steel HT-9 using deuterium ion beams from an accelerator. The results include measurements of the implantation flux and fluence dependence of the deuterium reemission and permeation for specimens heated to approximately 430(0)C. Simultaneous measurements of the ions sputtered from the specimen front surface with a secondary ion mass spectrometer provided some characterization of the surface condition throughout an experiment. For both materials, the permeation rate was lowered by the implantation process. However, the steady state permeation rate for HT-9 was found to be at least a factor of 5 greater than that for PCA.

Spinning fluids reactor

DOEpatents

Miller, Jan D; Hupka, Jan; Aranowski, Robert

2012-11-20

A spinning fluids reactor, includes a reactor body (24) having a circular cross-section and a fluid contactor screen (26) within the reactor body (24). The fluid contactor screen (26) having a plurality of apertures and a circular cross-section concentric with the reactor body (24) for a length thus forming an inner volume (28) bound by the fluid contactor screen (26) and an outer volume (30) bound by the reactor body (24) and the fluid contactor screen (26). A primary inlet (20) can be operatively connected to the reactor body (24) and can be configured to produce flow-through first spinning flow of a first fluid within the inner volume (28). A secondary inlet (22) can similarly be operatively connected to the reactor body (24) and can be configured to produce a second flow of a second fluid within the outer volume (30) which is optionally spinning.

Nitrate treatment effects on bacterial community biofilm formed on carbon steel in produced water stirred tank bioreactor.

PubMed

Marques, Joana Montezano; de Almeida, Fernando Pereira; Lins, Ulysses; Seldin, Lucy; Korenblum, Elisa

2012-06-01

To better understand the impact of nitrate in Brazilian oil reservoirs under souring processes and corrosion, the goal of this study was to analyse the effect of nitrate on bacterial biofilms formed on carbon steel coupons using reactors containing produced water from a Brazilian oil platform. Three independent experiments were carried out (E1, E2 and E3) using the same experimental conditions and different incubation times (5, 45 and 80 days, respectively). In every experiment, two biofilm-reactors were operated: one was treated with continuous nitrate flow (N reactor), and the other was a control reactor without nitrate (C reactor). A Polymerase Chain Reaction-Denaturing Gradient Gel Electrophoresis approach using the 16S rRNA gene was performed to compare the bacterial groups involved in biofilm formation in the N and C reactors. DGGE profiles showed remarkable changes in community structure only in experiments E2 and E3. Five bands extracted from the gel that represented the predominant bacterial groups were identified as Bacillus aquimaris, B. licheniformis, Marinobacter sp., Stenotrophomonas maltophilia and Thioclava sp. A reduction in the sulfate-reducing bacteria (SRB) most probable number counts was observed only during the longer nitrate treatment (E3). Carbon steel coupons used for biofilm formation had a slightly higher weight loss in N reactors in all experiments. When the coupon surfaces were analysed by scanning electron microscopy, an increase in corrosion was observed in the N reactors compared with the C reactors. In conclusion, nitrate reduced the viable SRB counts. Nevertheless, the nitrate dosing increased the pitting of coupons.

Fluidizing a mixture of particulate coal and char

DOEpatents

Green, Norman W.

1979-08-07

Method of mixing particulate materials comprising contacting a primary source and a secondary source thereof whereby resulting mixture ensues; preferably at least one of the two sources has enough motion to insure good mixing and the particulate materials may be heat treated if desired. Apparatus for such mixing comprising an inlet for a primary source, a reactor communicating therewith, a feeding means for supplying a secondary source to the reactor, and an inlet for the secondary source. Feeding means is preferably adapted to supply fluidized materials.

Mixing method and apparatus

DOEpatents

Green, Norman W.

1982-06-15

Method of mixing particulate materials comprising contacting a primary source and a secondary source thereof whereby resulting mixture ensues; preferably at least one of the two sources has enough motion to insure good mixing and the particulate materials may be heat treated if desired. Apparatus for such mixing comprising an inlet for a primary source, a reactor communicating therewith, a feeding means for supplying a secondary source to the reactor, and an inlet for the secondary source. Feeding means is preferably adapted to supply fluidized materials.

Cladding and duct materials for advanced nuclear recycle reactors

NASA Astrophysics Data System (ADS)

Allen, T. R.; Busby, J. T.; Klueh, R. L.; Maloy, S. A.; Toloczko, M. B.

2008-01-01

The expanded use of nuclear energy without risk of nuclear weapons proliferation and with safe nuclear waste disposal is a primary goal of the Global Nuclear Energy Partnership (GNEP). To achieve that goal the GNEP is exploring advanced technologies for recycling spent nuclear fuel that do not separate pure plutonium, and advanced reactors that consume transuranic elements from recycled spent fuel. The GNEP’s objectives will place high demands on reactor clad and structural materials. This article discusses the materials requirements of the GNEP’s advanced nuclear recycle reactors program.

Laboratory Measurements of Biomass Cook-stove Emissions Aged in an Oxidation Flow Reactor: Influence of Combustion and Aging Conditions on Aerosols

NASA Astrophysics Data System (ADS)

Grieshop, A. P.; Reece, S. M.; Sinha, A.; Wathore, R.

2016-12-01

Combustion in rudimentary and improved cook-stoves used by billions in developing countries can be a regionally dominant contributor to black carbon (BC), primary organic aerosols (POA) and precursors for secondary organic aerosol (SOA). Recent studies suggest that SOA formed during photo-oxidation of primary emissions from biomass burning may make important contribution to its atmospheric impacts. However, the extent to which stove type and operating conditions affect the amount, composition and characteristics of SOA formed from the aging of cookstoves emissions is still largely undetermined. Here we present results from experiments with a field portable oxidation flow reactor (F-OFR) designed to assess aging of cook-stove emissions in both laboratory and field settings. Laboratory tests results are used to compare the quantity and properties of fresh and aged emissions from a traditional open fire and twp alternative stove designs operated on the standard and alternate testing protocols. Diluted cookstove emissions were exposed to a range of oxidant concentrations in the F-OFR. Primary emissions were aged both on-line, to study the influence of combustion variability, and sampled from batched emissions in a smog chamber to examine different aging conditions. Data from real-time particle- and gas-phase instruments and integrated filter samples were collected up and down stream of the OFR. The properties of primary emissions vary strongly with stove type and combustion conditions (e.g. smoldering versus flaming). Experiments aging diluted biomass emissions from distinct phases of stove operation (smoldering and flaming) showed peak SOA production for both phases occurred between 3 and 6 equivalent days of aging with slightly greater production observed in flaming phase emissions. Changing combustion conditions had a stronger influence than aging on POA+SOA `emission factors'. Aerosol Chemical Speciation Monitor data show a substantial evolution of aerosol composition with aging. These results highlight the importance of both stoves' operating conditions and aging on composition and characteristics of emissions, which have important implications for regional air quality and climate forcing.

Measurement of neutron spectra in the experimental reactor LR-0

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

Prenosil, Vaclav; Mravec, Filip; Veskrna, Martin

2015-07-01

The measurement of fast neutron fluxes is important in many areas of nuclear technology. It affects the stability of the reactor structural components, performance of fuel, and also the fuel manner. The experiments performed at the LR-0 reactor were in the past focused on the measurement of neutron field far from the core, in reactor pressure vessel simulator or in biological shielding simulator. In the present the measurement in closer regions to core became more important, especially measurements in structural components like reactor baffle. This importance increases with both reactor power increase and also long term operation. Other important taskmore » is an increasing need for the measurement close to the fuel. The spectra near the fuel are aimed due to the planned measurements with the FLIBE salt, in FHR / MSR research, where one of the task is the measurement of the neutron spectra in it. In both types of experiments there is strong demand for high working count rate. The high count rate is caused mainly by high gamma background and by high fluxes. The fluxes in core or in its vicinity are relatively high to ensure safe reactor operation. This request is met in the digital spectroscopic apparatus. All experiments were realized in the LR-0 reactor. It is an extremely flexible light water zero-power research reactor, operated by the Research Center Rez (Czech Republic). (authors)« less

Benchmark Evaluation of HTR-PROTEUS Pebble Bed Experimental Program

DOE PAGES

Bess, John D.; Montierth, Leland; Köberl, Oliver; ...

2014-10-09

Benchmark models were developed to evaluate 11 critical core configurations of the HTR-PROTEUS pebble bed experimental program. Various additional reactor physics measurements were performed as part of this program; currently only a total of 37 absorber rod worth measurements have been evaluated as acceptable benchmark experiments for Cores 4, 9, and 10. Dominant uncertainties in the experimental keff for all core configurations come from uncertainties in the ²³⁵U enrichment of the fuel, impurities in the moderator pebbles, and the density and impurity content of the radial reflector. Calculations of k eff with MCNP5 and ENDF/B-VII.0 neutron nuclear data aremore » greater than the benchmark values but within 1% and also within the 3σ uncertainty, except for Core 4, which is the only randomly packed pebble configuration. Repeated calculations of k eff with MCNP6.1 and ENDF/B-VII.1 are lower than the benchmark values and within 1% (~3σ) except for Cores 5 and 9, which calculate lower than the benchmark eigenvalues within 4σ. The primary difference between the two nuclear data libraries is the adjustment of the absorption cross section of graphite. Simulations of the absorber rod worth measurements are within 3σ of the benchmark experiment values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.« less

A comparison of the technological effectiveness of dairy wastewater treatment in anaerobic UASB reactor and anaerobic reactor with an innovative design.

PubMed

Jedrzejewska-Cicinska, M; Kozak, K; Krzemieniewski, M

2007-10-01

The present research was an investigation of the influence of an innovative design of reactor filled with polyethylene (PE) granulate on model dairy wastewater treatment efficiency under anaerobic conditions compared to that obtained in a typical UASB reactor. The experiment was conducted at laboratory scale. An innovative reactor was designed with the reaction chamber inclined 30 degrees in relation to the ground with upward waste flow and was filled with PE granular material. Raw model dairy wastewater was fed to two anaerobic reactors of different design at the organic loading rate of 4 kg COD m(-3)d(-1). Throughout the experiment, a higher removal efficiency of organic compounds was observed in the reactor with an innovative design and it was higher by 7.1% on average than in the UASB reactor. The total suspended solids was lower in the wastewater treated in the anaerobic reactor with the innovative design. Applying a PE granulated filling in the chamber of the innovative reactor contributed to an even distribution of sludge biomass in the reactor, reducing washout of anaerobic sludge biomass from the reaction chamber and giving a higher organic compounds removal efficiency.

The PROSPECT physics program

DOE PAGES

Ashenfelter, J.; Balantekin, A. B.; Band, H. R.; ...

2016-10-17

The precision reactor oscillation and spectrum experiment, PROSPECT, is designed to make a precise measurement of the antineutrino spectrum from a highly-enriched uranium reactor and probe eV-scale sterile neutrinos by searching for neutrino oscillations over a distance of several meters. The subject of this paper, PROSPECT, is conceived as a 2-phase experiment utilizing segmented 6Li-doped liquid scintillator detectors for both efficient detection of reactor antineutrinos through the inverse beta decay reaction and excellent background discrimination. PROSPECT Phase I consists of a movable 3 ton antineutrino detector at distances of 7–12 m from the reactor core. It will probe the best-fitmore » point of the ν e disappearance experiments at 4σ in 1 year and the favored region of the sterile neutrino parameter space at > 3σ in 3 years. With a second antineutrino detector at 15–19 m from the reactor, Phase II of PROSPECT can probe the entire allowed parameter space below 10 eV 2 at 5σ in 3 additional years. Finally, the measurement of the reactor antineutrino spectrum and the search for short-baseline oscillations with PROSPECT will test the origin of the spectral deviations observed in recent θ 13 experiments, search for sterile neutrinos, and conclusively address the hypothesis of sterile neutrinos as an explanation of the reactor anomaly.« less

Preliminary design and hazards report. Boiling Reactor Experiment V (BORAX V)

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

Rice, R. E.

1960-02-01

The preliminary objectives of the proposed BORAX V program are to test nuclear superheating concepts and to advance the technology of boiling-water-reactor design by performing experiments which will improve the understanding of factors limiting the stability of boiling reactors at high power densities. The reactor vessel is a cylinder with ellipsoidal heads, made of carbon steel clad internally with stainless steel. Each of the three cores is 24 in. high and has an effective diameter of 39 in. This is a preliminary report. (W.D.M.)

Decay heat of sodium fast reactor: Comparison of experimental measurements on the PHENIX reactor with calculations performed with the French DARWIN package

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

Benoit, J. C.; Bourdot, P.; Eschbach, R.

2012-07-01

A Decay Heat (DH) experiment on the whole core of the French Sodium-Cooled Fast Reactor PHENIX has been conducted in May 2008. The measurements began an hour and a half after the shutdown of the reactor and lasted twelve days. It is one of the experiments used for the experimental validation of the depletion code DARWIN thereby confirming the excellent performance of the aforementioned code. Discrepancies between measured and calculated decay heat do not exceed 8%. (authors)

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.

An overview of the Daya Bay reactor neutrino experiment

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

Cao, Jun; Luk, Kam-Biu

2016-04-26

The Daya Bay Reactor Neutrino Experiment discovered an unexpectedly large neutrino oscillation related to the mixing angle θ 13 in 2012. This finding paved the way to the next generation of neutrino oscillation experiments. In this article, we review the history, featured design, and scientific results of Daya Bay. Prospects of the experiment are also described.

Novel micro-reactor flow cell for investigation of model catalysts using in situ grazing-incidence X-ray scattering

PubMed Central

Kehres, Jan; Pedersen, Thomas; Masini, Federico; Andreasen, Jens Wenzel; Nielsen, Martin Meedom; Diaz, Ana; Nielsen, Jane Hvolbæk; Hansen, Ole

2016-01-01

The design, fabrication and performance of a novel and highly sensitive micro-reactor device for performing in situ grazing-incidence X-ray scattering experiments of model catalyst systems is presented. The design of the reaction chamber, etched in silicon on insulator (SIO), permits grazing-incidence small-angle X-ray scattering (GISAXS) in transmission through 10 µm-thick entrance and exit windows by using micro-focused beams. An additional thinning of the Pyrex glass reactor lid allows simultaneous acquisition of the grazing-incidence wide-angle X-ray scattering (GIWAXS). In situ experiments at synchrotron facilities are performed utilizing the micro-reactor and a designed transportable gas feed and analysis system. The feasibility of simultaneous in situ GISAXS/GIWAXS experiments in the novel micro-reactor flow cell was confirmed with CO oxidation over mass-selected Ru nanoparticles. PMID:26917133

Catalyst and process development for synthesis gas conversion to isobutylene. Quarterly report, October 1, 1992--December 31, 1992

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

Anthony, R.G.; Akgerman, A.

1993-02-01

The objectives of this project are to develop a new catalyst, the kinetics for this catalyst, reactor models for trickle bed, slurry and fixed bed reactors, and simulate the performance of fixed bed trickle flow reactors, slurry flow reactors, and fixed bed gas phase reactors for conversion of a hydrogen lean synthesis gas to isobutylene. The goals for the quarter include: (1) Conduct experiments using a trickle bed reactor to determine the effect of reactor type on the product distribution. (2) Use spherical pellets of silica as a support for zirconia for the purpose of increasing surface, area and performancemore » of the catalysts. (3) Conduct exploratory experiments to determine the effect of super critical drying of the catalyst on the catalyst surface area and performance. (4) Prepare a ceria/zirconia catalyst by the precipitation method.« less

Effect of enzymes on anaerobic digestion of primary sludge and septic tank performance.

PubMed

Diak, James; Örmeci, Banu; Kennedy, Kevin J

2012-11-01

Enzyme additives are believed to improve septic tank performance by increasing the hydrolysis and digestion rates and maintaining a healthy microbial population. Previous studies reported mixed results on the effectiveness of enzymes on mesophilic and thermophilic digestion, and it is not clear whether enzymes would be effective under septic tank conditions where there is no heating or mixing, quantities of enzymes added are small, and they can be washed out quickly. In this study, batch reactors and continuous-flow reactors designed and operated as septic tanks were used to evaluate whether enzymatic treatment would increase the hydrolysis and digestion rates in primary sludge. Total solids, volatile solids, total suspended solids, total and soluble chemical oxygen demand, concentrations of protein, carbohydrate, ammonia and volatile acids in sludge and effluent samples were measured to determine the differences in digestion rates in the presence and absence of enzymes. Overall, no significant improvement was observed in enzyme-treated reactors compared with the control reactors.

Resolving Some Paradoxes in the Thermal Decomposition Mechanism of Acetaldehyde

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

Sivaramakrishnan, Raghu; Michael, Joe V.; Harding, Lawrence B.

2015-07-16

The mechanism for the thermal decomposition of acetaldehyde has been revisited with an analysis of literature kinetics experiments using theoretical kinetics. The present modeling study was motivated by recent observations, with very sensitive diagnostics, of some unexpected products in high temperature micro-tubular reactor experiments on the thermal decomposition of CH3CHO and its deuterated analogs, CH3CDO, CD3CHO, and CD3CDO. The observations of these products prompted the authors of these studies to suggest that the enol tautomer, CH2CHOH (vinyl alcohol), is a primary intermediate in the thermal decomposition of acetaldehyde. The present modeling efforts on acetaldehyde decomposition incorporate a master equation re-analysismore » of the CH3CHO potential energy surface (PES). The lowest energy process on this PES is an isomerization of CH3CHO to CH2CHOH. However, the subsequent product channels for CH2CHOH are substantially higher in energy, and the only unimolecular process that can be thermally accessed is a re-isomerization to CH3CHO. The incorporation of these new theoretical kinetics predictions into models for selected literature experiments on CH3CHO thermal decomposition confirms our earlier experiment and theory based conclusions that the dominant decomposition process in CH3CHO at high temperatures is C-C bond fission with a minor contribution (~10-20%) from the roaming mechanism to form CH4 and CO. The present modeling efforts also incorporate a master-equation analysis of the H + CH2CHOH potential energy surface. This bimolecular reaction is the primary mechanism for removal of CH2CHOH, which can accumulate to minor amounts at high temperatures, T > 1000 K, in most lab-scale experiments that use large initial concentrations of CH3CHO. Our modeling efforts indicate that the observation of ketene, water and acetylene in the recent micro-tubular experiments are primarily due to bimolecular reactions of CH3CHO and CH2CHOH with H-atoms, and have no bearing on the unimolecular decomposition mechanism of CH3CHO. The present simulations also indicate that experiments using these micro-tubular reactors when interpreted with the aid of high-level theoretical calculations and kinetics modeling can offer insights into the chemistry of elusive intermediates in high temperature pyrolysis of organic molecules.« less

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

Bess, John D.; Sterbentz, James W.; Snoj, Luka

PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen criticalmore » configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.« less

Tritium Mitigation/Control for Advanced Reactor System

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

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

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

Pre-test analysis of protected loss of primary pump transients in CIRCE-HERO facility

NASA Astrophysics Data System (ADS)

Narcisi, V.; Giannetti, F.; Del Nevo, A.; Tarantino, M.; Caruso, G.

2017-11-01

In the frame of LEADER project (Lead-cooled European Advanced Demonstration Reactor), a new configuration of the steam generator for ALFRED (Advanced Lead Fast Reactor European Demonstrator) was proposed. The new concept is a super-heated steam generator, double wall bayonet tube type with leakage monitoring [1]. In order to support the new steam generator concept, in the framework of Horizon 2020 SESAME project (thermal hydraulics Simulations and Experiments for the Safety Assessment of MEtal cooled reactors), the ENEA CIRCE pool facility will be refurbished to host the HERO (Heavy liquid mEtal pRessurized water cOoled tubes) test section to investigate a bundle of seven full scale bayonet tubes in ALFRED-like thermal hydraulics conditions. The aim of this work is to verify thermo-fluid dynamic performance of HERO during the transition from nominal to natural circulation condition. The simulations have been performed with RELAP5-3D© by using the validated geometrical model of the previous CIRCE-ICE test section [2], in which the preceding heat exchanger has been replaced by the new bayonet bundle model. Several calculations have been carried out to identify thermal hydraulics performance in different steady state conditions. The previous calculations represent the starting points of transient tests aimed at investigating the operation in natural circulation. The transient tests consist of the protected loss of primary pump, obtained by reducing feed-water mass flow to simulate the activation of DHR (Decay Heat Removal) system, and of the loss of DHR function in hot conditions, where feed-water mass flow rate is absent. According to simulations, in nominal conditions, HERO bayonet bundle offers excellent thermal hydraulic behavior and, moreover, it allows the operation in natural circulation.

Thermal-hydraulic interfacing code modules for CANDU reactors

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

Liu, W.S.; Gold, M.; Sills, H.

1997-07-01

The approach for CANDU reactor safety analysis in Ontario Hydro Nuclear (OHN) and Atomic Energy of Canada Limited (AECL) is presented. Reflecting the unique characteristics of CANDU reactors, the procedure of coupling the thermal-hydraulics, reactor physics and fuel channel/element codes in the safety analysis is described. The experience generated in the Canadian nuclear industry may be useful to other types of reactors in the areas of reactor safety analysis.

Reactor shutdown experience

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

Cletcher, J.W.

1995-10-01

This is a regular report of summary statistics relating to recent reactor shutdown experience. The information includes both number of events and rates of occurence. It was compiled from data about operating events that were entered into the SCSS data system by the Nuclear Operations Analysis Center at the Oak ridge National Laboratory and covers the six mont period of July 1 to December 31, 1994. Cumulative information, starting from May 1, 1994, is also reported. Updates on shutdown events included in earlier reports is excluded. Information on shutdowns as a function of reactor power at the time of themore » shutdown for both BWR and PWR reactors is given. Data is also discerned by shutdown type and reactor age.« less

Anaerobic co-digestion of fruit and vegetable wastes and primary sewage sludge.

PubMed

Velmurugan, B; Arathy, E C; Hemalatha, R; Philip, Jerry Elsa; Alwar Ramanujam, R

2010-01-01

Anaerobic co-digestion of fruit and vegetable wastes (FVW) and primary sewage sludge was carried out in a fed-batch reactor having a volume of 21 under ambient temperature conditions. Three different proportions (25:75, 50:50 and 75:25 in terms ofVS) of fruit and vegetable wastes and primary sewage sludge were studied for an organic loading rate (OLR) of 1.0 g VS/ l.d and with a hydraulic retention time (HRT) of 25 days. The reactor with 75% FVW and 25% sewage sludge (in terms of VS) showed better performance in terms of VS reduction and biogas yield when compared to other two proportions.

Method for removing cesium from a nuclear reactor coolant

DOEpatents

Colburn, R.P.

1983-08-10

A method of and system for removing cesium from a liquid metal reactor coolant including a carbon packing trap in the primary coolant system for absorbing a major portion of the radioactive cesium from the coolant flowing therethrough at a reduced temperature. A regeneration subloop system having a secondary carbon packing trap is selectively connected to the primary system for isolating the main trap therefrom and connecting it to the regeneration system. Increasing the temperature of the sodium flowing through the primary trap diffuses a portion of the cesium inventory thereof further into the carbon matrix while simultaneously redispersing a portion into the regeneration system for absorption at a reduced temperature by the secondary trap.

Design and Status of RERTR Irradiation Tests in the Advanced Test Reactor

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

Daniel M. Wachs; Richard G. Ambrosek; Gray Chang

2006-10-01

Irradiation testing of U-Mo based fuels is the central component of the Reduced Enrichment for Research and Test Reactors (RERTR) program fuel qualification plan. Several RERTR tests have recently been completed or are planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory in Idaho Falls, ID. Four mini-plate experiments in various stages of completion are described in detail, including the irradiation test design, objectives, and irradiation conditions. Observations made during and after the in-reactor RERTR-7A experiment breach are summarized. The irradiation experiment design and planned irradiation conditions for full-size plate test are described. Progressmore » toward element testing will be reviewed.« less

Current Results of NEUTRINO-4 Experiment

NASA Astrophysics Data System (ADS)

Serebrov, A.; Ivochkin, V.; Samoilov, R.; Fomin, A.; Polyushkin, A.; Zinoviev, V.; Neustroev, P.; Golovtsov, V.; Chernyj, A.; Zherebtsov, O.; Martemyanov, V.; Tarasenkov, V.; Aleshin, V.; Petelin, A.; Izhutov, A.; Tuzov, A.; Sazontov, S.; Ryazanov, D.; Gromov, M.; Afanasiev, V.; Zaytsev, M.; Chaikovskii, M.

2017-12-01

The main goal of experiment “Neutrino-4” is to search for the oscillation of reactor antineutrino to a sterile state. Experiment is conducted on SM-3 research reactor (Dimitrovgrad, Russia). Data collection with full-scale detector with liquid scintillator volume of 3m3 was started in June 2016. We present the results of measurements of reactor antineutrino flux dependence on the distance in range 6- 12 meters from the center of the reactor. At that distance range, the fit of experimental dependence has good agreement with the law 1/L2. Which means, at achieved during the data collecting accuracy level oscillations to sterile state are not observed. In addition, the spectrum of prompt signals of neutrino-like events at different distances have been presented.

Comprehensive Experiments on Subcritical Assemblies of Cascade Reactor Systems

NASA Astrophysics Data System (ADS)

Zavyalov, N. V.; Il'kaev, R. I.; Kolesov, V. F.; Ivanin, I. A.; Zhitnik, A. K.; Kuvshinov, M. I.; Nefedov, Yu. Ya.; Punin, V. T.; Tel'nov, A. V.; Khoruzhi, V. Kh.

2017-12-01

Cascade reactors attract particular attention because of their capability of improving the parameters of pulsed reactors and achieving the feasibility of electronuclear facilities. The paper presents the results of three series of experiments on uranium-neptunium cascade assemblies at the Institute of Nuclear and Radiation Physics of the All-Russian Research Institute of Experimental Physics conducted in 2003-2004. The experiments confirmed theoretical conclusions on positive properties of cascade blankets and effectiveness of using neptunium-237 as a means of creating a one-sided connection between the sections.

Cosmic muon background and reactor neutrino detectors: the Angra experiment

NASA Astrophysics Data System (ADS)

Casimiro, E.; Anjos, J. C.

2008-06-01

We discuss on the importance of appropriately taking into account the cosmic background in the design of reactor neutrino detectors. In particular, as a practical study case, we describe the Angra Project, a new reactor neutrino oscillation experiment proposed to be built in the coming years at the Brazilian nuclear power complex, located near the Angra dos Reis city. The main goal of the experiment is to measure with high precision θ13, the last unknown of the three neutrino mixing angles. The experiment will in addition explore the possibility of using neutrino detectors for purposes of safeguards and non-proliferation of nuclear weapons.

Technical Information on the Carbonation of the EBR-II Reactor, Summary Report Part 1: Laboratory Experiments and Application to EBR-II Secondary Sodium System

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

Steven R. Sherman

Residual sodium is defined as sodium metal that remains behind in pipes, vessels, and tanks after the bulk sodium metal has been melted and drained from such components. The residual sodium has the same chemical properties as bulk sodium, and differs from bulk sodium only in the thickness of the sodium deposit. Typically, sodium is considered residual when the thickness of the deposit is less than 5-6 cm. This residual sodium must be removed or deactivated when a pipe, vessel, system, or entire reactor is permanently taken out of service, in order to make the component or system safer and/ormore » to comply with decommissioning regulations. As an alternative to the established residual sodium deactivation techniques (steam-and-nitrogen, wet vapor nitrogen, etc.), a technique involving the use of moisture and carbon dioxide has been developed. With this technique, sodium metal is converted into sodium bicarbonate by reacting it with humid carbon dioxide. Hydrogen is emitted as a by-product. This technique was first developed in the laboratory by exposing sodium samples to humidified carbon dioxide under controlled conditions, and then demonstrated on a larger scale by treating residual sodium within the Experimental Breeder Reactor II (EBR-II) secondary cooling system, followed by the primary cooling system, respectively. The EBR-II facility is located at the Idaho National Laboratory (INL) in southeastern Idaho, U.S.A. This report is Part 1 of a two-part report. It is divided into three sections. The first section describes the chemistry of carbon dioxide-water-sodium reactions. The second section covers the laboratory experiments that were conducted in order to develop the residual sodium deactivation process. The third section discusses the application of the deactivation process to the treatment of residual sodium within the EBR-II secondary sodium cooling system. Part 2 of the report, under separate cover, describes the application of the technique to residual sodium treatment within the EBR-II primary sodium cooling system and related systems.« less

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.

Reforming and decomposition of glucose in an aqueous phase

NASA Technical Reports Server (NTRS)

Amin, S.; Reid, R. C.; Modell, M.

1975-01-01

Exploratory experiments have been carried out to study the decomposition of glucose, a typical carbohydrate, in a high temperature-high pressure water reactor. The objective of the study was to examine the feasibility of such a process to decompose cellulosic waste materials in long-term space missions. At temperatures below the critical point of water, glucose decomposed to form liquid products and char. Little gas was noted with or without reforming catalysts present. The rate of the primary glucose reaction increased significantly with temperature. Partial identification of the liquid phase was made and the C:H:O ratios determined for both the liquid and solid products. One of the more interesting results from this study was the finding that when glucose was injected into a reactor held at the critical temperature (and pressure) of water, no solid products formed. Gas production increased, but the majority of the carbon was found in soluble furans (and furan derivatives). This significant result is now being investigated further.

Toward a Mechanistic Source Term in Advanced Reactors: A Review of Past U.S. SFR Incidents, Experiments, and Analyses

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

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

In 2015, as part of a Regulatory Technology Development Plan (RTDP) effort for sodium-cooled fast reactors (SFRs), Argonne National Laboratory investigated the current state of knowledge of source term development for a metal-fueled, pool-type SFR. This paper provides a summary of past domestic metal-fueled SFR incidents and experiments and highlights information relevant to source term estimations that were gathered as part of the RTDP effort. The incidents described in this paper include fuel pin failures at the Sodium Reactor Experiment (SRE) facility in July of 1959, the Fermi I meltdown that occurred in October of 1966, and the repeated meltingmore » of a fuel element within an experimental capsule at the Experimental Breeder Reactor II (EBR-II) from November 1967 to May 1968. The experiments described in this paper include the Run-Beyond-Cladding-Breach tests that were performed at EBR-II in 1985 and a series of severe transient overpower tests conducted at the Transient Reactor Test Facility (TREAT) in the mid-1980s.« less

Biparticle fluidized bed reactor

DOEpatents

Scott, Charles D.; Marasco, Joseph A.

1995-01-01

A fluidized bed reactor system utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary particulate phase, continuously introduced and removed in either cocurrent or countercurrent mode, acts in a secondary role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Introduction and removal of the sorbent phase is accomplished through the use of feed screw mechanisms and multivane slurry valves.

Biparticle fluidized bed reactor

DOEpatents

Scott, Charles D.; Marasco, Joseph A.

1996-01-01

A fluidized bed reactor system which utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary and tertiary particulate phases, continuously introduced and removed simultaneously in the cocurrent and countercurrent mode, act in a role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Means for introducing and removing the sorbent phases include feed screw mechanisms and multivane slurry valves.

Biparticle fluidized bed reactor

DOEpatents

Scott, C.D.; Marasco, J.A.

1995-04-25

A fluidized bed reactor system utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary particulate phase, continuously introduced and removed in either cocurrent or countercurrent mode, acts in a secondary role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Introduction and removal of the sorbent phase is accomplished through the use of feed screw mechanisms and multivane slurry valves. 3 figs.

Biparticle fluidized bed reactor

DOEpatents

Scott, C.D.

1993-12-14

A fluidized bed reactor system which utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase is described. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary particulate phase, continuously introduced and removed in either cocurrent or countercurrent mode, acts in a secondary role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Introduction and removal of the sorbent phase is accomplished through the use of feed screw mechanisms and multivane slurry valves. 3 figures.

Biparticle fluidized bed reactor

DOEpatents

Scott, C.D.; Marasco, J.A.

1996-02-27

A fluidized bed reactor system is described which utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary and tertiary particulate phases, continuously introduced and removed simultaneously in the cocurrent and countercurrent mode, act in a role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Means for introducing and removing the sorbent phases include feed screw mechanisms and multivane slurry valves. 3 figs.

Biparticle fluidized bed reactor

DOEpatents

Scott, Charles D.

1993-01-01

A fluidized bed reactor system which utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary particulate phase, continuously introduced and removed in either cocurrent or countercurrent mode, acts in a secondary role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Introduction and removal of the sorbent phase is accomplished through the use of feed screw mechanisms and multivane slurry valves.

Latest results from Daya Bay

NASA Astrophysics Data System (ADS)

Vorobel, Vit; Daya Bay Collaboration

2017-07-01

The Daya Bay Reactor Neutrino Experiment was designed to measure θ 13, the smallest mixing angle in the three-neutrino mixing framework, with unprecedented precision. The experiment consists of eight functionally identical detectors placed underground at different baselines from three pairs of nuclear reactors in South China. Since Dec. 2011, the experiment has been running stably for more than 4 years, and has collected the largest reactor anti-neutrino sample to date. Daya Bay is able to greatly improve the precision on θ 13 and to make an independent measurement of the effective mass splitting in the electron antineutrino disappearance channel. Daya Bay can also perform a number of other precise measurements, such as a high-statistics determination of the absolute reactor antineutrino flux and spectrum, as well as a search for sterile neutrino mixing, among others. The most recent results from Daya Bay are discussed in this paper, as well as the current status and future prospects of the experiment.

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.

Biodegradation of pharmaceuticals in hospital wastewater by staged Moving Bed Biofilm Reactors (MBBR).

PubMed

Casas, Mònica Escolà; Chhetri, Ravi Kumar; Ooi, Gordon; Hansen, Kamilla M S; Litty, Klaus; Christensson, Magnus; Kragelund, Caroline; Andersen, Henrik R; Bester, Kai

2015-10-15

Hospital wastewater represents a significant input of pharmaceuticals into municipal wastewater. As Moving Bed Biofilm Reactors (MBBRs) appear to remove organic micro-pollutants, hospital wastewater was treated with a pilot plant consisting of three MBBRs in series. The removal of pharmaceuticals was studied in two experiments: 1) A batch experiment where pharmaceuticals were spiked to each reactor and 2) a continuous flow experiment at native concentrations. DOC removal, nitrification as well as removal of pharmaceuticals (including X-ray contrast media, β-blockers, analgesics and antibiotics) occurred mainly in the first reactor. In the batch experiment most of the compounds followed a single first-order kinetics degradation function, giving degradation rate constants ranged from 5.77 × 10(-3) to 4.07 h(-1), from -5.53 × 10(-3) to 9.24 × 10(-1) h(-1) and from 1.83 × 10(-3) to 2.42 × 10(-1) h(-1) for first, second and third reactor respectively. Generally, the highest removal rate constants were found in the first reactor while the lowest were found in the third one. This order was inverted for most compounds, when the removal rate constants were normalized to biomass, indicating that the last tank had the most effective biofilms. In the batch experiment, 21 out of 26 compounds were assessed to be degraded with more than 20% within the MBBR train. In the continuous flow experiment the measured removal rates were lower than those estimated from the batch experiments. Copyright © 2015 Elsevier Ltd. All rights reserved.

Heat-transfer analysis of double-pipe heat exchangers for indirect-cycle SCW NPP

NASA Astrophysics Data System (ADS)

Thind, Harwinder

SuperCritical-Water-cooled Reactors (SCWRs) are being developed as one of the Generation-IV nuclear-reactor concepts. SuperCritical Water (SCW) Nuclear Power Plants (NPPs) are expected to have much higher operating parameters compared to current NPPs, i.e., pressure of about 25 MPa and outlet temperature up to 625 °C. This study presents the heat transfer analysis of an intermediate Heat exchanger (HX) design for indirect-cycle concepts of Pressure-Tube (PT) and Pressure-Vessel (PV) SCWRs. Thermodynamic configurations with an intermediate HX gives a possibility to have a single-reheat option for PT and PV SCWRs without introducing steam-reheat channels into a reactor. Similar to the current CANDU and Pressurized Water Reactor (PWR) NPPs, steam generators separate the primary loop from the secondary loop. In this way, the primary loop can be completely enclosed in a reactor containment building. This study analyzes the heat transfer from a SCW primary (reactor) loop to a SCW and Super-Heated Steam (SHS) secondary (turbine) loop using a double-pipe intermediate HX. The numerical model is developed with MATLAB and NIST REFPROP software. Water from the primary loop flows through the inner pipe, and water from the secondary loop flows through the annulus in the counter direction of the double-pipe HX. The analysis on the double-pipe HX shows temperature and profiles of thermophysical properties along the heated length of the HX. It was found that the pseudocritical region has a significant effect on the temperature profiles and heat-transfer area of the HX. An analysis shows the effect of variation in pressure, temperature, mass flow rate, and pipe size on the pseudocritical region and the heat-transfer area of the HX. The results from the numerical model can be used to optimize the heat-transfer area of the HX. The higher pressure difference on the hot side and higher temperature difference between the hot and cold sides reduces the pseudocritical-region length, thus decreases the heat-transfer surface area of the HX.

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

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.

ENGINEERING APPLICATIONS OF ANALOG COMPUTERS

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

Bryant, L.T.; Janicke, M.J.; Just, L.C.

1963-10-31

Six experiments from the fields of reactor engineering, heat transfer, and dynamics are presented to illustrate the engineering applications of analog computers. The steps required for producing the analog solution are shown, as well as complete information for duplicating the solution. Graphical results are provided. The experiments include: deceleration of a reactor control rod, pressure variations through a packed bed, reactor kinetics over many decades with thermal feedback, a vibrating system with two degrees of freedom, temperature distribution in a radiating fin, temperature distribution in an infinite slab considering variable thermal properties, and iodine -xenon buildup in a reactor. (M.C.G.)

In situ monitored in-pile creep testing of zirconium alloys

NASA Astrophysics Data System (ADS)

Kozar, R. W.; Jaworski, A. W.; Webb, T. W.; Smith, R. W.

2014-01-01

The experiments described herein were designed to investigate the detailed irradiation creep behavior of zirconium based alloys in the HALDEN Reactor spectrum. The HALDEN Test Reactor has the unique capability to control both applied stress and temperature independently and externally for each specimen while the specimen is in-reactor and under fast neutron flux. The ability to monitor in situ the creep rates following a stress and temperature change made possible the characterization of creep behavior over a wide stress-strain-rate-temperature design space for two model experimental heats, Zircaloy-2 and Zircaloy-2 + 1 wt%Nb, with only 12 test specimens in a 100-day in-pile creep test program. Zircaloy-2 specimens with and without 1 wt% Nb additions were tested at irradiation temperatures of 561 K and 616 K and stresses ranging from 69 MPa to 455 MPa. Various steady state creep models were evaluated against the experimental results. The irradiation creep model proposed by Nichols that separates creep behavior into low, intermediate, and high stress regimes was the best model for predicting steady-state creep rates. Dislocation-based primary creep, rather than diffusion-based transient irradiation creep, was identified as the mechanism controlling deformation during the transitional period of evolving creep rate following a step change to different test conditions.

Posttest analysis of a 1:6-scale reinforced concrete reactor containment building

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

Weatherby, J.R.

In an experiment conducted at Sandia National Laboratories, 1:6-scale model of a reinforced concrete light water reactor containment building was pressurized with nitrogen gas to more than three times its design pressure. The pressurization produced one large tear and several smaller tears in the steel liner plate that functioned as the primary pneumatic seal for the structure. The data collected from the overpressurization test have been used to evaluate and further refine methods of structural analysis that can be used to predict the performance of containment buildings under conditions produced by a severe accident. This report describes posttest finite elementmore » analyses of the 1:6-scale model tests and compares pretest predictions of the structural response to the experimental results. Strain and displacements calculated in axisymmetric finite element analyses of the 1:6-scale model are compared to strains and displacement measured in the experiment. Detailed analyses of the liner plate are also described in the report. The region of the liner surrounding the large tear was analyzed using two different two-dimensional finite elements model. The results from these analyzed indicate that the primary mechanisms that initiated the tear can be captured in a two- dimensional finite element model. Furthermore, the analyses show that studs used to anchor the liner to the concrete wall, played an important role in initiating the liner tear. Three-dimensional finite element analyses of liner plates loaded by studs are also presented. Results from the three-dimensional analyses are compared to results from two-dimensional analyses of the same problems. 12 refs., 56 figs., 1 tab.« less

Development and Characterization of 6Li-doped Liquid Scintillator Detectors for PROSPECT

NASA Astrophysics Data System (ADS)

Gaison, Jeremy; Prospect Collaboration

2016-09-01

PROSPECT, the Precision Reactor Oscillation and Spectrum experiment, is a phased reactor antineutrino experiment designed to search for eV-scale sterile neutrinos via short-baseline neutrino oscillations and to make a precision measurement of the 235U reactor antineutrino spectrum. A multi-ton, optically segmented detector will be deployed at Oak Ridge National Laboratory's (ORNL) High Flux Isotope Reactor (HFIR) to measure the reactor spectrum at baselines ranging from 7-12m. A two-segment detector prototype with 50 liters of active liquid scintillator target has been built to verify the detector design and to benchmark its performance. In this presentation, we will summarize the performance of this detector prototype and describe the optical and energy calibration of the segmented PROSPECT detectors.

HTR-PROTEUS pebble bed experimental program cores 9 & 10: columnar hexagonal point-on-point packing with a 1:1 moderator-to-fuel pebble ratio

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

Bess, John D.

2014-03-01

PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen criticalmore » configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.« less

HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORES 5, 6, 7, & 8: COLUMNAR HEXAGONAL POINT-ON-POINT PACKING WITH A 1:2 MODERATOR-TO-FUEL PEBBLE RATIO

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

John D. Bess

2013-03-01

PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen criticalmore » configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.« less

HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORES 9 & 10: COLUMNAR HEXAGONAL POINT-ON-POINT PACKING WITH A 1:1 MODERATOR-TO-FUEL PEBBLE RATIO

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

John D. Bess

2013-03-01

PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen criticalmore » configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.« less

Experimental investigation of a new method for advanced fast reactor shutdown cooling

NASA Astrophysics Data System (ADS)

Pakholkov, V. V.; Kandaurov, A. A.; Potseluev, A. I.; Rogozhkin, S. A.; Sergeev, D. A.; Troitskaya, Yu. I.; Shepelev, S. F.

2017-07-01

We consider a new method for fast reactor shutdown cooling using a decay heat removal system (DHRS) with a check valve. In this method, a coolant from the decay heat exchanger (DHX) immersed into the reactor upper plenum is supplied to the high-pressure plenum and, then, inside the fuel subassemblies (SAs). A check valve installed at the DHX outlet opens by the force of gravity after primary pumps (PP-1) are shut down. Experimental studies of the new and alternative methods of shutdown cooling were performed at the TISEY test facility at OKBM. The velocity fields in the upper plenum of the reactor model were obtained using the optical particle image velocimetry developed at the Institute of Applied Physics (Russian Academy of Sciences). The study considers the process of development of natural circulation in the reactor and the DHRS models and the corresponding evolution of the temperature and velocity fields. A considerable influence of the valve position in the displacer of the primary pump on the natural circulation of water in the reactor through the DHX was discovered (in some modes, circulation reversal through the DHX was obtained). Alternative DHRS designs without a shell at the DHX outlet with open and closed check valve are also studied. For an open check valve, in spite of the absence of a shell, part of the flow is supplied through the DHX pipeline and then inside the SA simulators. When simulating power modes of the reactor operation, temperature stratification of the liquid was observed, which increased in the cooling mode via the DHRS. These data qualitatively agree with the results of tests at BN-600 and BN-800 reactors.

SoLid: Search for Oscillations with Lithium-6 Detector at the SCK-CEN BR2 reactor

NASA Astrophysics Data System (ADS)

Ban, G.; Beaumont, W.; Buhour, J. M.; Coupé, B.; Cucoanes, A. S.; D'Hondt, J.; Durand, D.; Fallot, M.; Fresneau, S.; Giot, L.; Guillon, B.; Guilloux, G.; Janssen, X.; Kalcheva, S.; Koonen, E.; Labare, M.; Moortgat, C.; Pronost, G.; Raes, L.; Ryckbosch, D.; Ryder, N.; Shitov, Y.; Vacheret, A.; Van Mulders, P.; Van Remortel, N.; Weber, A.; Yermia, F.

2016-04-01

Sterile neutrinos have been considered as a possible explanation for the recent reactor and Gallium anomalies arising from reanalysis of reactor flux and calibration data of previous neutrino experiments. A way to test this hypothesis is to look for distortions of the anti-neutrino energy caused by oscillation from active to sterile neutrino at close stand-off (˜ 6- 8m) of a compact reactor core. Due to the low rate of anti-neutrino interactions the main challenge in such measurement is to control the high level of gamma rays and neutron background. The SoLid experiment is a proposal to search for active-to-sterile anti-neutrino oscillation at very short baseline of the SCK•CEN BR2 research reactor. This experiment uses a novel approach to detect anti-neutrino with a highly segmented detector based on Lithium-6. With the combination of high granularity, high neutron-gamma discrimination using 6LiF:ZnS(Ag) and precise localization of the Inverse Beta Decay products, a better experimental sensitivity can be achieved compared to other state-of-the-art technology. This compact system requires minimum passive shielding allowing for very close stand off to the reactor. The experimental set up of the SoLid experiment and the BR2 reactor will be presented. The new principle of neutrino detection and the detector design with expected performance will be described. The expected sensitivity to new oscillations of the SoLid detector as well as the first measurements made with the 8 kg prototype detector deployed at the BR2 reactor in 2013-2014 will be reported.

Catalog of experimental projects for a fissioning plasma reactor

NASA Technical Reports Server (NTRS)

Lanzo, C. D.

1973-01-01

Experimental and theoretical investigations were carried out to determine the feasibility of using a small scale fissioning uranium plasma as the power source in a driver reactor. The driver system is a light water cooled and moderated reactor of the MTR type. The eight experiments and proposed configurations for the reactor are outlined.

MTR, SOUTH FACE OF REACTOR. SPECIAL SUPPLEMENTAL SHIELDING WAS REQUIRED ...

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

MTR, SOUTH FACE OF REACTOR. SPECIAL SUPPLEMENTAL SHIELDING WAS REQUIRED OUTSIDE OF MTR FOR EXPERIMENTS. THE AIRCRAFT NUCLEAR PROPULSION PROJECT DOMINATED THE USE OF THIS PART OF THE MTR. INL NEGATIVE NO. 7225. Unknown Photographer, 11/28/1952 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

SoLid Detector Technology

NASA Astrophysics Data System (ADS)

Labare, Mathieu

2017-09-01

SoLid is a reactor anti-neutrino experiment where a novel detector is deployed at a minimum distance of 5.5 m from a nuclear reactor core. The purpose of the experiment is three-fold: to search for neutrino oscillations at a very short baseline; to measure the pure 235U neutrino energy spectrum; and to demonstrate the feasibility of neutrino detectors for reactor monitoring. This report presents the unique features of the SoLid detector technology. The technology has been optimised for a high background environment resulting from low overburden and the vicinity of a nuclear reactor. The versatility of the detector technology is demonstrated with a 288 kg detector prototype which was deployed at the BR2 nuclear reactor in 2015. The data presented includes both reactor on, reactor off and calibration measurements. The measurement results are compared with Monte Carlo simulations. The 1.6t SoLid detector is currently under construction, with an optimised design and upgraded material technology to enhance the detector capabilities. Its deployement on site is planned for the begin of 2017 and offers the prospect to resolve the reactor anomaly within about two years.

Issues relating to spent nuclear fuel storage on the Oak Ridge Reservation

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

Klein, J.A.; Turner, D.W.

1994-12-31

Currently, about 2,800 metric tons of spent nuclear fuel (SNF) is stored in the US, 1,000 kg of SNF (or about 0.03% of the nation`s total) are stored at the US Department of Energy (DOE) complex in Oak Ridge, Tennessee. However small the total quantity of material stored at Oak Ridge, some of the material is quite singular in character and, thus, poses unique management concerns. The various types of SNF stored at Oak Ridge will be discussed including: (1) High-Flux Isotope Reactor (HFIR) and future Advanced Neutron Source (ANS) fuels; (2) Material Testing Reactor (MTR) fuels, including Bulk Shieldingmore » Reactor (BSR) and Oak Ridge Research Reactor (ORR) fuels; (3) Molten Salt Reactor Experiment (MSRE) fuel; (4) Homogeneous Reactor Experiment (HRE) fuel; (5) Miscellaneous SNF stored in Oak Ridge National Laboratory`s (ORNL`s) Solid Waste Storage Areas (SWSAs); (6) SNF stored in the Y-12 Plant 9720-5 Warehouse including Health. Physics Reactor (HPRR), Space Nuclear Auxiliary Power (SNAP-) 10A, and DOE Demonstration Reactor fuels.« less

RACEWAY REACTOR FOR MICROALGAL BIODIESEL PRODUCTION

EPA Science Inventory

The proposed mathematical model incorporating mass transfer, hydraulics, carbonate/aquatic chemistry, biokinetics, biology and reactor design will be calibrated and validated using the data to be generated from the experiments. The practical feasibility of the proposed reactor...

77 FR 63897 - Notice of License Terminations for National Aeronautics and Space Administration; Plum Brook...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-17

... test reactor, constructed to perform irradiation testing of fueled and unfueled experiments for space... constructed to test ``mock-up'' irradiation components for the Plum Brook Reactor. The reactors operated from...

Spatial variation of a short-lived intermediate chemical species in a Couette reactor

NASA Astrophysics Data System (ADS)

Vigil, R. Dennis; Ouyang, Q.; Swinney, Harry L.

1992-04-01

We have conducted experiments and simulations of the spatial variation of a short-lived intermediate species (triiodide) in the autocatalytic oxidation of arsenite by iodate in a reactor that is essentially one dimensional—the Couette reactor. (This reactor consists of two concentric cylinders with the inner one rotating and the outer one at rest; reagents are continuously fed and removed at each end in such a way that there is no net axial flux and there are opposing arsenite and iodate gradients.) The predictions of a one-dimensional reaction-diffusion model, which has no adjustable parameters, are in good qualitative (and, in some cases, quantitative) agreement with experiments. Thus, the Couette reactor, which is used to deliberately create spatial inhomogeneities, can be exploited to enhance the recovery of short-lived intermediate species relative to that which can be obtained with either a batch or continuous-flow stirred-tank reactor.

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.

Catalytic wet oxidation: mathematical modeling of multicompound destruction.

PubMed

Yang, J; Hand, D W; Hokanson, D R; Crittenden, J C; Oman, E J

2003-01-01

A mathematical model of a three-phase catalytic reactor, CatReac, was developed for analysis and optimization of a catalytic oxidation reactor that is used in the International Space Station potable water processor. The packed-bed catalytic reactor, known as the volatile reactor assembly (VRA), is operated as a three-phase reactor and contains a proprietary catalyst, a pure-oxygen gas phase, and the contaminated water. The contaminated water being fed to the VRA primarily consists of acetic acid, acetone, ethanol, 1-propanol, 2-propanol, and propionic acid ranging in concentration from 1 to 10 mg/L. The Langmuir-Hinshelwood Hougen-Watson (L-H) (Hougen, 1943) expression was used to describe the surface reaction rate for these compounds. Single and multicompound short-column experiments were used to determine the L-H rate parameters and calibrate the model. The model was able to predict steady-state multicomponent effluent profiles for short and full-scale reactor experiments.

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.

Proceedings of the 1992 topical meeting on advances in reactor physics. Volume 2

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

Not Available

1992-04-01

This document, Volume 2, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Transport Theory; Fast Reactors; Plant Analyzers; Integral Experiments/Measurements & Analysis; Core Computational Systems; Reactor Physics; Monte Carlo; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual reports have been cataloged separately. (FI)

Summary of ORSphere critical and reactor physics measurements

NASA Astrophysics Data System (ADS)

Marshall, Margaret A.; Bess, John D.

2017-09-01

In the early 1970s Dr. John T. Mihalczo (team leader), J.J. Lynn, and J.R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s. The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with the GODIVA I experiments. This critical configuration has been evaluated. Preliminary results were presented at ND2013. Since then, the evaluation was finalized and judged to be an acceptable benchmark experiment for the International Criticality Safety Benchmark Experiment Project (ICSBEP). Additionally, reactor physics measurements were performed to determine surface button worths, central void worth, delayed neutron fraction, prompt neutron decay constant, fission density and neutron importance. These measurements have been evaluated and found to be acceptable experiments and are discussed in full detail in the International Handbook of Evaluated Reactor Physics Benchmark Experiments. The purpose of this paper is to summarize all the evaluated critical and reactor physics measurements evaluations.

Measurement of neutrino mixing angle θ13 and mass difference Δ mee2 from reactor antineutrino disappearance in the RENO experiment

NASA Astrophysics Data System (ADS)

Kim, Soo-Bong

2016-07-01

RENO (Reactor Experiment for Neutrino Oscillation) made a definitive measurement of the smallest neutrino mixing angle θ13 in 2012, based on the disappearance of reactor electron antineutrinos. The experiment has obtained a more precise value of the mixing angle and the first result on neutrino mass difference Δ mee2 from an energy and baseline dependent reactor neutrino disappearance using ∼500 days of data. Based on the ratio of inverse-beta-decay (IBD) prompt spectra measured in two identical far and near detectors, we obtain sin2 ⁡ (2θ13) = 0.082 ± 0.009 (stat .) ± 0.006 (syst .) and | Δ mee2 | = [2.62-0.23+0.21 (stat.)-0.13+0.12 (syst .) ] ×10-3 eV2. An excess of reactor antineutrinos near 5 MeV is observed in the measured prompt spectrum with respect to the most commonly used models. The excess is found to be consistent with coming from reactors. A successful measurement of θ13 is also made in an IBD event sample with a delayed signal of neutron capture on hydrogen. A precise value of θ13 would provide important information on determination of the leptonic CP phase if combined with a result of an accelerator neutrino beam experiment.

77 FR 7610 - Notice of Availability of Environmental Assessment and Finding of No Significant Impact for...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-02-13

... and reinforced concrete floors acting as diaphragms in distributing loads to vertically resisting... reinforced concrete foundation. The reactor is fueled with standard low-enriched TRIGA (Training, Research... cooled by a light water primary system consisting of the reactor pool and a heat removal system to remove...

Detecting Dark Photons with Reactor Neutrino Experiments

NASA Astrophysics Data System (ADS)

Park, H. K.

2017-08-01

We propose to search for light U (1 ) dark photons, A', produced via kinetically mixing with ordinary photons via the Compton-like process, γ e-→A'e-, in a nuclear reactor and detected by their interactions with the material in the active volumes of reactor neutrino experiments. We derive 95% confidence-level upper limits on ɛ , the A'-γ mixing parameter, ɛ , for dark-photon masses below 1 MeV of ɛ <1.3 ×10-5 and ɛ <2.1 ×10-5, from NEOS and TEXONO experimental data, respectively. This study demonstrates the applicability of nuclear reactors as potential sources of intense fluxes of low-mass dark photons.

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.

Large-scale boiling experiments of the flooded cavity concept for in-vessel core retention

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

Chu, T.Y.; Slezak, S.E.; Bentz, J.H.

1994-03-01

This paper presents results of ex-vessel boiling experiments performed in the CYBL (CYlindrical BoiLing) facility. CYBL is a reactor-scale facility for confirmatory research of the flooded cavity concept for accident management. CYBL has a tank-within-a-tank design; the inner tank simulates the reactor vessel and the outer tank simulates the reactor cavity. Experiments with uniform and edge-peaked heat flux distributions up to 20 W/cm{sup 2} across the vessel bottom were performed. Boiling outside the reactor vessel was found to be subcooled nucleate boiling. The subcooling is mainly due to the gravity head which results from flooding the sides of the reactormore » vessel. The boiling process exhibits a cyclic pattern with four distinct phases: direct liquid/solid contact, bubble nucleation and growth, coalescence, and vapor mass dispersion (ejection). The results suggest that under prototypic heat load and heat flux distributions, the flooded cavity in a passive pressurized water reactor like the AP-600 should be capable of cooling the reactor pressure vessel in the central region of the lower head that is addressed by these tests.« less

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.

The Fukushima Nuclear Disaster and the U.S. Customs and Border Protection Response

NASA Astrophysics Data System (ADS)

McCormick, Kathy

2013-10-01

On 3/11/11, the reactors at the Fukushima Nuclear Plant in Japan were damaged by a magnitude 9.0 earthquake. Of the six reactors at the site, three were in operation prior to the event, and were automatically shut-down during the earthquake. Emergency cooling systems came online and were subsequently destroyed by a tsunami generated by the earthquake. For the operating reactors, all the reactor cores were exposed, resulting in overheating and the release of steam and hydrogen gas to the containment vessels, several of which subsequently exploded, releasing radioactivity into the atmosphere. The cores of the operating reactors melted down, and radioactive water was released to the ocean in cooling efforts. The primary radiation concerns in the United States from the disaster were radioactive plumes driven by westerly winds and contaminated commercial products and travelers. In the United States, one of the primary governmental organizations to respond to the disaster was U.S. Customs and Border Protection (CBP), which has responsibility to oversee the safety and security of cargo and travelers entering the United States. This talk will describe the various types of radioactive commodities and events encountered by CBP in the U.S. from the Fukushima disaster. Thanks to the CBP Teleforensics Center for their assistance with this presentation.

Summary of reactor plant conditions during L2-2 pre-LOCE maneuver

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

Tsang, F.Y.; Yarbrough, W.M.; Cannon, J.W.

1979-04-26

This document presents the experimental results obtained during the pre-Loss of Coolant Experiment (LOCE) manuever and the core conditions prior to the L2-2 LOCE. The peak linear heat rate prior to the blowdown was 8.04 kW/ft, the primary coolant mass flow rate was 1.539 Mlbm/hr, the hot leg temperature was 585.9/sup 0/F, and the core ..delta..T was 42/sup 0/F. These conditions satisfied the requirements specified for the L2-2 LOCE except for the hot leg temperature being 12/sup 0/F below the desired 598/sup 0/F.

Experience with ALARA and ALARA procedures in a nuclear power plant

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

Abrahamse, J.C.

1995-03-01

The nuclear power plant Borssele is a Siemens two-loop Pressurized Water Reactor having a capacity of 480 MWe and in operation since 1973. The nuclear power plant Borssle is located in the southwest of the Netherlands, near the Westerschelde River. In the first nine years of operation the radiation level in the primary system increased, reaching a maximum in 1983. The most important reason for this high radiation level was the cobalt content of the grid assemblies of the fuel elements. After resolving this problem, the radiation level decreased to a level comparable with that of other nuclear power plants.

High-solid mesophilic methane fermentation of food waste with an emphasis on Iron, Cobalt, and Nickel requirements.

PubMed

Qiang, Hong; Lang, Dong-Li; Li, Yu-You

2012-01-01

The effect of trace metals on the mesophilic methane fermentation of high-solid food waste was investigated using both batch and continuous experiments. The continuous experiment was conducted by using a CSTR-type reactor with three run. During the first run, the HRT of the reactor was stepwise decreased from 100 days to 30 days. From operation day 50, the reactor efficiency deteriorated due to the lack of trace metals. The batch experiment showed that iron, cobalt, and nickel combinations had a significant effect on food waste. According to the results of the batch experiment, a combination of iron, cobalt, and nickel was added into the CSTR reactor by two different methods at run II, and III. Based on experimental results and theoretical calculations, the most suitable values of Fe/COD, Co/COD, and Ni/COD in the substrate were identified as 200, 6.0, and 5.7 mg/kg COD, respectively. Copyright © 2011 Elsevier Ltd. All rights reserved.

Impact of fine mesh sieve primary treatment on nitrogen removal in moving bed biofilm reactors.

PubMed

Rusten, B; Razafimanantsoa, V A; Andriamiarinjaka, M A; Otis, C L; Sahu, A K; Bilstad, T

2016-01-01

The purpose of this project was to investigate the effect of selective particle removal during primary treatment on nitrogen removal in moving bed biofilm reactors (MBBRs). Two small MBBR pilot plants were operated in parallel, where one train treated 2 mm screened municipal wastewater and the other train treated wastewater that had passed through a Salsnes Filter SF1000 rotating belt sieve (RBS) with a 33 µs sieve cloth. The SF1000 was operated without a filter mat on the belt. The tests confirmed that, for the wastewater characteristics at the test plant, Salsnes Filter primary treatment with a 33 µs RBS and no filter mat produced a primary effluent that was close to optimum. Removal of organic matter with the 33 µs sieve had no negative effect on the denitrification process. Nitrification rates improved by 10-15% in the train with 33 µs RBS primary treatment. Mass balance calculations showed that without RBS primary treatment, the oxygen demand in the biological system was 36% higher. Other studies have shown that the sludge produced by RBS primary treatment is beneficial for biogas production and will also significantly improve sludge dewatering of the combined primary and biological sludge.

Response of a continuous anaerobic digester to temperature transitions: A critical range for restructuring the microbial community structure and function.

PubMed

Kim, Jaai; Lee, Changsoo

2016-02-01

Temperature is a crucial factor that significantly influences the microbial activity and so the methanation performance of an anaerobic digestion (AD) process. Therefore, how to control the operating temperature for optimal activity of the microbes involved is a key to stable AD. This study examined the response of a continuous anaerobic reactor to a series of temperature shifts over a wide range of 35-65 °C using a dairy-processing byproduct as model wastewater. During the long-term experiment for approximately 16 months, the reactor was subjected to stepwise temperature increases by 5 °C at a fixed HRT of 15 days. The reactor showed stable performance within the temperature range of 35-45 °C, with the methane production rate and yield being maximum at 45 °C (18% and 26% greater, respectively, than at 35 °C). However, the subsequent increase to 50 °C induced a sudden performance deterioration with a complete cessation of methane recovery, indicating that the temperature range between 45 °C and 50 °C had a critical impact on the transition of the reactor's methanogenic activity from mesophilic to thermophilic. This serious process perturbation was associated with a severe restructuring of the reactor microbial community structure, particularly of methanogens, quantitatively as well as qualitatively. Once restored by interrupted feeding for about two months, the reactor maintained fairly stable performance under thermophilic conditions until it was upset again at 65 °C. Interestingly, in contrast to most previous reports, hydrogenotrophs largely dominated the methanogen community at mesophilic temperatures while acetotrophs emerged as a major group at thermophilic temperature. This implies that the primary methanogenesis route of the reactor shifted from hydrogen- to acetate-utilizing pathways with the temperature shifts from mesophilic to thermophilic temperatures. Our observations suggest that a mesophilic digester may not need to be cooled at up to 45 °C in case of undesired temperature rise, for example, by excessive self-heating, which offers a possibility to reduce operating costs. Copyright © 2015 Elsevier Ltd. All rights reserved.

Basic Nuclear Physics.

ERIC Educational Resources Information Center

Bureau of Naval Personnel, Washington, DC.

Basic concepts of nuclear structures, radiation, nuclear reactions, and health physics are presented in this text, prepared for naval officers. Applications to the area of nuclear power are described in connection with pressurized water reactors, experimental boiling water reactors, homogeneous reactor experiments, and experimental breeder…

Summary of ORSphere Critical and Reactor Physics Measurements

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

Marshall, Margaret A.; Bess, John D.

In the early 1970s Dr. John T. Mihalczo (team leader), J. J. Lynn, and J. R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s. The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with the GODIVAmore » I experiments. This critical configuration has been evaluated. Preliminary results were presented at ND2013. Since then, the evaluation was finalized and judged to be an acceptable benchmark experiment for the International Criticality Safety Benchmark Experiment Project (ICSBEP). Additionally, reactor physics measurements were performed to determine surface button worths, central void worth, delayed neutron fraction, prompt neutron decay constant, fission density and neutron importance. These measurements have been evaluated and found to be acceptable experiments and are discussed in full detail in the International Handbook of Evaluated Reactor Physics Benchmark Experiments. The purpose of this paper is summary summarize all the critical and reactor physics measurements evaluations and, when possible, to compare them to GODIVA experiment results.« less

Kilowatt Reactor Using Stirling TechnologY (KRUSTY) Demonstration. CEDT Phase 1 Preliminary Design Documentation

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

Sanchez, Rene Gerardo; Hutchinson, Jesson D.; Mcclure, Patrick Ray

2015-08-20

The intent of the integral experiment request IER 299 (called KiloPower by NASA) is to assemble and evaluate the operational performance of a compact reactor configuration that closely resembles the flight unit to be used by NASA to execute a deep space exploration mission. The reactor design will include heat pipes coupled to Stirling engines to demonstrate how one can generate electricity when extracting energy from a “nuclear generated” heat source. This series of experiments is a larger scale follow up to the DUFF series of experiments1,2 that were performed using the Flat-Top assembly.

Comparative study on neutron data in integral experiments of MYRRHA mockup critical cores in the VENUS-F reactor

NASA Astrophysics Data System (ADS)

Krása, Antonín; Kochetkov, Anatoly; Baeten, Peter; Vittiglio, Guido; Wagemans, Jan; Bécares, Vicente

2017-09-01

VENUS-F is a fast, zero-power reactor with 30% wt. metallic uranium fuel and solid lead as coolant simulator. It serves as a mockup of the MYRRHA reactor core. This paper describes integral experiments performed in two critical VENUS-F core configurations (with and without graphite reflector). Discrepancies between experiments and Monte Carlo calculations (MCNP5) of keff, fission rate spatial distribution and reactivity effects (lead void and fuel Doppler) depending on a nuclear data library used (JENDL-4.0, ENDF-B-VII.1, JEFF-3.1.2, 3.2, 3.3T2) are presented.

Ultrahigh temperature vapor core reactor-MHD system for space nuclear electric power

NASA Technical Reports Server (NTRS)

Maya, Isaac; Anghaie, Samim; Diaz, Nils J.; Dugan, Edward T.

1991-01-01

The conceptual design of a nuclear space power system based on the ultrahigh temperature vapor core reactor with MHD energy conversion is presented. This UF4 fueled gas core cavity reactor operates at 4000 K maximum core temperature and 40 atm. Materials experiments, conducted with UF4 up to 2200 K, demonstrate acceptable compatibility with tungsten-molybdenum-, and carbon-based materials. The supporting nuclear, heat transfer, fluid flow and MHD analysis, and fissioning plasma physics experiments are also discussed.

Characterization of the Caliban and Prospero Critical Assemblies Neutron Spectra for Integral Measurements Experiments

NASA Astrophysics Data System (ADS)

Casoli, P.; Authier, N.; Jacquet, X.; Cartier, J.

2014-04-01

Caliban and Prospero are two highly enriched uranium metallic core reactors operated on the CEA Center of Valduc. These critical assemblies are suitable for integral experiments, such as fission yields measurements or perturbation measurements, which have been carried out recently on the Caliban reactor. Different unfolding methods, based on activation foils and fission chambers measurements, are used to characterize the reactor spectra and especially the Caliban spectrum, which is very close to a pure fission spectrum.

Methods and systems for the production of hydrogen

DOEpatents

Oh, Chang H [Idaho Falls, ID; Kim, Eung S [Ammon, ID; Sherman, Steven R [Augusta, GA

2012-03-13

Methods and systems are disclosed for the production of hydrogen and the use of high-temperature heat sources in energy conversion. In one embodiment, a primary loop may include a nuclear reactor utilizing a molten salt or helium as a coolant. The nuclear reactor may provide heat energy to a power generation loop for production of electrical energy. For example, a supercritical carbon dioxide fluid may be heated by the nuclear reactor via the molten salt and then expanded in a turbine to drive a generator. An intermediate heat exchange loop may also be thermally coupled with the primary loop and provide heat energy to one or more hydrogen production facilities. A portion of the hydrogen produced by the hydrogen production facility may be diverted to a combustor to elevate the temperature of water being split into hydrogen and oxygen by the hydrogen production facility.

A simulator-based nuclear reactor emergency response training exercise.

PubMed

Waller, Edward; Bereznai, George; Shaw, John; Chaput, Joseph; Lafortune, Jean-Francois

Training offsite emergency response personnel basic awareness of onsite control room operations during nuclear power plant emergency conditions was the primary objective of a week-long workshop conducted on a CANDU® virtual nuclear reactor simulator available at the University of Ontario Institute of Technology, Oshawa, Canada. The workshop was designed to examine both normal and abnormal reactor operating conditions, and to observe the conditions in the control room that may have impact on the subsequent offsite emergency response. The workshop was attended by participants from a number of countries encompassing diverse job functions related to nuclear emergency response. Objectives of the workshop were to provide opportunities for participants to act in the roles of control room personnel under different reactor operating scenarios, providing a unique experience for participants to interact with the simulator in real-time, and providing increased awareness of control room operations during accident conditions. The ability to "pause" the simulator during exercises allowed the instructors to evaluate and critique the performance of participants, and to provide context with respect to potential offsite emergency actions. Feedback from the participants highlighted (i) advantages of observing and participating "hands-on" with operational exercises, (ii) their general unfamiliarity with control room operational procedures and arrangements prior to the workshop, (iii) awareness of the vast quantity of detailed control room procedures for both normal and transient conditions, and (iv) appreciation of the increased workload for the operators in the control room during a transient from normal operations. Based upon participant feedback, it was determined that the objectives of the training had been met, and that future workshops should be conducted.

Experiments and models of general corrosion and flow-assisted corrosion of materials in nuclear reactor environments

NASA Astrophysics Data System (ADS)

Cook, William Gordon

Corrosion and material degradation issues are of concern to all industries. However, the nuclear power industry must conform to more stringent construction, fabrication and operational guidelines due to the perceived additional risk of operating with radioactive components. Thus corrosion and material integrity are of considerable concern for the operators of nuclear power plants and the bodies that govern their operations. In order to keep corrosion low and maintain adequate material integrity, knowledge of the processes that govern the material's breakdown and failure in a given environment are essential. The work presented here details the current understanding of the general corrosion of stainless steel and carbon steel in nuclear reactor primary heat transport systems (PHTS) and examines the mechanisms and possible mitigation techniques for flow-assisted corrosion (FAC) in CANDU outlet feeder pipes. Mechanistic models have been developed based on first principles and a 'solution-pores' mechanism of metal corrosion. The models predict corrosion rates and material transport in the PHTS of a pressurized water reactor (PWR) and the influence of electrochemistry on the corrosion and flow-assisted corrosion of carbon steel in the CANDU outlet feeders. In-situ probes, based on an electrical resistance technique, were developed to measure the real-time corrosion rate of reactor materials in high-temperature water. The probes were used to evaluate the effects of coolant pH and flow on FAC of carbon steel as well as demonstrate of the use of titanium dioxide as a coolant additive to mitigated FAC in CANDU outlet feeder pipes.

Accelerator and reactor complementarity in coherent neutrino-nucleus scattering

NASA Astrophysics Data System (ADS)

Dent, James B.; Dutta, Bhaskar; Liao, Shu; Newstead, Jayden L.; Strigari, Louis E.; Walker, Joel W.

2018-02-01

We study the complementarity between accelerator and reactor coherent elastic neutrino-nucleus elastic scattering (CE ν NS ) experiments for constraining new physics in the form of nonstandard neutrino interactions (NSI). First, considering just data from the recent observation by the Coherent experiment, we explore interpretive degeneracies that emerge when activating either two or four unknown NSI parameters. Next, we demonstrate that simultaneous treatment of reactor and accelerator experiments, each employing at least two distinct target materials, can break a degeneracy between up and down flavor-diagonal NSI terms that survives analysis of neutrino oscillation experiments. Considering four flavor-diagonal (e e /μ μ ) up- and down-type NSI parameters, we find that all terms can be measured with high local precision (to a width as small as ˜5 % in Fermi units) by next-generation experiments, although discrete reflection ambiguities persist.

Analog to digital converter system for temperature monitoring -- B, C, D, DR, F, and H reactors

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

Ballowe, J.W.

1961-03-23

This document discusses a proposal that certain presently installed reactor process water outlet temperature data logging equipment in subject reactors to be replaced with new functionally simplified equipment of a more adequate design. The primary purpose of the proposed installation is to replace existing equipment which is obsolete and in three reactors is worn out to the point where the equipment is out of service frequently for periods of time up to 8 hours or more. The new equipment will provide reliable process tube temperature information for use in the functions of reactor control and product accountability. Based upon anticipatedmore » incremental production gains resulting from use of the new equipment, the amortization period for the project is calculated at 2.7 years.« less

Optimization of 200 MWth and 250 MWt Ship Based Small Long Life NPP

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

Fitriyani, Dian; Su'ud, Zaki

2010-06-22

Design optimization of ship-based 200 MWth and 250 MWt nuclear power reactors have been performed. The neutronic and thermo-hydraulic programs of the three-dimensional X-Y-Z geometry have been developed for the analysis of ship-based nuclear power plant. Quasi-static approach is adopted to treat seawater effect. The reactor are loop type lead bismuth cooled fast reactor with nitride fuel and with relatively large coolant pipe above reactor core, the heat from primary coolant system is directly transferred to watersteam loop through steam generators. Square core type are selected and optimized. As the optimization result, the core outlet temperature distribution is changing withmore » the elevation angle of the reactor system and the characteristics are discussed.« less

Evaluation of anticipatory signal to steam generator pressure control program for 700 MWe Indian pressurized heavy water reactor

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

Pahari, S.; Hajela, S.; Rammohan, H. P.

2012-07-01

700 MWe Indian Pressurized Heavy Water Reactor (IPHWR) is horizontal channel type reactor with partial boiling at channel outlet. Due to boiling, it has a large volume of vapor present in the primary loops. It has two primary loops connected with the help of pressurizer surge line. The pressurizer has a large capacity and is partly filled by liquid and partly by vapor. Large vapor volume improves compressibility of the system. During turbine trip or load rejection, pressure builds up in Steam Generator (SG). This leads to pressurization of Primary Heat Transport System (PHTS). To control pressurization of SG andmore » PHTS, around 70% of the steam generated in SG is dumped into the condenser by opening Condenser Steam Dump Valves (CSDVs) and rest of the steam is released to the atmosphere by opening Atmospheric Steam Discharge Valves (ASDVs) immediately after sensing the event. This is accomplished by adding anticipatory signal to the output of SG pressure controller. Anticipatory signal is proportional to the thermal power of reactor and the proportionality constant is set so that SG pressure controller's output jacks up to ASDV opening range when operating at 100% FP. To simulate this behavior for 700 MWe IPHWR, Primary and secondary heat transport system is modeled. SG pressure control and other process control program have also been modeled to capture overall plant dynamics. Analysis has been carried out with 3-D neutron kinetics coupled thermal hydraulic computer code ATMIKA.T to evaluate the effect of the anticipatory signal on PHT pressure and over all plant dynamics during turbine trip in 700 MWe IPHWR. This paper brings out the results of the analysis with and without considering anticipatory signal in SG pressure control program during turbine trip. (authors)« less

Micropollutant removal from black water and grey water sludge in a UASB-GAC reactor.

PubMed

Butkovskyi, A; Sevenou, L; Meulepas, R J W; Hernandez Leal, L; Zeeman, G; Rijnaarts, H H M

2018-02-01

The effect of granular activated carbon (GAC) addition on the removal of diclofenac, ibuprofen, metoprolol, galaxolide and triclosan in a up-flow anaerobic sludge blanket (UASB) reactor was studied. Prior to the reactor studies, batch experiments indicated that addition of activated carbon to UASB sludge can decrease micropollutant concentrations in both liquid phase and sludge. In continuous experiments, two UASB reactors were operated for 260 days at an HRT of 20 days, using a mixture of source separated black water and sludge from aerobic grey water treatment as influent. GAC (5.7 g per liter of reactor volume) was added to one of the reactors on day 138. No significant difference in COD removal and biogas production between reactors with and without GAC addition was observed. In the presence of GAC, fewer micropollutants were washed out with the effluent and a lower accumulation of micropollutants in sludge and particulate organic matter occurred, which is an advantage in micropollutant emission reduction from wastewater. However, the removal of micropollutants by adding GAC to a UASB reactor would require more activated carbon compared to effluent post-treatment. Additional research is needed to estimate the effect of bioregeneration on the lifetime of activated carbon in a UASB-GAC reactor.

Study of dietary supplements compositions by neutron activation analysis at the VR-1 training reactor

NASA Astrophysics Data System (ADS)

Stefanik, Milan; Rataj, Jan; Huml, Ondrej; Sklenka, Lubomir

2017-11-01

The VR-1 training reactor operated by the Czech Technical University in Prague is utilized mainly for education of students and training of various reactor staff; however, R&D is also carried out at the reactor. The experimental instrumentation of the reactor can be used for the irradiation experiments and neutron activation analysis. In this paper, the neutron activation analysis (NAA) is used for a study of dietary supplements containing the zinc (one of the essential trace elements for the human body). This analysis includes the dietary supplement pills of different brands; each brand is represented by several different batches of pills. All pills were irradiated together with the standard activation etalons in the vertical channel of the VR-1 reactor at the nominal power (80 W). Activated samples were investigated by the nuclear gamma-ray spectrometry technique employing the semiconductor HPGe detector. From resulting saturated activities, the amount of mineral element (Zn) in the pills was determined using the comparative NAA method. The results show clearly that the VR-1 training reactor is utilizable for neutron activation analysis experiments.

Utilization of solar energy in sewage sludge composting: Fertilizer effect and application

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

Chen, Yiqun; Yu, Fang; Liang, Shengwen

2014-11-15

Highlights: • Solar energy technologies were utilized in aerobic sewage sludge composting. • Greenhouse and solar reactors were constructed to compare impacts on the composting. • Impatiens balsamina was planted in pot experiments to evaluate fertilizer effect. - Abstract: Three reactors, ordinary, greenhouse, and solar, were constructed and tested to compare their impacts on the composting of municipal sewage sludge. Greenhouse and solar reactors were designed to evaluate the use of solar energy in sludge composting, including their effects on temperature and compost quality. After 40 days of composting, it was found that the solar reactor could provide more stablemore » heat for the composting process. The average temperature of the solar reactor was higher than that of the other two systems, and only the solar reactor could maintain the temperature above 55 °C for more than 3 days. Composting with the solar reactor resulted in 31.3% decrease in the total organic carbon, increased the germination index to 91%, decreased the total nitrogen loss, and produced a good effect on pot experiments.« less

Experience of on-site disposal of production uranium-graphite nuclear reactor.

PubMed

Pavliuk, Alexander O; Kotlyarevskiy, Sergey G; Bespala, Evgeny V; Zakharova, Elena V; Ermolaev, Vyacheslav M; Volkova, Anna G

2018-04-01

The paper reported the experience gained in the course of decommissioning EI-2 Production Uranium-Graphite Nuclear Reactor. EI-2 was a production Uranium-Graphite Nuclear Reactor located on the Production and Demonstration Center for Uranium-Graphite Reactors JSC (PDC UGR JSC) site of Seversk City, Tomsk Region, Russia. EI-2 commenced its operation in 1958, and was shut down on December 28, 1990, having operated for the period of 33 years all together. The extra pure grade graphite for the moderator, water for the coolant, and uranium metal for the fuel were used in the reactor. During the operation nitrogen gas was passed through the graphite stack of the reactor. In the process of decommissioning the PDC UGR JSC site the cavities in the reactor space were filled with clay-based materials. A specific composite barrier material based on clays and minerals of Siberian Region was developed for the purpose. Numerical modeling demonstrated the developed clay composite would make efficient geological barriers preventing release of radionuclides into the environment. Copyright © 2018 Elsevier Ltd. All rights reserved.

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.

Interface requirements for coupling a containment code to a reactor system thermal hydraulic codes

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

Baratta, A.J.

1997-07-01

To perform a complete analysis of a reactor transient, not only the primary system response but the containment response must also be accounted for. Such transients and accidents as a loss of coolant accident in both pressurized water and boiling water reactors and inadvertent operation of safety relief valves all challenge the containment and may influence flows because of containment feedback. More recently, the advanced reactor designs put forth by General Electric and Westinghouse in the US and by Framatome and Seimens in Europe rely on the containment to act as the ultimate heat sink. Techniques used by analysts andmore » engineers to analyze the interaction of the containment and the primary system were usually iterative in nature. Codes such as RELAP or RETRAN were used to analyze the primary system response and CONTAIN or CONTEMPT the containment response. The analysis was performed by first running the system code and representing the containment as a fixed pressure boundary condition. The flows were usually from the primary system to the containment initially and generally under choked conditions. Once the mass flows and timing are determined from the system codes, these conditions were input into the containment code. The resulting pressures and temperatures were then calculated and the containment performance analyzed. The disadvantage of this approach becomes evident when one performs an analysis of a rapid depressurization or a long term accident sequence in which feedback from the containment can occur. For example, in a BWR main steam line break transient, the containment heats up and becomes a source of energy for the primary system. Recent advances in programming and computer technology are available to provide an alternative approach. The author and other researchers have developed linkage codes capable of transferring data between codes at each time step allowing discrete codes to be coupled together.« less

10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

Code of Federal Regulations, 2014 CFR

2014-01-01

... air test pressure and to assure they will be subjected to the post accident differential pressure... Table of Contents I. Introduction. II. Explanation of terms. III. Leakage test requirements. A. Type A test. B. Type B test. C. Type C test. D. Periodic retest schedule. IV. Special test requirements. A...

10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

Code of Federal Regulations, 2012 CFR

2012-01-01

... air test pressure and to assure they will be subjected to the post accident differential pressure... Table of Contents I. Introduction. II. Explanation of terms. III. Leakage test requirements. A. Type A test. B. Type B test. C. Type C test. D. Periodic retest schedule. IV. Special test requirements. A...

10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

Code of Federal Regulations, 2010 CFR

2010-01-01

... air test pressure and to assure they will be subjected to the post accident differential pressure... Table of Contents I. Introduction. II. Explanation of terms. III. Leakage test requirements. A. Type A test. B. Type B test. C. Type C test. D. Periodic retest schedule. IV. Special test requirements. A...

10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

Code of Federal Regulations, 2013 CFR

2013-01-01

... air test pressure and to assure they will be subjected to the post accident differential pressure... Table of Contents I. Introduction. II. Explanation of terms. III. Leakage test requirements. A. Type A test. B. Type B test. C. Type C test. D. Periodic retest schedule. IV. Special test requirements. A...

10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

Code of Federal Regulations, 2011 CFR

2011-01-01

.... Containment inspection. B. Repordkeeping of test results. I. Introduction One of the conditions of all... following: A. Type A test—1. Pretest requirements. (a) Containment inspection in accordance with V. A. shall.... During the period between the completion of one Type A test and the initiation of the containment...

75 FR 54657 - University of Florida; University of Florida Training Reactor; Environmental Assessment and...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-09-08

... operation of the UFTR to routinely provide teaching, research, and services to numerous institutions for a... confinement. The Nuclear Reactor Building and its annex, the Nuclear Sciences Center, are located in an area... primary system consisting of a 200-gallon coolant storage tank, a heat removal system, and a processing...

Reactor experiments to study luminescence of He-Ne and He-Kr gaseous mixtures, excited by the products of 6Li (n, α) 3H nuclear reaction

NASA Astrophysics Data System (ADS)

Batyrbekov, E. G.; Gordienko, Yu. N.; Barsukov, N. I.; Ponkratov, Yu. V.; Kulsartov, T. V.; Khassenov, M. U.; Zaurbekova, Zh. A.; Tulubayev, Ye. Y.; Samarkhanov, K. K.

2018-04-01

The spectral studies of optical radiation of gaseous mixtures are of interest for solving problems associated with finding gaseous media with high energy conversion efficiency of nuclear reactions into the energy of laser or spontaneous emission [1, 2]. Such media can be used to extract energy from nuclear and fusion reactors in the form of optical radiation, and also to control and adjust the nuclear reactors parameters. This paper presents the preliminary results of the reactor experiments to study the spectral-luminescent properties of gas mixtures (based on He, Ne and Kr noble gases) excited by the products of 6Li(n,α)3H nuclear reaction at different levels of the stationary power of the IVG.1M reactor.

Analysis of the Browns Ferry Unit 3 irradiation experiments. Final report

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

Simmons, G.L.

1984-11-01

The results of the analysis of two experiments performed at the Browns Ferry-3 reactor are presented. These calculations utilize state-of-the-art neutron transport techniques and a new neutron cross-section library that has been developed for LWR applications. The calculations agree well with the experimental data obtained in irradiations inside the reactor vessel. For the measurements performed in the reactor cavity, the calculations agree well at the reactor midplane. Accurate determination of the axial distribution of the neutron fluence in the reactor cavity depends on having a concise representation of the axial-void distribution in the core. Detailed data are presented describing themore » procedures used in the generation of the new cross-section library that has been named SAILOR. This library is available from the Radiation-Shielding Information Center.« less

Report detailing comparative analysis of results from high flux isotope reactor and national institute of standards technology small-angle neutron scattering experiments

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

Sokolov, Mikhail A.; Littrell, Ken; Wells, Peter

The major issues regarding irradiation effects are discussed in [1-3] and have also been discussed in previous progress and milestone reports. As noted previously, of the many significant issues discussed, the issue considered to have the most impact on the current regulatory process is that associated with effects of neutron irradiation on RPV steels at high fluence, for long irradiation times, and as affected by neutron flux. It is clear that embrittlement of RPV steels is a critical issue that may limit LWR plant life extension. The primary objective of the LWRSP RPV task is to develop robust predictions ofmore » transition temperature shifts (TTS) at high fluence ( t) to at least 1020 n/cm 2 (>1 MeV) pertinent to plant operation of some pressurized water reactors (PWR) for 80 full power years. Correlations between the high flux test reactor results and low flux surveillance specimens must be established for proper RPV embrittlement predictions of the current nuclear power fleet. Additionally, a complete understanding of defect evolution for high nickel RPV steels is needed to characterize the embrittlement potential of Mn-Ni-enriched precipitates (MNPs), particularly for the high fluence regime. While understanding of copper-enriched precipitates (CRPs) have been fully developed, the recent discovery and experimental verification [4] of late blooming MNPs with little to no copper for nucleation has stimulated research efforts to understand the evolution of these phases. New and existing databases will be combined to support developing physically based models of TTS for high fluence-low flux ( < 10 11n/cm 2-s) conditions, beyond the existing surveillance database, to neutron fluences of at least 1 1020 n/cm2 (>1 MeV). Moreover, large number of various RPV materials have been irradiated in ATR-2 experiment and will be jointly studied by University of California Santa Barbara (UCSB) and ORNL to address majority of microstructural characteristics discussed above, see Ref. [5] and [6] for details. UCSB has performed a large number of SANS experiments in the past at the National Institute of Standards and Technology (NIST) Center for Neutron Research (NCNR). These data are taken from RPV steels irradiated in a wide range of flux-fluence space and will be very useful in comparing to the upcoming UCSB ATR-2 irradiation characterization since most of the SANS experiments with ATR-2 materials will be performed at ORNL High Flux Isotope Reactor (HFIR). However in the previous report [7], some discrepancies were observed between HFIR and NCNR generated data. One of the hypotheses was that there was some kind of extra scattering occurring off the sample holders that results in the HFIR curves falling above the NCNR curves. To test this hypothesis, UCSB provided thermally aged samples that have been previously run at NCNR to ORNL for testing at HFIR while ORNL performed some improvements to experimental set up at HFIR. This report provides the status for the Level 3 Milestone (M3LW-15OR0402013), Complete report detailing comparative analysis of results from High Flux Isotope Reactor and National Institute of Standards and Technology small-angle neutron scattering experiments. This milestone is associated with small-angle neutron scattering characterization at the High Flux Isotope Reactor of various model alloys that had been previously characterized at NCNR by UCSB.« less

National Environmental Policy Act Hazards Assessment for the TREAT Alternative

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

Boyd D. Christensen; Annette L. Schafer

2013-11-01

This document provides an assessment of hazards as required by the National Environmental Policy Act for the alternative of restarting the reactor at the Transient Reactor Test (TREAT) facility by the Resumption of Transient Testing Program. Potential hazards have been identified and screening level calculations have been conducted to provide estimates of unmitigated dose consequences that could be incurred through this alternative. Consequences considered include those related to use of the TREAT Reactor, experiment assembly handling, and combined events involving both the reactor and experiments. In addition, potential safety structures, systems, and components for processes associated with operating TREAT andmore » onsite handling of nuclear fuels and experiments are listed. If this alternative is selected, a safety basis will be prepared in accordance with 10 CFR 830, “Nuclear Safety Management,” Subpart B, “Safety Basis Requirements.”« less

National Environmental Policy Act Hazards Assessment for the TREAT Alternative

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

Christensen, Boyd D.; Schafer, Annette L.

2014-02-01

This document provides an assessment of hazards as required by the National Environmental Policy Act for the alternative of restarting the reactor at the Transient Reactor Test (TREAT) facility by the Resumption of Transient Testing Program. Potential hazards have been identified and screening level calculations have been conducted to provide estimates of unmitigated dose consequences that could be incurred through this alternative. Consequences considered include those related to use of the TREAT Reactor, experiment assembly handling, and combined events involving both the reactor and experiments. In addition, potential safety structures, systems, and components for processes associated with operating TREAT andmore » onsite handling of nuclear fuels and experiments are listed. If this alternative is selected, a safety basis will be prepared in accordance with 10 CFR 830, “Nuclear Safety Management,” Subpart B, “Safety Basis Requirements.”« less

PRELIMINARY RESULTS OF THE AGC-4 IRRADIATION IN THE ADVANCED TEST REACTOR AND DESIGN OF AGC-5 (HTR16-18469)

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

Davenport, Michael; Petti, D. A.

The United States Department of Energy’s Advanced Reactor Technologies (ART) Program will irradiate up to six nuclear graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The graphite experiments are being irradiated over an approximate eight year period to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Very High Temperature Gasmore » Reactor (VHTR), as well as other future gas reactors. The experiments each consist of a single capsule that contain six stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens are not be subjected to a compressive load during irradiation. The six stacks have differing compressive loads applied to the top half of diametrically opposite pairs of specimen stacks. A seventh specimen stack in the center of the capsule does not have a compressive load. The specimens are being irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There are also samples taken of the sweep gas effluent to measure any oxidation or off-gassing of the specimens that may occur during initial start-up of the experiment. The first experiment, AGC-1, started its irradiation in September 2009, and the irradiation was completed in January 2011. The second experiment, AGC-2, started its irradiation in April 2011 and completed its irradiation in May 2012. The third experiment, AGC-3, started its irradiation in late November 2012 and completed in the April of 2014. AGC-4 is currently being irradiated in the ATR. This paper will briefly discuss the preliminary irradiation results of the AGC-4 experiment, as well as the design of AGC-5.« less

Detecting Dark Photons with Reactor Neutrino Experiments.

PubMed

Park, H K

2017-08-25

We propose to search for light U(1) dark photons, A^{'}, produced via kinetically mixing with ordinary photons via the Compton-like process, γe^{-}→A^{'}e^{-}, in a nuclear reactor and detected by their interactions with the material in the active volumes of reactor neutrino experiments. We derive 95% confidence-level upper limits on ε, the A^{'}-γ mixing parameter, ε, for dark-photon masses below 1 MeV of ε<1.3×10^{-5} and ε<2.1×10^{-5}, from NEOS and TEXONO experimental data, respectively. This study demonstrates the applicability of nuclear reactors as potential sources of intense fluxes of low-mass dark photons.

A comparative study of thermophilic and mesophilic anaerobic co-digestion of food waste and wheat straw: Process stability and microbial community structure shifts.

PubMed

Shi, Xuchuan; Guo, Xianglin; Zuo, Jiane; Wang, Yajiao; Zhang, Mengyu

2018-05-01

Renewable energy recovery from organic solid waste via anaerobic digestion is a promising way to provide sustainable energy supply and eliminate environmental pollution. However, poor efficiency and operational problems hinder its wide application of anaerobic digestion. The effects of two key parameters, i.e. temperature and substrate characteristics on process stability and microbial community structure were studied using two lab-scale anaerobic reactors under thermophilic and mesophilic conditions. Both the reactors were fed with food waste (FW) and wheat straw (WS). The organic loading rates (OLRs) were maintained at a constant level of 3 kg VS/(m 3 ·d). Five different FW:WS substrate ratios were utilized in different operational phases. The synergetic effects of co-digestion improved the stability and performance of the reactors. When FW was mono-digested, both reactors were unstable. The mesophilic reactor eventually failed due to volatile fatty acid accumulation. The thermophilic reactor had better performance compared to mesophilic one. The biogas production rate of the thermophilic reactor was 4.9-14.8% higher than that of mesophilic reactor throughout the experiment. The shifts in microbial community structures throughout the experiment in both thermophilic and mesophilic reactors were investigated. With increasing FW proportions, bacteria belonging to the phylum Thermotogae became predominant in the thermophilic reactor, while the phylum Bacteroidetes was predominant in the mesophilic reactor. The genus Methanosarcina was the predominant methanogen in the thermophilic reactor, while the genus Methanothrix remained predominant in the mesophilic reactor. The methanogenesis pathway shifted from acetoclastic to hydrogenotrophic when the mesophilic reactor experienced perturbations. Moreover, the population of lignocellulose-degrading microorganisms in the thermophilic reactor was higher than those in mesophilic reactor, which explained the better performance of the thermophilic reactor. Copyright © 2018. Published by Elsevier Ltd.

An induction reactor for studying crude-oil oxidation relevant to in situ combustion.

PubMed

Bazargan, Mohammad; Lapene, Alexandre; Chen, Bo; Castanier, Louis M; Kovscek, Anthony R

2013-07-01

In a conventional ramped temperature oxidation kinetics cell experiment, an electrical furnace is used to ramp temperature at a prescribed rate. Thus, the heating rate of a kinetics cell experiment is limited by furnace performance to heating rates of about 0.5-3 °C/min. A new reactor has been designed to overcome this limit. It uses an induction heating method to ramp temperature. Induction heating is fast and easily controlled. The new reactor covers heating rates from 1 to 30 °C/min. This is the first time that the oxidation profiles of a crude oil are available over such a wide range of heating rate. The results from an induction reactor and a conventional kinetics cell at roughly 2 °C/min are compared to illustrate consistency between the two reactors. The results at low heating rate are the same as the conventional kinetics cell. As presented in the paper, the new reactor couples well with the isoconversional method for interpretation of reaction kinetics.

Acidogenic fermentation of iron-enhanced primary sedimentation sludge under different pH conditions for production of volatile fatty acids.

PubMed

Lin, Lin; Li, Xiao-Yan

2018-03-01

Iron-based chemically enhanced primary sedimentation (CEPS) is increasingly adopted for wastewater treatment in mega cities, producing a large amount of sludge (Fe-sludge) with a high content of organics for potential organic resource recovery. In this experimental study, acidogenic fermentation was applied treat FeCl 3 -based CEPS sludge for production of volatile fatty acids (VFAs) at different pHs. Batch fermentation tests on the Fe-sludge with an organic content of 10 g-COD/L showed that the maximum VFAs production reached 2782.2 mg-COD/L in the reactor without pH control, and it reached 688.4, 3095.3, and 2603.7 mg-COD/L in reactors with pHs kept at 5.0, 6.0 and 8.0, respectively. Analysis of the acidogenesis kinetics and enzymatic activity indicated that the alkaline pH could accelerate the rate of organic hydrolysis but inhibited the further organic conversion to VFAs. In semi-continuous sludge fermentation tests, the VFAs yield in the pH6 reactor was 20% higher than that in the control reactor without pH regulation, while the VFAs yield in the pH8 reactor was 10% lower than the control. Illumina MiSeq sequencing revealed that key functional microorganisms known for effective sludge fermentation, including Bacteroidia and Erysipelotrichi, were enriched in the pH6 reactor with an enhanced VFAs production, while Clostridia became more abundant in the pH8 reactor to stand the unfavorable pH condition. The research presented acidogenic fermentation as an effective process for CEPS sludge treatment and organic resource recovery and provided the first insight into the related microbial community dynamics. Copyright © 2017 Elsevier Ltd. All rights reserved.

INL Experimental Program Roadmap for Thermal Hydraulic Code Validation

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

Glenn McCreery; Hugh McIlroy

2007-09-01

Advanced computer modeling and simulation tools and protocols will be heavily relied on for a wide variety of system studies, engineering design activities, and other aspects of the Next Generation Nuclear Power (NGNP) Very High Temperature Reactor (VHTR), the DOE Global Nuclear Energy Partnership (GNEP), and light-water reactors. The goal is for all modeling and simulation tools to be demonstrated accurate and reliable through a formal Verification and Validation (V&V) process, especially where such tools are to be used to establish safety margins and support regulatory compliance, or to design a system in a manner that reduces the role ofmore » expensive mockups and prototypes. Recent literature identifies specific experimental principles that must be followed in order to insure that experimental data meet the standards required for a “benchmark” database. Even for well conducted experiments, missing experimental details, such as geometrical definition, data reduction procedures, and manufacturing tolerances have led to poor Benchmark calculations. The INL has a long and deep history of research in thermal hydraulics, especially in the 1960s through 1980s when many programs such as LOFT and Semiscle were devoted to light-water reactor safety research, the EBRII fast reactor was in operation, and a strong geothermal energy program was established. The past can serve as a partial guide for reinvigorating thermal hydraulic research at the laboratory. However, new research programs need to fully incorporate modern experimental methods such as measurement techniques using the latest instrumentation, computerized data reduction, and scaling methodology. The path forward for establishing experimental research for code model validation will require benchmark experiments conducted in suitable facilities located at the INL. This document describes thermal hydraulic facility requirements and candidate buildings and presents examples of suitable validation experiments related to VHTRs, sodium-cooled fast reactors, and light-water reactors. These experiments range from relatively low-cost benchtop experiments for investigating individual phenomena to large electrically-heated integral facilities for investigating reactor accidents and transients.« 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

Two-phase reduced gravity experiments for a space reactor design

NASA Technical Reports Server (NTRS)

Antoniak, Zenen I.

1987-01-01

Future space missions researchers envision using large nuclear reactors with either a single or a two-phase alkali-metal working fluid. The design and analysis of such reactors require state-of-the-art computer codes that can properly treat alkali-metal flow and heat transfer in a reduced-gravity environment. New flow regime maps, models, and correlations are required if the codes are to be successfully applied to reduced-gravity flow and heat transfer. General plans are put forth for the reduced-gravity experiments which will have to be performed, at NASA facilities, with benign fluids. Data from the reduced-gravity experiments with innocuous fluids are to be combined with normal gravity data from two-phase alkali-metal experiments. Because these reduced-gravity experiments will be very basic, and will employ small test loops of simple geometry, a large measure of commonality exists between them and experiments planned by other organizations. It is recommended that a committee be formed to coordinate all ongoing and planned reduced gravity flow experiments.

Neutrino parameters from reactor and accelerator neutrino experiments

NASA Astrophysics Data System (ADS)

Lindner, Manfred; Rodejohann, Werner; Xu, Xun-Jie

2018-04-01

We revisit correlations of neutrino oscillation parameters in reactor and long-baseline neutrino oscillation experiments. A framework based on an effective value of θ13 is presented, which can be used to analytically study the correlations and explain some questions including why and when δC P has the best fit value of -π /2 , why current and future long-baseline experiments will have less precision of δC P around ±π /2 than that around zero, etc. Recent hints on the C P phase are then considered from the point of view that different reactor and long-baseline neutrino experiments provide currently different best-fit values of θ13 and θ23. We point out that the significance of the hints changes for the different available best-fit values.

Expert system for surveillance and diagnosis of breach fuel elements

DOEpatents

Gross, K.C.

1988-01-21

An apparatus and method are disclosed for surveillance and diagnosis of breached fuel elements in a nuclear reactor. A delayed neutron monitoring system provides output signals indicating the delayed neutron activity and age and the equivalent recoil area of a breached fuel element. Sensors are used to provide outputs indicating the status of each component of the delayed neutron monitoring system. Detectors also generate output signals indicating the reactor power level and the primary coolant flow rate of the reactor. The outputs from the detectors and sensors are interfaced with an artificial intelligence-based knowledge system which implements predetermined logic and generates output signals indicating the operability of the reactor. 2 figs.

Expert system for surveillance and diagnosis of breach fuel elements

DOEpatents

Gross, Kenny C.

1989-01-01

An apparatus and method are disclosed for surveillance and diagnosis of breached fuel elements in a nuclear reactor. A delayed neutron monitoring system provides output signals indicating the delayed neutron activity and age and the equivalent recoil areas of a breached fuel element. Sensors are used to provide outputs indicating the status of each component of the delayed neutron monitoring system. Detectors also generate output signals indicating the reactor power level and the primary coolant flow rate of the reactor. The outputs from the detectors and sensors are interfaced with an artificial intelligence-based knowledge system which implements predetermined logic and generates output signals indicating the operability of the reactor.

Initial experimental evaluation of crud-resistant materials for light water reactors

NASA Astrophysics Data System (ADS)

Dumnernchanvanit, I.; Zhang, N. Q.; Robertson, S.; Delmore, A.; Carlson, M. B.; Hussey, D.; Short, M. P.

2018-01-01

The buildup of fouling deposits on nuclear fuel rods, known as crud, continues to challenge the worldwide fleet of light water reactors (LWRs). Crud causes 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 assumes that it will always be present. In this study, we report on the development of crud-resistant materials as fuel cladding coatings, to reduce or eliminate these problems altogether. Integrated loop testing experiments at flowing LWR conditions show significantly reduced crud adhesion and surface crud coverage, respectively, for TiC and ZrN coatings compared to ZrO2. The loop testing results roughly agree with the London dispersion component of van der Waals force predictions, suggesting that they contribute most significantly to the adhesion of crud to fuel cladding in out-of-pile conditions. These results motivate a new look at ways of reducing crud, thus avoiding many expensive LWR operational issues.

Assessment of Sensor Technologies for Advanced Reactors

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

Korsah, Kofi; Kisner, R. A.; Britton Jr., C. L.

This paper provides an assessment of sensor technologies and a determination of measurement needs for advanced reactors (AdvRx). It is a summary of a study performed to provide the technical basis for identifying and prioritizing research targets within the instrumentation and control (I&C) Technology Area under the Department of Energy’s (DOE’s) Advanced Reactor Technology (ART) program. The study covered two broad reactor technology categories: High Temperature Reactors and Fast Reactors. The scope of “High temperature reactors” included Gen IV reactors whose coolant exit temperatures exceed ≈650 °C and are moderated (as opposed to fast reactors). To bound the scope formore » fast reactors, this report reviewed relevant operating experience from US-operated Sodium Fast Reactor (SFR) and relevant test experience from the Fast Flux Test Facility (FFTF). For high temperature reactors the study showed that in many cases instrumentation have performed reasonably well in research and demonstration reactors. However, even in cases where the technology is “mature” (such as thermocouples), HTGRs can benefit from improved technologies. Current HTGR instrumentation is generally based on decades-old technology and adapting newer technologies could provide significant advantages. For sodium fast reactors, the study found that several key research needs arise around (1) radiation-tolerant sensor design for in-vessel or in-core applications, where possible non-invasive sensing approaches for key parameters that minimize the need to deploy sensors in-vessel, (2) approaches to exfiltrating data from in-vessel sensors while minimizing penetrations, (3) calibration of sensors in-situ, and (4) optimizing sensor placements to maximize the information content while minimizing the number of sensors needed.« less

Benchmark gas core critical experiment.

NASA Technical Reports Server (NTRS)

Kunze, J. F.; Lofthouse, J. H.; Cooper, C. G.; Hyland, R. E.

1972-01-01

A critical experiment with spherical symmetry has been conducted on the gas core nuclear reactor concept. The nonspherical perturbations in the experiment were evaluated experimentally and produce corrections to the observed eigenvalue of approximately 1% delta k. The reactor consisted of a low density, central uranium hexafluoride gaseous core, surrounded by an annulus of void or low density hydrocarbon, which in turn was surrounded with a 97-cm-thick heavy water reflector.

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

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

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

2011-07-01

The purpose of this document is to provide a guide for using the current version of the Verifiable Fuel Cycle Simulation (VISION) model. This is a complex model with many parameters and options; the user is strongly encouraged to read this user guide before attempting to run the model. This model is an R&D work in progress and may contain errors and omissions. It is based upon numerous assumptions. This model is intended to assist in evaluating 'what if' scenarios and in comparing fuel, reactor, and fuel processing alternatives at a systems level. The model is not intended as amore » tool for process flow and design modeling of specific facilities nor for tracking individual units of fuel or other material through the system. The model is intended to examine the interactions among the components of a fuel system as a function of time varying system parameters; this model represents a dynamic rather than steady-state approximation of the nuclear fuel system. VISION models the nuclear cycle at the system level, not individual facilities, e.g., 'reactor types' not individual reactors and 'separation types' not individual separation plants. Natural uranium can be enriched, which produces enriched uranium, which goes into fuel fabrication, and depleted uranium (DU), which goes into storage. Fuel is transformed (transmuted) in reactors and then goes into a storage buffer. Used fuel can be pulled from storage into either separation or disposal. If sent to separations, fuel is transformed (partitioned) into fuel products, recovered uranium, and various categories of waste. Recycled material is stored until used by its assigned reactor type. VISION is comprised of several Microsoft Excel input files, a Powersim Studio core, and several Microsoft Excel output files. All must be co-located in the same folder on a PC to function. You must use Powersim Studio 8 or better. We have tested VISION with the Studio 8 Expert, Executive, and Education versions. The Expert and Education versions work with the number of reactor types of 3 or less. For more reactor types, the Executive version is currently required. The input files are Excel2003 format (xls). The output files are macro-enabled Excel2007 format (xlsm). VISION 3.4 was designed with more flexibility than previous versions, which were structured for only three reactor types - LWRs that can use only uranium oxide (UOX) fuel, LWRs that can use multiple fuel types (LWR MF), and fast reactors. One could not have, for example, two types of fast reactors concurrently. The new version allows 10 reactor types and any user-defined uranium-plutonium fuel is allowed. (Thorium-based fuels can be input but several features of the model would not work.) The user identifies (by year) the primary fuel to be used for each reactor type. The user can identify for each primary fuel a contingent fuel to use if the primary fuel is not available, e.g., a reactor designated as using mixed oxide fuel (MOX) would have UOX as the contingent fuel. Another example is that a fast reactor using recycled transuranic (TRU) material can be designated as either having or not having appropriately enriched uranium oxide as a contingent fuel. Because of the need to study evolution in recycling and separation strategies, the user can now select the recycling strategy and separation technology, by year.« less

Fate of LCFA in the co-digestion of cow manure, food waste and discontinuous addition of oil.

PubMed

Neves, L; Oliveira, R; Alves, M M

2009-12-01

Different concentrations of oily waste were added in a discontinuous mode and recurrently to anaerobic continuous stirred tank reactors fed with cow manure and food waste. Four continuous stirred tank reactors were run in parallel. A control reactor (R1) received no additional oil and R2, R3 and R4 received increasing concentrations of oil in two different experimental approaches. First, the lipids composition was forced to change suddenly, in three moments, without changing the total chemical oxygen demand (COD) fed to the reactors. The only long chain fatty acid (LCFA) detected onto the R1 solid matrix was palmitic acid (C16:0). Nevertheless in the solid matrix of R2, R3 and R4C16:0 and stearic acid were detected. For occasional increase in the oil concentration up to 7.7gCOD(oil)/L(reactor) (55% Oil(COD)/Total(COD)) no statistical differences were detected between the reactors, in terms of methane production, effluent soluble COD, effluent volatile fatty acids and total and volatile solids removal. Therefore this experiment allowed to conclude that cow manure-food waste co-digestion presents sufficient buffer capacity to endure solid-associated LCFA concentration up to 20-25gCOD-LCFA/kgTS. In a second experiment higher concentrations of oil were added, raising occasionally the concentration in the reactors to 9, 12, 15 and 18gCOD(oil)/L(reactor). All pulses had a positive effect in methane production, with the exception of the highest oil pulse concentration, that persistently impaired the reactor performance. This experiment demonstrates that threshold values for LCFA and C16:0 accumulation onto the solid matrix, of about 180-220gCOD-LCFA/kgTS and 120-150gCOD-C16:0/kgTS, should not be surpassed in order to prevent persistent reactor failure, as occurs in some full scale co-digestion plants.

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

Merzari, E.; Yuan, Haomin; Kraus, A.

The NEAMS program aims to develop an integrated multi-physics simulation capability “pellet-to-plant” for the design and analysis of future generations of nuclear power plants. In particular, the Reactor Product Line code suite's multi-resolution hierarchy is being designed to ultimately span the full range of length and time scales present in relevant reactor design and safety analyses, as well as scale from desktop to petaflop computing platforms. Flow-induced vibration (FIV) is widespread problem in energy systems because they rely on fluid movement for energy conversion. Vibrating structures may be damaged as fatigue or wear occurs. Given the importance of reliable componentsmore » in the nuclear industry, flow-induced vibration has long been a major concern in safety and operation of nuclear reactors. In particular, nuclear fuel rods and steam generators have been known to suffer from flow-induced vibration and related failures. Advanced reactors, such as integral Pressurized Water Reactors (PWRs) considered for Small Modular Reactors (SMR), often rely on innovative component designs to meet cost and safety targets. One component that is the subject of advanced designs is the steam generator, some designs of which forego the usual shell-and-tube architecture in order to fit within the primary vessel. In addition to being more cost- and space-efficient, such steam generators need to be more reliable, since failure of the primary vessel represents a potential loss of coolant and a safety concern. A significant amount of data exists on flow-induced vibration in shell-and-tube heat exchangers, and heuristic methods are available to predict their occurrence based on a set of given assumptions. In contrast, advanced designs have far less data available. Advanced modeling and simulation based on coupled structural and fluid simulations have the potential to predict flow-induced vibration in a variety of designs, reducing the need for expensive experimental programs, especially at the design stage. Over the past five years, the Reactor Product Line has developed the integrated multi-physics code suite SHARP. The goal of developing such a tool is to perform multi-physics neutronics, thermal/fluid, and structural mechanics modeling of the components inside the full reactor core or portions of it with a user-specified fidelity. In particular SHARP contains high-fidelity single-physics codes Diablo for structural mechanics and Nek5000 for fluid mechanics calculations. Both codes are state-of-the-art, highly scalable tools that have been extensively validated. These tools form a strong basis on which to build a flow-induced vibration modeling capability. In this report we discuss one-way coupled calculations performed with Nek5000 and Diablo aimed at simulating available FIV experiments in helical steam generators in the turbulent buffeting regime. In this regime one-way coupling is judged sufficient because the pressure loads do not cause substantial displacements. It is also the most common source of vibration in helical steam generators at the low flows expected in integral PWRs. The legacy data is obtained from two datasets developed at Argonne and B&W.« less

Developments in Sensitivity Methodologies and the Validation of Reactor Physics Calculations

DOE PAGES

Palmiotti, Giuseppe; Salvatores, Massimo

2012-01-01

The sensitivity methodologies have been a remarkable story when adopted in the reactor physics field. Sensitivity coefficients can be used for different objectives like uncertainty estimates, design optimization, determination of target accuracy requirements, adjustment of input parameters, and evaluations of the representativity of an experiment with respect to a reference design configuration. A review of the methods used is provided, and several examples illustrate the success of the methodology in reactor physics. A new application as the improvement of nuclear basic parameters using integral experiments is also described.

Validation of light water reactor calculation methods and JEF-1-based data libraries by TRX and BAPL critical experiments

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

Paratte, J.M.; Pelloni, S.; Grimm, P.

1991-04-01

This paper analyzes the capability of various code systems and JEF-1-based nuclear data libraries to compute light water reactor lattices by comparing calculations with results from thermal reactor benchmark experiments TRX and BAPL and with previously published values. With the JEF-1 evaluation, eigenvalues are generally well predicted within 8 mk (1 mk = 0.001) or less by all code systems, and all methods give reasonable results for the measured reaction rate ratios within, or not too far from, the experimental uncertainty.

Chemical Kinetics, Heat Transfer, and Sensor Dynamics Revisited in a Simple Experiment

ERIC Educational Resources Information Center

Sad, Maria E.; Sad, Mario R.; Castro, Alberto A.; Garetto, Teresita F.

2008-01-01

A simple experiment about thermal effects in chemical reactors is described, which can be used to illustrate chemical reactor models, the determination and validation of their parameters, and some simple principles of heat transfer and sensor dynamics. It is based in the exothermic reaction between aqueous solutions of sodium thiosulfate and…

Vapor phase synthesis of compound semiconductors, from thin films to nanoparticles

NASA Astrophysics Data System (ADS)

Sarigiannis, Demetrius

A counterflow jet reactor was developed to study the gas-phase decomposition kinetics of organometallics used in the vapor phase synthesis of compound semiconductors. The reactor minimized wall effects by generating a reaction zone near the stagnation point of two vertically opposed counterflowing jets. Smoke tracing experiments were used to confirm the stability of the flow field and validate the proposed heat, mass and flow models of the counterflow jet reactor. Transport experiments using ethyl acetate confirmed the overall mass balance for the system and verified the ability of the model to predict concentrations at various points in the reactor under different flow conditions. Preliminary kinetic experiments were performed with ethyl acetate and indicated a need to redesign the reactor. The counterflow jet reactor was adapted for the synthesis of ZnSe nanoparticles. Hydrogen selenide was introduced through one jet and dimethylzinc-triethylamine through the other. The two precursors reacted in a region near the stagnation zone and polycrystalline particles of zinc selenide were reproducibly synthesized at room temperature and collected for analysis. Raman spectroscopy confirmed that the particles were crystalline zinc selenide, Morphological analysis using SEM clearly showed the presence of aggregates of particles, 40 to 60 nanometers in diameter. Analysis by TEM showed that the particles were polycrystalline in nature and composed of smaller single crystalline nanocrystallites, five to ten nanometers in diameter. The particles in the aggregate had the appearance of being sintered together. To prevent this sintering, a split inlet lower jet was designed to introduce dimethylzinc through the inner tube and a surface passivator through the outer one. This passivating agent appeared to prevent the particles from agglomerating. An existing MOVPE reactor for II-VI thin film growth was modified to grow III-V semiconductors. A novel new heater was designed and built around an easily replaceable, economical, 650-watt, tungsten-halogen lamp. The heater was successfully tested to temperatures up to 1500°F. The deposition reactor was successfully tested by growing a thin film of GaP on GaAs <100>. The film surface was imperfect but the experiments proved that the reactor was ready for service.

Cleanup Verification Package for the 118-C-1, 105-C Solid Waste Burial Ground

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

M. J. Appel and J. M. Capron

2007-07-25

This cleanup verification package documents completion of remedial action for the 118-C-1, 105-C Solid Waste Burial Ground. This waste site was the primary burial ground for general wastes from the operation of the 105-C Reactor and received process tubes, aluminum fuel spacers, control rods, reactor hardware, spent nuclear fuel and soft wastes.

Understanding Victims of Technological Disaster: Beliefs and Worries of Three Mile Island.

ERIC Educational Resources Information Center

Prince-Embury, Sandra; Rooney, James

The primary purpose of the present study was to examine how prevalent were concerns about restarting Three Mile Island nuclear reactor Unit I among people within a five-mile radius of the plant four years after the accident involving reactor Unit II. Also explored were concerns related to expectations about the restart of Unit I, perception of…

High-throughput bioconjugation for enhanced 193 nm photodissociation via droplet phase initiated ion/ion chemistry using a front-end dual spray reactor.

PubMed

Cotham, Victoria C; Shaw, Jared B; Brodbelt, Jennifer S

2015-09-15

Fast online chemical derivatization of peptides with an aromatic label for enhanced 193 nm ultraviolet photodissociation (UVPD) is demonstrated using a dual electrospray reactor implemented on the front-end of a linear ion trap (LIT) mass spectrometer. The reactor facilitates the intersection of protonated peptides with a second population of chromogenic 4-formyl-1,3-benzenedisulfonic acid (FBDSA) anions to promote real-time formation of ion/ion complexes at atmospheric pressure. Subsequent collisional activation of the ion/ion intermediate results in Schiff base formation generated via reaction between a primary amine in the peptide cation and the aldehyde moiety of the FBDSA anion. Utilizing 193 nm UVPD as the subsequent activation step in the MS(3) workflow results in acquisition of greater primary sequence information relative to conventional collision induced dissociation (CID). Furthermore, Schiff-base-modified peptides exhibit on average a 20% increase in UVPD efficiency compared to their unmodified counterparts. Due to the efficiency of covalent labeling achieved with the dual spray reactor, we demonstrate that this strategy can be integrated into a high-throughput LC-MS(n) workflow for rapid derivatization of peptide mixtures.

Full reactor coolant system chemical decontamination qualification programs

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

Miller, P.E.

1995-03-01

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

Radiant vessel auxiliary cooling system

DOEpatents

Germer, John H.

1987-01-01

In a modular liquid-metal pool breeder reactor, a radiant vessel auxiliary cooling system is disclosed for removing the residual heat resulting from the shutdown of a reactor by a completely passive heat transfer system. A shell surrounds the reactor and containment vessel, separated from the containment vessel by an air passage. Natural circulation of air is provided by air vents at the lower and upper ends of the shell. Longitudinal, radial and inwardly extending fins extend from the shell into the air passage. The fins are heated by radiation from the containment vessel and convect the heat to the circulating air. Residual heat from the primary reactor vessel is transmitted from the reactor vessel through an inert gas plenum to a guard or containment vessel designed to contain any leaking coolant. The containment vessel is conventional and is surrounded by the shell.

NEUTRONIC REACTOR

DOEpatents

Hurwitz, H. Jr.; Brooks, H.; Mannal, C.; Payne, J.H.; Luebke, E.A.

1959-03-24

A reactor of the heterogeneous, liquid cooled type is described. This reactor is comprised of a central region of a plurality of vertically disposed elongated tubes surrounded by a region of moderator material. The central region is comprised of a central core surrounded by a reflector region which is surrounded by a fast neutron absorber region, which in turn is surrounded by a slow neutron absorber region. Liquid sodium is used as the primary coolant and circulates through the core which contains the fuel elements. Control of the reactor is accomplished by varying the ability of the reflector region to reflect neutrons back into the core of the reactor. For this purpose the reflector is comprised of moderator and control elements having varying effects on reactivity, the control elements being arranged and actuated by groups to give regulation, shim, and safety control.

Preparation of Gd Loaded Liquid Scintillator for Daya Bay Neutrino Experiment

NASA Astrophysics Data System (ADS)

Ya-yun, Ding; Zhi-yong, Zhang

2010-05-01

Gadolinium loaded liquid scintillator (Gd-LS) is an excellent target material for reactor antineutrino experiments. Ideal Gd-LS should have long attenuation length, high light yield, long term stability, low toxicity, and should be compatible with the material used to build the detector. We have developed a new Gd-LS recipe in which carboxylic acid 3,5,5-trimethylhexanoic acid is used as the complexing ligand to gadolinium, 2,5-diphenyloxazole (PPO) and 1,4-bis[2-methylstyryl]benzene (bis-MSB) are used as primary fluor and wavelength shifter, respectively. The scintillator base is linear alkyl benzene (LAB). Eight hundred liters of Gd-LS has been synthesized and tested in a prototype detector. Results show that the Gd-LS has high quality and is suitable for underground experiments in large quantity. Large scale production facility has been built. A full batch production of 4 t Gd-LS has been produced and monitored for several months. The production of 180 t Gd-LS will be carried out in the near future.

Superfund record of decision (EPA region 10): Idaho National Engineering Lab, (USDOE) Operable Unit 26 (Stationary Low-Power Reactor-1 and Boiling Water Reactor Experiment-I Burial Grounds), Idaho Falls, ID, December 1, 1995

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

NONE

1997-03-01

This document presents the selected remedial action for the Stationary Low-Power Reactor-1 (SL-1) burial ground, the Boiling Water Reactor Experiment-I (BORAX-I) burial ground, and 10 no action sites in Waste Area Group 5. Actual or threatened releases of hazardous substances from the SL-1 and BORAX-I burial grounds, if not addressed by implementing the response action selected in this Record of Decision, may present a current or potential threat to public health, welfare, or the environment. The 10 no action sites do not present a threat to human health or the environment.

Modular assembly for supporting, straining, and directing flow to a core in a nuclear reactor

DOEpatents

Pennell, William E.

1977-01-01

A reactor core support arrangement for supporting, straining, and providing fluid flow to the core and periphery of a nuclear reactor during normal operation. A plurality of removable inlet modular units are contained within permanent liners in the lower supporting plate of the reactor vessel lower internals. During normal operation (1) each inlet modular unit directs main coolant flow to a plurality of core assemblies, the latter being removably supported in receptacles in the upper portion of the modular unit and (2) each inlet modular unit may direct bypass flow to a low pressure annular region of the reactor vessel. Each inlet modular unit may include special fluid seals interposed between mating surfaces of the inlet modular units and the core assemblies and between the inlet modular units and the liners, to minimize leakage and achieve an hydraulic balance. Utilizing the hydraulic balance, the modular units are held in the liners and the assemblies are held in the modular unit receptacles by their own respective weight. Included as part of the permanent liners below the horizontal support plate are generally hexagonal axial debris barriers. The axial debris barriers collectively form a bottom boundary of a secondary high pressure plenum, the upper boundary of which is the bottom surface of the horizontal support plate. Peripheral liners include radial debris barriers which collectively form a barrier against debris entry radially. During normal operation primary coolant inlet openings in the liner, below the axial debris barriers, pass a large amount of coolant into the inlet modular units, and secondary coolant inlet openings in the portion of the liners within the secondary plenum pass a small amount of coolant into the inlet modular units. The secondary coolant inlet openings also provide alternative coolant inlet flow paths in the unlikely event of blockage of the primary inlet openings. The primary inlet openings have characteristics which limit the entry of debris and minimize the potential for debris entering the primary inlets blocking the secondary inlets from inside the modular unit.

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.

The influence of perforation of foil reactors on greenhouse gas emission rates during aerobic biostabilization of the undersize fraction of municipal wastes.

PubMed

Stegenta, Sylwia; Dębowski, Marcin; Bukowski, Przemysław; Randerson, Peter F; Białowiec, Andrzej

2018-02-01

The opinion, that the use of foil reactors for the aerobic biostabilization of municipal wastes is not a valid method, due to vulnerability to perforation, and risk of uncontrolled release of exhaust gasses, was verified. This study aimed to determine the intensity of greenhouse gas (GHG) emissions to the atmosphere from the surface of foil reactors in relation to the extent of foil surface perforation. Three scenarios were tested: intact (airtight) foil reactor, perforated foil reactor, and torn foil reactor. Each experimental variant was triplicated, and the duration of each experiment cycle was 5 weeks. Temperature measurements demonstrated a significant decrease in temperature of the biostabilization in the torn reactor. The highest emissions of CO 2 , CO and SO 2 were observed at the beginning of the process, and mostly in the torn reactor. During the whole experiment, observed emissions of CO, H 2 S, NO, NO 2 , and SO 2 were at a very low level which in extreme cases did not exceed 0.25 mg t -1 .h -1 (emission of gasses mass unit per waste mass unit per unit time). The lowest average emissions of greenhouse gases were determined in the case of the intact reactor, which shows that maintaining the foil reactors in an airtight condition during the process is extremely important. Copyright © 2017 Elsevier Ltd. All rights reserved.

LCRE and SNAP 50-DR-1 programs. Engineering progress report, October 1, 1962--December 31, 1962

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

None

Declassified 5 Sep 1973. Information is presented concerning LCRE specifications, reactor kinetics, fuel elements, primary coolant circuit, secondary coolant circuit, materials development, and fabrication; and SNAP50-DR- 1 specifications, primary pump, and materials development. (DCC)

Experiment for search for sterile neutrino at SM-3 reactor

NASA Astrophysics Data System (ADS)

Serebrov, A. P.; Ivochkin, V. G.; Samoylov, R. M.; Fomin, A. K.; Zinoviev, V. G.; Neustroev, P. V.; Golovtsov, V. L.; Gruzinsky, N. V.; Solovey, V. A.; Cherniy, A. V.; Zherebtsov, O. M.; Martemyanov, V. P.; Zinoev, V. G.; Tarasenkov, V. G.; Aleshin, V. I.; Petelin, A. L.; Pavlov, S. V.; Izhutov, A. L.; Sazontov, S. A.; Ryazanov, D. K.; Gromov, M. O.; Afanasiev, V. V.; Matrosov, L. N.; Matrosova, M. Yu.

2016-11-01

In connection with the question of possible existence of sterile neutrino the laboratory on the basis of SM-3 reactor was created to search for oscillations of reactor antineutrino. A prototype of a neutrino detector with scintillator volume of 400 l can be moved at the distance of 6-11 m from the reactor core. The measurements of background conditions have been made. It is shown that the main experimental problem is associated with cosmic radiation background. Test measurements of dependence of a reactor antineutrino flux on the distance from a reactor core have been made. The prospects of search for oscillations of reactor antineutrino at short distances are discussed.

Multidimensional effects in the thermal response of fuel rod simulators. [PWR

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

Dabbs, R.D.; Ott, L.J.

1980-01-01

One of the primary objectives of the Oak Ridge National Laboratory Pressurized-Water Reactor Blowdown Heat Transfer Separate-Effects Program is the determination of the transient surface temperature and surface heat flux of fuel pin simulators (FPSs) from internal thermocouple signals obtained during a loss-of-coolant experiment (LOCE) in the Thermal-Hydraulics Test Facility. This analysis requires the solution of the classical inverse heat conduction problem. The assumptions that allow the governing differential equation to be reduced to one dimension can introduce significant errors in the computed surface heat flux and surface temperature. The degree to which these computed variables are perturbed is addressedmore » and quantified.« less

Corrosion and Corrosion Control in Light Water Reactors

NASA Astrophysics Data System (ADS)

Gordon, Barry M.

2013-08-01

Serious corrosion problems have plagued the light water reactor (LWR) industry for decades. The complex corrosion mechanisms involved and the development of practical engineering solutions for their mitigation will be discussed in this article. After a brief overview of the basic designs of the boiling water reactor (BWR) and pressurized water reactor (PWR), emphasis will be placed on the general corrosion of LWR containments, flow-accelerated corrosion of carbon steel components, intergranular stress corrosion cracking (IGSCC) in BWRs, primary water stress corrosion cracking (PWSCC) in PWRs, and irradiation-assisted stress corrosion cracking (IASCC) in both systems. Finally, the corrosion future of both plants will be discussed as plants extend their period of operation for an additional 20 to 40 years.

Passive heat-transfer means for nuclear reactors. [LMFBR

DOEpatents

Burelbach, J.P.

1982-06-10

An improved passive cooling arrangement is disclosed for maintaining adjacent or related components of a nuclear reactor within specified temperature differences. Specifically, heat pipes are operatively interposed between the components, with the vaporizing section of the heat pipe proximate the hot component operable to cool it and the primary condensing section of the heat pipe proximate the other and cooler component operable to heat it. Each heat pipe further has a secondary condensing section that is located outwardly beyond the reactor confinement and in a secondary heat sink, such as air ambient the containment, that is cooler than the other reactor component. By having many such heat pipes, an emergency passive cooling system is defined that is operative without electrical power.

Low-Temperature Oxidation Reactions and Cool Flames at Earth and Reduced Gravity

NASA Technical Reports Server (NTRS)

Pearlman, Howard

1999-01-01

Non-isothermal studies of cool flames and low temperature oxidation reactions in unstirred closed vessels are complicated by the perturbing effects of natural convection at earth gravity. Buoyant convection due to self-heating during the course of slow reaction produces spatio-temporal variations in the thermal and thus specie concentration fields due to the Arrhenius temperature dependence of the reaction rates. Such complexities have never been quantitatively modeled and were the primary impetus for the development of CSTR's (continuously stirred tank reactors) 30 years ago. While CSTR's have been widely adopted since they offer the advantage of spatial uniformity in temperature and concentration, all gradients are necessarily destroyed along with any structure that may otherwise develop. Microgravity offers a unique environment where buoyant convection can be effectively minimized and the need for stirring eliminated. Moreover, eliminating buoyancy and the need for stirring eliminates complications associated with the induced hydrodynamic field whose influence on heat transport and hot spot formation, hence explosion limits, is not fully realized. The objective of this research is to quantitatively determine and understand the fundamental mechanisms that control the onset and evolution of low temperature reactions and cool flames in both static and flow reactors. Microgravity experiments will be conducted to obtain benchmark data on the structure (spatio-temporal temperature, concentration, flow fields), the dynamics of the chemical fronts, and the ignition diagrams (pressure vs. temperature). Ground-based experiments will be conducted to ascertain the role of buoyancy. Numerical simulations including detailed kinetics will be conducted and compared to experiment.

A SURVEY OF CONVENTIONAL STEAM BOILER EXPERIENCE APPLICABLE TO THE HTGR STEAM GENERATORS

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

Paget, J.A.

1959-10-01

BS>The steam generator of a high temperature gas-cooled reactor consists of tubular heating surface inside a shell which forms part of the primary He circuit of the reactor. When a tube fails in such a steam generator, moisture in the form of steam is released into the He steam and is carried through the reactor where it will cause corrosion and mass transfer of C in the core. A paramount consideration in the design of a steam generator for a high temperature gas-cooled reactor is the prevention of tube failures. Preference, therefore, should be given to a forced circulation design.more » The Loeffler Boiler would be the best from this standpoint alone since only steam enters the tubes, and its circulation rate can be maintained at an adequate value to insure cool tubes regardless of load fluctuations. The next type in the order of preference would be the forced recirculation boiler, since at least the boiier tubes always have an adequate cooling flow regardless of output. The third type in order of preference would be a Sulzer Type boiler since it has a separator to remove dissolved material from the water which is comparible in efficiency to a standard boiler drum and although the flow through evaporator and superheater fluctuates with load, the Sulzer Boiler can be operated as a forced recirculation boiler at low loads. The least desirable type would be a Benson or supercritical boiler which is completely dependent on input water purity for its survival. It is not claimed that Benson or supercritical boilers should not or will not be used in the future for gas-cooled reactors, but only that their use would be the least conservative choice from a tube failure standpoint at the present time. (auth)« less

Decommissioning of the Dragon High Temperature Reactor (HTR) Located at the Former United Kingdom Atomic Energy Authority (UKAEA) Research Site at Winfrith - 13180

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

Smith, Anthony A.

2013-07-01

The Dragon Reactor was constructed at the United Kingdom Atomic Energy Research Establishment at Winfrith in Dorset through the late 1950's and into the early 1960's. It was a High Temperature Gas Cooled Reactor (HTR) with helium gas coolant and graphite moderation. It operated as a fuel testing and demonstration reactor at up to 20 MW (Thermal) from 1964 until 1975, when international funding for this project was terminated. The fuel was removed from the core in 1976 and the reactor was put into Safestore. To meet the UK's Nuclear Decommissioning Authority (NDA) objective to 'drive hazard reduction' [1] itmore » is necessary to decommission and remediate all the Research Sites Restoration Ltd (RSRL) facilities. This includes the Dragon Reactor where the activated core, pressure vessel and control rods and the contaminated primary circuit (including a {sup 90}Sr source) still remain. It is essential to remove these hazards at the appropriate time and return the area occupied by the reactor to a safe condition. (author)« less

A molten Salt Am242M Production Reactor for Space Applications

NASA Technical Reports Server (NTRS)

Emrich, William

2005-01-01

The use of Am242m holds great promise for increasing the efficiency nuclear thermal rocket engines. Because Am242m has the highest fission cross section of any known isotope (1000's of barns), its extremely high reactivity may be used to directly heat a propellant gas with fission fragments. Since this isotope does not occur naturally, it must be bred in special production reactors designed for that purpose. The primary advantage to using molten salt reactors for breeding Am242m is that the reactors can be reprocessed continually yielding a constant rate of production of the isotope. Once built and initially fueled, the reactor will continually breed the additional fuel it needs to remain critical. The only feedstock required is a salt of U238. No enriched fuel is required during normal operation and all fissile material, except the Am242m, is maintained in a closed loop. For a reactor operating at 200 MW several kilograms of Am242m may be bred each year.

Operation of the WWR-S Reactor in Poland and in the Period 1961-1962; DOKLAD OB ISPOL'ZOVANII I EKSPLUATATSII REAKTORA TIPA VVR-C V POL'SHE ZA 1961-1962 GODA

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

Aleksandrowicz, J.

1963-03-01

The experimental equipment used in the work at the horizontal reactor channels is listed. Diagrams of the utilization of the nominal reactor power and core loading are given, the reactivity fractions of the separate fuel assemblies are detonated, together with the diagram of reactivity versus burnup. Reactor channels and space used for sample irradiation and isotope production are described, and the total number of irradiations is given. Results of the measurements connected with the routine reactor operation are quoted, namely: analysis of water purity in the primary circuit, analysis of the work of the ion exchanger and mechanical filter, andmore » analysis of air activity in the special ventilation system. Data are given concerning radiation protection of personnel, including individual monitoring. Leak testing of the fuel elements is discussed. Damage of the reactor equipment and appearance of alarm signals are described. (auth)« less

Support vector regression model of wastewater bioreactor performance using microbial community diversity indices: effect of stress and bioaugmentation.

PubMed

Seshan, Hari; Goyal, Manish K; Falk, Michael W; Wuertz, Stefan

2014-04-15

The relationship between microbial community structure and function has been examined in detail in natural and engineered environments, but little work has been done on using microbial community information to predict function. We processed microbial community and operational data from controlled experiments with bench-scale bioreactor systems to predict reactor process performance. Four membrane-operated sequencing batch reactors treating synthetic wastewater were operated in two experiments to test the effects of (i) the toxic compound 3-chloroaniline (3-CA) and (ii) bioaugmentation targeting 3-CA degradation, on the sludge microbial community in the reactors. In the first experiment, two reactors were treated with 3-CA and two reactors were operated as controls without 3-CA input. In the second experiment, all four reactors were additionally bioaugmented with a Pseudomonas putida strain carrying a plasmid with a portion of the pathway for 3-CA degradation. Molecular data were generated from terminal restriction fragment length polymorphism (T-RFLP) analysis targeting the 16S rRNA and amoA genes from the sludge community. The electropherograms resulting from these T-RFs were used to calculate diversity indices - community richness, dynamics and evenness - for the domain Bacteria as well as for ammonia-oxidizing bacteria in each reactor over time. These diversity indices were then used to train and test a support vector regression (SVR) model to predict reactor performance based on input microbial community indices and operational data. Considering the diversity indices over time and across replicate reactors as discrete values, it was found that, although bioaugmentation with a bacterial strain harboring a subset of genes involved in the degradation of 3-CA did not bring about 3-CA degradation, it significantly affected the community as measured through all three diversity indices in both the general bacterial community and the ammonia-oxidizer community (α = 0.5). The impact of bioaugmentation was also seen qualitatively in the variation of community richness and evenness over time in each reactor, with overall community richness falling in the case of bioaugmented reactors subjected to 3-CA and community evenness remaining lower and more stable in the bioaugmented reactors as opposed to the unbioaugmented reactors. Using diversity indices, 3-CA input, bioaugmentation and time as input variables, the SVR model successfully predicted reactor performance in terms of the removal of broad-range contaminants like COD, ammonia and nitrate as well as specific contaminants like 3-CA. This work was the first to demonstrate that (i) bioaugmentation, even when unsuccessful, can produce a change in community structure and (ii) microbial community information can be used to reliably predict process performance. However, T-RFLP may not result in the most accurate representation of the microbial community itself, and a much more powerful prediction tool can potentially be developed using more sophisticated molecular methods. Copyright © 2014 Elsevier Ltd. All rights reserved.

Impact of conversion to mixed-oxide fuels on reactor structural components

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

Yahr, G.T.

1997-04-01

The use of mixed-oxide (MOX) fuel to replace conventional uranium fuel in commercial light-water power reactors will result in an increase in the neutron flux. The impact of the higher flux on the structural integrity of reactor structural components must be evaluated. This report briefly reviews the effects of radiation on the mechanical properties of metals. Aging degradation studies and reactor operating experience provide a basis for determining the areas where conversion to MOX fuels has the potential to impact the structural integrity of reactor components.

Current status of the Double Chooz experiment

NASA Astrophysics Data System (ADS)

Haser, J.; Double Chooz Collaboration

2016-04-01

The Double Chooz reactor antineutrino experiment aims for a precision measurement of the neutrino mixing angle θ13. Located at the Chooz nuclear power plant in France, it observes an energy dependent deficit in the electron antineutrino spectrum, currently with one detector filled with gadolinium-loaded liquid scintillator at a baseline of 1.05 km. The Double Chooz analysis utilizes different approaches to extract θ13: A combined rate and spectral shape fit as well as a background-model-independent analysis based on reactor power variations are performed, giving consistent results. Among the recent reactor-based oscillation experiments with comparable baseline it was the only one to observe reactor shutdown phases, during which all reactors are turned off. These enabled to measure the backgrounds solely, allowing to crosscheck the background models used in the oscillation analysis. At present an improved analysis was put forward with twice as much data statistics collected compared to the last publication. Revised selection criteria and background studies enhance the signal to background ratio while a decrease in the corresponding uncertainties is achieved. Along with an improved energy calibration the overall systematic uncertainty on θ13 is reduced, preparing for a two detector analysis. The new analysis obtains from 467.90 live days with 66.5 GW-ton-years of exposure (reactor power × detector mass × live time) a value of sin2 ⁡ 2θ13 =0.090-0.029+0.032(stat + syst).

The Nucifer Experiment

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

Cucoanes, A.S., E-mail: cucoanes@subatech.in2p3.fr

In nuclear reactors, a large number of antineutrinos are generated in the decay chains of the fission products; thus a survey of the antineutrino flux could provide valuable information related to the uranium and plutonium content of the core. This application generated interest by the IAEA in using antineutrino detectors as a potential safeguard tool. Here we present the Nucifer experiment, developed in France, by CEA and CNRS/IN2P3. The design of this new antineutrino detector has focused on safety, size reduction, reliability and high detection efficiency with a good background rejection. The Nucifer detector is currently taking data at themore » OSIRIS research reactor, inside CEA-Saclay. Presently, the ongoing analyses are considering the main sources of background for the antineutrino detection; the first antineutrino result is expected in 2013. A possible contribution to the understanding of the so called “reactor antineutrino anomaly” is also discussed. Finally, we present a brief description of the proposed experiments at very short baselines (VSBL) from reactors in France.« less

Experiments on rehabilitation of radioactive metallic waste (RMW) of reactor stainless steels of Siberian chemical plant

NASA Astrophysics Data System (ADS)

Kolpakov, G. N.; Zakusilov, V. V.; Demyanenko, N. V.; Mishin, A. S.

2016-06-01

Stainless steel pipes, used to cool a reactor plant, have a high cost, and after taking a reactor out of service they must be buried together with other radioactive waste. Therefore, the relevant problem is the rinse of pipes from contamination, followed by returning to operation.

The Entrance and Exit Effects in Small Electrochemical Filter-Press Reactors Used in the Laboratory

ERIC Educational Resources Information Center

Frias-Ferrer, Angel; Gonzalez-Garcia, Jose; Saez, Veronica; Exposito, Eduardo; Sanchez-Sanchez, Carlos M.; Mantiel, Vicente; Walsh, Frank C.; Aldaz, Antonio; Walsh, Frank C.

2005-01-01

A laboratory experiment designed to examine the entrance and exit effects in small electrochemical filter-press reactors used in the laboratory is presented. The single compartment of the filter-press reactor is filled with different turbulence promoters to study their influence as compared to the empty configuration.

The rate of decay of fresh fission products from a nuclear reactor

NASA Astrophysics Data System (ADS)

Dolan, David J.

Determining the rate of decay of fresh fission products from a nuclear reactor is complex because of the number of isotopes involved, different types of decay, half-lives of the isotopes, and some isotopes decay into other radioactive isotopes. Traditionally, a simplified rule of 7s and 10s is used to determine the dose rate from nuclear weapons and can be to estimate the dose rate from fresh fission products of a nuclear reactor. An experiment was designed to determine the dose rate with respect to time from fresh fission products of a nuclear reactor. The experiment exposed 0.5 grams of unenriched Uranium to a fast and thermal neutron flux from a TRIGA Research Reactor (Lakewood, CO) for ten minutes. The dose rate from the fission products was measured by four Mirion DMC 2000XB electronic personal dosimeters over a period of six days. The resulting dose rate following a rule of 10s: the dose rate of fresh fission products from a nuclear reactor decreases by a factor of 10 for every 10 units of time.

Hydrodynamics of Packed Bed Reactor in Low Gravity

NASA Technical Reports Server (NTRS)

Motil, Brian J.; Nahra, Henry K.; Balakotaiah, Vemuri

2005-01-01

Packed bed reactors are well known for their vast and diverse applications in the chemical industry; from gas absorption, to stripping, to catalytic conversion. Use of this type of reactor in terrestrial applications has been rather extensive because of its simplicity and relative ease of operation. Developing similar reactors for use in microgravity is critical to many space-based advanced life support systems. However, the hydrodynamics of two-phase flow packed bed reactors in this new environment and the effects of one physiochemical process on another has not been adequately assessed. Surface tension or capillary forces play a much greater role which results in a shifting in flow regime transitions and pressure drop. Results from low gravity experiments related to flow regimes and two-phase pressure drop models are presented in this paper along with a description of plans for a flight experiment on the International Space Station (ISS). Understanding the packed bed hydrodynamics and its effects on mass transfer processes in microgravity is crucial for the design of packed bed chemical or biological reactors to be used for water reclamation and other life support processes involving water purification.

Large-scale testing of in-vessel debris cooling through external flooding of the reactor pressure vessel in the CYBL facility

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

Chu, T.Y.; Bentz, J.H.; Bergeron, K.D.

1994-04-01

The possibility of achieving in-vessel core retention by flooding the reactor cavity, or the ``flooded cavity``, is an accident management concept currently under consideration for advanced light water reactors (ALWR), as well as for existing light water reactors (LWR). The CYBL (CYlindrical BoiLing) facility is a facility specifically designed to perform large-scale confirmatory testing of the flooded cavity concept. CYBL has a tank-within-a-tank design; the inner 3.7 m diameter tank simulates the reactor vessel, and the outer tank simulates the reactor cavity. The energy deposition on the bottom head is simulated with an array of radiant heaters. The array canmore » deliver a tailored heat flux distribution corresponding to that resulting from core melt convection. The present paper provides a detailed description of the capabilities of the facility, as well as results of recent experiments with heat flux in the range of interest to those required for in-vessel retention in typical ALWRs. The paper concludes with a discussion of other experiments for the flooded cavity applications.« less

THERMAL NEUTRONIC REACTOR

DOEpatents

Spinrad, B.I.

1960-01-12

A novel thermal reactor was designed in which a first reflector formed from a high atomic weight, nonmoderating material is disposed immediately adjacent to the reactor core. A second reflector composed of a moderating material is disposed outwardly of the first reflector. The advantage of this novel reflector arrangement is that the first reflector provides a high slow neutron flux in the second reflector, where irradiation experiments may be conducted with a small effect on reactor reactivity.

NUCLEAR POWER INSTALLATIONS

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

Bell, F.R.

1962-12-01

An arrangement is described for nuclear power plants including a reactor and at least one heat exchanger having primary and secondary circuits through which are passed heat-conveying fluids. Pressure-resisting walls about the heat exchangers and the reactor are either integral with or rigidly connected to one another. The heat exchangers are arranged so that their casings tend to shield withdrawn control rods from damage by radiation. (R.J.S.)

Year One Summary of X-energy Pebble Fuel Development at ORNL

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

Helmreich, Grant W.; Hunn, John D.; McMurray, Jake W.

2017-06-01

The Advanced Reactor Concepts X-energy (ARC-Xe) Pebble Fuel Development project at Oak Ridge National Laboratory (ORNL) has successfully completed its first year, having made excellent progress in accomplishing programmatic objectives. The primary focus of research at ORNL in support of X-energy has been the training of X-energy fuel fabrication engineers and the establishment of US pebble fuel production capabilities able to supply the Xe-100 pebble-bed reactor. These efforts have been strongly supported by particle fuel fabrication and characterization expertise present at ORNL from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program.

Fast Pyrolysis of Poplar Using a Captive Sample Reactor: Effects of Inorganic Salts on Primary Pyrolysis Products

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

Mukarakate, C.; Robichaud, D.; Donohoe, B.

2012-01-01

We have constructed a captive sample reactor (CSR) to study fast pyrolysis of biomass. The reactor uses a stainless steel wire mesh to surround biomass materials with an isothermal environment by independent controlling of heating rates and pyrolysis temperatures. The vapors produced during pyrolysis are immediately entrained and transported in He carrier gas to a molecular beam mass spectrometer (MBMS). Formation of secondary products is minimized by rapidly quenching the sample support with liquid nitrogen. A range of alkali and alkaline earth metal (AAEM) and transition metal salts were tested to study their effect on composition of primary pyrolysis products.more » Multivariate curve resolution (MCR) analysis of the MBMS data shows that transition metal salts enhance pyrolysis of carbohydrates and AAEM salts enhances pyrolysis of lignin. This was supported by performing similar separate studies on cellulose, hemicellulose and extracted lignin. The effect of salts on char formation is also discussed.« less

Analysis of decommissioning costs for the AFRRI TRIGA reactor facility. Technical report

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

Forsbacka, M.; Moore, M.

1989-12-01

This report provides a cost analysis for decommissioning the Armed Forces Radiobiology Research Institute (AFRRI) TRIGA reactor facility. AFRRI is not suggesting that the AFRRI TRIGA reactor facility be decommissioned. This report was prepared in compliance with paragraph 50.33 of Title 10, Code of Federal Regulations, which requires that funding for the decommissioning of reactor facilities be available when licensed activities cease. The planned method of decommissioning is complete decontamination (DECON) of the AFRRI TRIGA reactor site to allow for restoration of the site to full public access. The cost of DECON in 1990 dollars is estimated to be $3,200,000.more » The anticipated ancillary costs of facility site demobilization and spent fuel shipment will be an additional $600,000. Thus, the total cost of terminating reactor operations at AFRRI will be about $3,800,000. The primary basis for developing this cost estimate was a study of the decommissioning costs of similar reactor facility performed by Battelle Pacific Northwest Laboratory, as provided in U.S. Nuclear Regulatory Commission publication NUREG/CR-1756. The data in this study were adapted to reflect the decommissioning requirements of the AFRRI TRIGA reactor facility.« less

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

Light Water Reactor Sustainability Program Integrated Program Plan

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

McCarthy, Kathryn A.; Busby, Jeremy; Hallbert, Bruce

2014-04-01

Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution tomore » the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline—even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy’s Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration’s energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program’s plans.« less

Lessons from Fukushima for Improving the Safety of Nuclear Reactors

NASA Astrophysics Data System (ADS)

Lyman, Edwin

2012-02-01

The March 2011 accident at the Fukushima Daiichi nuclear power plant has revealed serious vulnerabilities in the design, operation and regulation of nuclear power plants. While some aspects of the accident were plant- and site-specific, others have implications that are broadly applicable to the current generation of nuclear plants in operation around the world. Although many of the details of the accident progression and public health consequences are still unclear, there are a number of lessons that can already be drawn. The accident demonstrated the need at nuclear plants for robust, highly reliable backup power sources capable of functioning for many days in the event of a complete loss of primary off-site and on-site electrical power. It highlighted the importance of detailed planning for severe accident management that realistically evaluates the capabilities of personnel to carry out mitigation operations under extremely hazardous conditions. It showed how emergency plans rooted in the assumption that only one reactor at a multi-unit site would be likely to experience a crisis fail miserably in the event of an accident affecting multiple reactor units simultaneously. It revealed that alternate water injection following a severe accident could be needed for weeks or months, generating large volumes of contaminated water that must be contained. And it reinforced the grim lesson of Chernobyl: that a nuclear reactor accident could lead to widespread radioactive contamination with profound implications for public health, the economy and the environment. While many nations have re-examined their policies regarding nuclear power safety in the months following the accident, it remains to be seen to what extent the world will take the lessons of Fukushima seriously and make meaningful changes in time to avert another, and potentially even worse, nuclear catastrophe.

Thermal hydraulic-severe accident code interfaces for SCDAP/RELAP5/MOD3.2

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

Coryell, E.W.; Siefken, L.J.; Harvego, E.A.

1997-07-01

The SCDAP/RELAP5 computer code is designed to describe the overall reactor coolant system thermal-hydraulic response, core damage progression, and fission product release during severe accidents. The code is being developed at the Idaho National Engineering Laboratory under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission. The code is the result of merging the RELAP5, SCDAP, and COUPLE codes. The RELAP5 portion of the code calculates the overall reactor coolant system, thermal-hydraulics, and associated reactor system responses. The SCDAP portion of the code describes the response of the core and associated vessel structures.more » The COUPLE portion of the code describes response of lower plenum structures and debris and the failure of the lower head. The code uses a modular approach with the overall structure, input/output processing, and data structures following the pattern established for RELAP5. The code uses a building block approach to allow the code user to easily represent a wide variety of systems and conditions through a powerful input processor. The user can represent a wide variety of experiments or reactor designs by selecting fuel rods and other assembly structures from a range of representative core component models, and arrange them in a variety of patterns within the thermalhydraulic network. The COUPLE portion of the code uses two-dimensional representations of the lower plenum structures and debris beds. The flow of information between the different portions of the code occurs at each system level time step advancement. The RELAP5 portion of the code describes the fluid transport around the system. These fluid conditions are used as thermal and mass transport boundary conditions for the SCDAP and COUPLE structures and debris beds.« less

40 CFR 63.9882 - What parts of my plant does this subpart cover?

Code of Federal Regulations, 2010 CFR

2010-07-01

... CATEGORIES (CONTINUED) National Emissions Standards for Hazardous Air Pollutants for Primary Magnesium... affected sources are each new and existing primary magnesium refining facility. (b) This subpart covers emissions from each spray dryer stack, magnesium chloride storage bins scrubber stack, melt/reactor system...

40 CFR 63.9882 - What parts of my plant does this subpart cover?

Code of Federal Regulations, 2012 CFR

2012-07-01

... CATEGORIES (CONTINUED) National Emissions Standards for Hazardous Air Pollutants for Primary Magnesium... affected sources are each new and existing primary magnesium refining facility. (b) This subpart covers emissions from each spray dryer stack, magnesium chloride storage bins scrubber stack, melt/reactor system...

40 CFR 63.9882 - What parts of my plant does this subpart cover?

Code of Federal Regulations, 2014 CFR

2014-07-01

... CATEGORIES (CONTINUED) National Emissions Standards for Hazardous Air Pollutants for Primary Magnesium... affected sources are each new and existing primary magnesium refining facility. (b) This subpart covers emissions from each spray dryer stack, magnesium chloride storage bins scrubber stack, melt/reactor system...

40 CFR 63.9882 - What parts of my plant does this subpart cover?

Code of Federal Regulations, 2011 CFR

2011-07-01

... CATEGORIES (CONTINUED) National Emissions Standards for Hazardous Air Pollutants for Primary Magnesium... affected sources are each new and existing primary magnesium refining facility. (b) This subpart covers emissions from each spray dryer stack, magnesium chloride storage bins scrubber stack, melt/reactor system...

40 CFR 63.9882 - What parts of my plant does this subpart cover?

Code of Federal Regulations, 2013 CFR

2013-07-01

... CATEGORIES (CONTINUED) National Emissions Standards for Hazardous Air Pollutants for Primary Magnesium... affected sources are each new and existing primary magnesium refining facility. (b) This subpart covers emissions from each spray dryer stack, magnesium chloride storage bins scrubber stack, melt/reactor system...

Cold Trap Dismantling and Sodium Removal at a Fast Breeder Reactor - 12327

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

Graf, A.; Petrick, H.; Stutz, U.

2012-07-01

The first German prototype Fast Breeder Nuclear Reactor (KNK) is currently being dismantled after being the only operating Fast Breeder-type reactor in Germany. As this reactor type used sodium as a coolant in its primary and secondary circuit, seven cold traps containing various amounts of partially activated sodium needed to be disposed of as part of the dismantling. The resulting combined difficulties of radioactive contamination and high chemical reactivity were handled by treating the cold traps differently depending on their size and the amount of sodium contained inside. Six small cold traps were processed onsite by cutting them up intomore » small parts using a band saw under a protective atmosphere. The sodium was then converted to sodium hydroxide by using water. The remaining large cold trap could not be handled in the same way due to its dimensions (2.9 m x 1.1 m) and the declared amount of sodium inside (1,700 kg). It was therefore manually dismantled inside a large box filled with a protective atmosphere, while the resulting pieces were packaged for later burning in a special facility. The experiences gained by KNK during this process may be advantageous for future dismantling projects in similar sodium-cooled reactors worldwide. The dismantling of a prototype fast breeder reactor provides the challenge not only to dismantle radioactive materials but also to handle sodium-contaminated or sodium-containing components. The treatment of sodium requires additional equipment and installations to ensure a safe handling. Since it is not permitted to bring sodium into a repository, all sodium has to be neutralized either through a controlled reaction with water or by incinerating. The resulting components can be disposed of as normal radioactive waste with no further conditions. The handling of sodium needs skilled and experienced workers to minimize the inherent risks. And the example of the disposal of the large KNK cold trap shows the interaction with others and also foreign decommissioning projects can provide solutions with were unknown before. (authors)« less

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

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

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

2001-11-01

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

Improving the oxidation resistance of 316L stainless steel in simulated pressurized water reactor primary water by electropolishing treatment

NASA Astrophysics Data System (ADS)

Han, Guangdong; Lu, Zhanpeng; Ru, Xiangkun; Chen, Junjie; Xiao, Qian; Tian, Yongwu

2015-12-01

The oxidation behavior of 316L stainless steel specimens after emery paper grounding, mechanical polishing, and electropolishing were investigated in simulated pressurized water reactor primary water at 310 °C for 120 and 500 h. Electropolishing afforded improved oxidation resistance especially during the early immersion stages. Duplex oxide films comprising a coarse Fe-rich outer layer and a fine Cr-rich inner layer formed on all specimens after 500 h of immersion. Only a compact layer was observed on the electropolished specimen after 120 h of immersion. The enrichment of chromium in the electropolished layer contributed to the passivity and protectiveness of the specimen.

Total absorption spectroscopy of fission fragments relevant for reactor antineutrino spectra

DOE PAGES

Fallot, M.; Porta, A.; Meur, L. Le; ...

2017-09-13

Here, the accurate determination of reactor antineutrino spectra remains a very active research topic for which new methods of study have emerged in recent years. Indeed, following the long-recognized reactor anomaly (measured antineutrino deficit in short baseline reactor experiments when compared with spectral predictions), the three international reactor neutrino experiments Double Chooz, Daya Bay and Reno have recently demonstrated the existence of spectral distortions in their measurements with respect to the same predictions. These spectral predictions were obtained through the conversion of integral beta-energy spectra obtained at the ILL research reactor. Several studies have shown that the underlying nuclear physicsmore » required for the conversion of these spectra into antineutrino spectra is not totally understood. An alternative to such converted spectra is a complementary approach that consists of determining the antineutrino spectrum by means of the measurement and processing of nuclear data. The beta properties of some key fission products suffer from the pandemonium effect which can be circumvented by the use of the Total Absorption Gamma-ray Spectroscopy technique (TAGS). The two main contributors to the Pressurized Water Reactor antineutrino spectrum in the region where the spectral distortion has been observed are 92Rb and 142Cs, which have been measured at the radioactive beam facility of the University of Jyvaskyla in two TAGS experiments. We present the results of the analysis of the TAGS measurements of the β-decay properties of 92Rb along with preliminary results on 142Cs and report on the measurements already performed.« less

Total absorption spectroscopy of fission fragments relevant for reactor antineutrino spectra

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

Fallot, M.; Porta, A.; Meur, L. Le

Here, the accurate determination of reactor antineutrino spectra remains a very active research topic for which new methods of study have emerged in recent years. Indeed, following the long-recognized reactor anomaly (measured antineutrino deficit in short baseline reactor experiments when compared with spectral predictions), the three international reactor neutrino experiments Double Chooz, Daya Bay and Reno have recently demonstrated the existence of spectral distortions in their measurements with respect to the same predictions. These spectral predictions were obtained through the conversion of integral beta-energy spectra obtained at the ILL research reactor. Several studies have shown that the underlying nuclear physicsmore » required for the conversion of these spectra into antineutrino spectra is not totally understood. An alternative to such converted spectra is a complementary approach that consists of determining the antineutrino spectrum by means of the measurement and processing of nuclear data. The beta properties of some key fission products suffer from the pandemonium effect which can be circumvented by the use of the Total Absorption Gamma-ray Spectroscopy technique (TAGS). The two main contributors to the Pressurized Water Reactor antineutrino spectrum in the region where the spectral distortion has been observed are 92Rb and 142Cs, which have been measured at the radioactive beam facility of the University of Jyvaskyla in two TAGS experiments. We present the results of the analysis of the TAGS measurements of the β-decay properties of 92Rb along with preliminary results on 142Cs and report on the measurements already performed.« less

Experimental Anomalies in Neutrino Physics

NASA Astrophysics Data System (ADS)

Palamara, Ornella

2014-03-01

In recent years, experimental anomalies ranging in significance (2.8-3.8 σ) have been reported from a variety of experiments studying neutrinos over baselines less than 1 km. Results from the LSND and MiniBooNE short-baseline νe /νe appearance experiments show anomalies which cannot be described by oscillations between the three standard model neutrinos (the ``LSND anomaly''). In addition, a re-analysis of the anti-neutrino flux produced by nuclear power reactors has led to an apparent deficit in νe event rates in a number of reactor experiments (the ``reactor anomaly''). Similarly, calibration runs using 51Cr and 37Ar radioactive sources in the Gallium solar neutrino experiments GALLEX and SAGE have shown an unexplained deficit in the electron neutrino event rate over very short distances (the ``Gallium anomaly''). The puzzling results from these experiments, which together may suggest the existence of physics beyond the Standard Model and hint at exciting new physics, including the possibility of additional low-mass sterile neutrino states, have raised the interest in the community for new experimental efforts that could eventually solve this puzzle. Definitive evidence for sterile neutrinos would be a revolutionary discovery, with implications for particle physics as well as cosmology. Proposals to address these signals by employing accelerator, reactor and radioactive source experiments are in the planning stages or underway worldwide. In this talk some of these will be reviewed, with emphasis on the accelerator programs.

Trench fast reactor design using the microcomputer

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

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

1987-01-01

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

Evaluation of Neutron Radiography Reactor LEU-Core Start-Up Measurements

DOE PAGES

Bess, John D.; Maddock, Thomas L.; Smolinski, Andrew T.; ...

2014-11-04

Benchmark models were developed to evaluate the cold-critical start-up measurements performed during the fresh core reload of the Neutron Radiography (NRAD) reactor with Low Enriched Uranium (LEU) fuel. Experiments include criticality, control-rod worth measurements, shutdown margin, and excess reactivity for four core loadings with 56, 60, 62, and 64 fuel elements. The worth of four graphite reflector block assemblies and an empty dry tube used for experiment irradiations were also measured and evaluated for the 60-fuel-element core configuration. Dominant uncertainties in the experimental k eff come from uncertainties in the manganese content and impurities in the stainless steel fuel claddingmore » as well as the 236U and erbium poison content in the fuel matrix. Calculations with MCNP5 and ENDF/B-VII.0 neutron nuclear data are approximately 1.4% (9σ) greater than the benchmark model eigenvalues, which is commonly seen in Monte Carlo simulations of other TRIGA reactors. Simulations of the worth measurements are within the 2σ uncertainty for most of the benchmark experiment worth values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.« less

Evaluation of Neutron Radiography Reactor LEU-Core Start-Up Measurements

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

Bess, John D.; Maddock, Thomas L.; Smolinski, Andrew T.

Benchmark models were developed to evaluate the cold-critical start-up measurements performed during the fresh core reload of the Neutron Radiography (NRAD) reactor with Low Enriched Uranium (LEU) fuel. Experiments include criticality, control-rod worth measurements, shutdown margin, and excess reactivity for four core loadings with 56, 60, 62, and 64 fuel elements. The worth of four graphite reflector block assemblies and an empty dry tube used for experiment irradiations were also measured and evaluated for the 60-fuel-element core configuration. Dominant uncertainties in the experimental k eff come from uncertainties in the manganese content and impurities in the stainless steel fuel claddingmore » as well as the 236U and erbium poison content in the fuel matrix. Calculations with MCNP5 and ENDF/B-VII.0 neutron nuclear data are approximately 1.4% (9σ) greater than the benchmark model eigenvalues, which is commonly seen in Monte Carlo simulations of other TRIGA reactors. Simulations of the worth measurements are within the 2σ uncertainty for most of the benchmark experiment worth values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.« less

Accelerated anaerobic hydrolysis rates under a combination of intermittent aeration and anaerobic conditions.

PubMed

Jensen, T R; Lastra Milone, T; Petersen, G; Andersen, H R

2017-04-01

Anaerobic hydrolysis in activated return sludge was investigated in laboratory scale experiments to find if intermittent aeration would accelerate anaerobic hydrolysis rates compared to anaerobic hydrolysis rates under strict anaerobic conditions. The intermittent reactors were set up in a 240 h experiment with intermittent aeration (3 h:3 h) in a period of 24 h followed by a subsequent anaerobic period of 24 h in a cycle of 48 h which was repeated five times during the experiment. The anaerobic reactors were kept under strict anaerobic conditions in the same period (240 h). Two methods for calculating hydrolysis rates based on soluble chemical oxygen demand were compared. Two-way analysis of variance with the Bonferroni post-test was performed in order to register any significant difference between reactors with intermittent aeration and strictly anaerobic conditions respectively. The experiment demonstrated a statistically significant difference in favor of the reactors with intermittent aeration showing a tendency towards accelerated anaerobic hydrolysis rates due to application of intermittent aeration. The conclusion of the work is thus that intermittent aeration applied in the activated return sludge process can improve the treatment capacity further in full scale applications.

Aluminum Carbothermic Technology

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

Bruno, Marshall J.

2005-03-31

This report documents the non-proprietary research and development conducted on the Aluminum Carbothermic Technology (ACT) project from contract inception on July 01, 2000 to termination on December 31, 2004. The objectives of the program were to demonstrate the technical and economic feasibility of a new carbothermic process for producing commercial grade aluminum, designated as the ''Advanced Reactor Process'' (ARP). The scope of the program ranged from fundamental research through small scale laboratory experiments (65 kW power input) to larger scale test modules at up to 1600 kW power input. The tasks included work on four components of the process, Stagesmore » 1 and 2 of the reactor, vapor recovery and metal alloy decarbonization; development of computer models; and economic analyses of capital and operating costs. Justification for developing a new, carbothermic route to aluminum production is defined by the potential benefits in reduced energy, lower costs and more favorable environmental characteristics than the conventional Hall-Heroult process presently used by the industry. The estimated metrics for these advantages include energy rates at approximately 10 kWh/kg Al (versus over 13 kWh/kg Al for Hall-Heroult), capital costs as low as $1250 per MTY (versus 4,000 per MTY for Hall-Heroult), operating cost reductions of over 10%, and up to 37% reduction in CO2 emissions for fossil-fuel power plants. Realization of these benefits would be critical to sustaining the US aluminum industries position as a global leader in primary aluminum production. One very attractive incentive for ARP is its perceived ability to cost effectively produce metal over a range of smelter sizes, not feasible for Hall-Heroult plants which must be large, 240,000 TPY or more, to be economical. Lower capacity stand alone carbothermic smelters could be utilized to supply molten metal at fabrication facilities similar to the mini-mill concept employed by the steel industry. Major accomplishments for the program include definition of the system thermo-chemistry, demonstration of reactor stage 1, development of reactor stage 2 critical components in a 500 kW module, experimental determination of the vapor recovery reactor fundamentals, detailed design and installation of an advanced stage 1/vapor recovery reactor, feasibility of efficient separation of Al-C metal alloy product, updated capital and operating cost estimates, and development of computer models for all steps of the Advanced Reactor Process.« less

(Boiling water reactor (BWR) CORA experiments)

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

Ott, L.J.

To participate in the 1990 CORA Workshop at Kernforschungszentrum Karlsruhe (KfK) GmbH, Karlsruhe, FRG, on October 1--4, and to participate in detailed discussions on October 5 with the KfK CORA Boiling Water Reactor (BWR) experiments. The traveler attended the 1990 CORA Workshop at KfK, FRG. Participation included the presentation of a paper on work performed by the Boiling Water Reactor Core Melt Progression Phenomena Program at Oak Ridge National Laboratory (ORNL) on posttest analyses of CORA BWR experiments. The Statement of Work (November 1989) for the BWR Core Melt Progression Phenomena Program provides for pretest and posttest analyses of themore » BWR CORA experiments performed at KfK. Additionally, it is intended that ORNL personnel participate in the planning process for future CORA BWR experiments. For these purposes, meetings were held with KfK staff to discuss such topics as (1) experimental test schedule, (2) BWR test conduct, (3) perceived BWR experimental needs, and (4) KfK operational staff needs with respect to ORNL support. 19 refs.« less

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

Adamov, E.O.; Kuklin, A.N.; Mityaev, Yu.I.

The nuclear power plants with boiling water reactors of improved safety are being developed. There is 26 years of operating experience with the plant VK-50 in Dimitrovgrad. The design and operation of the BWR reactors are described.

Design and manufacture of a D-shape coil-based toroid-type HTS DC reactor using 2nd generation HTS wire

NASA Astrophysics Data System (ADS)

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

2014-09-01

This paper describes the design specifications and performance of a real toroid-type high temperature superconducting (HTS) DC reactor. The HTS DC reactor was designed using 2G HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The target inductance of the HTS DC reactor was 400 mH. The expected operating temperature was under 20 K. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. Performances of the toroid-type HTS DC reactor were analyzed through experiments conducted under the steady-state and charge conditions. The fundamental design specifications and the data obtained from this research will be applied to the design of a commercial-type HTS DC reactor.

New results from RENO and prospects with RENO-50

NASA Astrophysics Data System (ADS)

Kim, Soo-Bong

2015-08-01

RENO (Reactor Experiment for Neutrino Oscillation) has made a definitive measurement of the smallest mixing angle θ13 in 2012, based on the disappearance of electron antineutrinos. More precise measurements have been obtained and presented on the mixing angle and the reactor neutrino spectrum, using ˜800 days of data. We have observed an excess of inverse-beta-decay (IBD) prompt spectrum near 5 MeV with respect to the most commonly used models. The excess is found to be consistent with coming from reactors. We have also successfully measured the reactor neutrino flux with a delayed signal of neutron capture on hydrogen. A future reactor neutrino experiment of RENO-50 is proposed to determine the neutrino mass hierarchy and to make highly precise measurements of θ12, Δm212, and Δ m312. Physics goals and sensitivities of RENO-50 are presented with a strategy of obtaining an extremely good energy resolution toward the neutrino mass hierarchy.

PRESSURE SYSTEM CONTROL

DOEpatents

Esselman, W.H.; Kaplan, G.M.

1961-06-20

The control of pressure in pressurized liquid systems, especially a pressurized liquid reactor system, may be achieved by providing a bias circuit or loop across a closed loop having a flow restriction means in the form of an orifice, a storage tank, and a pump connected in series. The subject invention is advantageously utilized where control of a reactor can be achieved by response to the temperature and pressure of the primary cooling system.

PWR PRELIMINARY DESIGN FOR PL-3

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

Humphries, G. E.

1962-02-28

The pressurized water reactor preliminary design, the preferred design developed under Phase I of the PL-3 contract, is presented. Plant design criteria, summary of plant selection, plant description, reactor and primary system description, thermal and hydraulic analysis, nuclear analysis, control and instrumentatlon description, shielding description, auxiliary systems, power plant equipment, waste dispusal, buildings and tunnels, services, operation and maintenance, logistics, erection, cost information, and a training program outline are given. (auth)

Moving bed reactor setup to study complex gas-solid reactions.

PubMed

Gupta, Puneet; Velazquez-Vargas, Luis G; Valentine, Charles; Fan, Liang-Shih

2007-08-01

A moving bed scale reactor setup for studying complex gas-solid reactions has been designed in order to obtain kinetic data for scale-up purpose. In this bench scale reactor setup, gas and solid reactants can be contacted in a cocurrent and countercurrent manner at high temperatures. Gas and solid sampling can be performed through the reactor bed with their composition profiles determined at steady state. The reactor setup can be used to evaluate and corroborate model parameters accounting for intrinsic reaction rates in both simple and complex gas-solid reaction systems. The moving bed design allows experimentation over a variety of gas and solid compositions in a single experiment unlike differential bed reactors where the gas composition is usually fixed. The data obtained from the reactor can also be used for direct scale-up of designs for moving bed reactors.

FERRET-SAND II physics-dosimetry analysis for N Reactor Pressure Tubes 2954, 3053 and 1165 using a WIMS calculated input spectrum

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

McElroy, W.N.; Kellogg, L.S.; Matsumoto, W.Y.

1988-05-01

This report is in response to a request from Westinghouse Hanford Company (WHC) that the PNL National Dosimetry Center (NDC) perform physics-dosimetry analyses (E > MeV) for N Reactor Pressure Tubes 2954 and 3053. As a result of these analyses, and recommendations for additional studies, two physics-dosimetry re-evaluations for Pressure Tube 1165 were also accomplished. The primary objective of Pacific Northwest Laboratories' (PNL) National Dosimetry Center (NDC) physics-dosimetry work for N Reactor was to provide FERRET-SAND II physics-dosimetry results to assist in the assessment of neutron radiation-induced changes in the physical and mechanical properties of N Reactor pressure tubes. 15more » refs., 6 figs., 5 tabs.« less

Alloying of steel and graphite by hydrogen in nuclear reactor

NASA Astrophysics Data System (ADS)

Krasikov, E.

2017-02-01

In traditional power engineering hydrogen may be one of the first primary source of equipment damage. This problem has high actuality for both nuclear and thermonuclear power engineering. Study of radiation-hydrogen embrittlement of the steel raises the question concerning the unknown source of hydrogen in reactors. Later unexpectedly high hydrogen concentrations were detected in irradiated graphite. It is necessary to look for this source of hydrogen especially because hydrogen flakes were detected in reactor vessels of Belgian NPPs. As a possible initial hypothesis about the enigmatical source of hydrogen one can propose protons generation during beta-decay of free neutrons поскольку inasmuch as protons detected by researches at nuclear reactors as witness of beta-decay of free neutrons.

On-line coating of glass with tin oxide by atmospheric pressure chemical vapor deposition.

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

Allendorf, Mark D.; Sopko, J.F.; Houf, William G.

2006-11-01

Atmospheric pressure chemical vapor deposition (APCVD) of tin oxide is a very important manufacturing technique used in the production of low-emissivity glass. It is also the primary method used to provide wear-resistant coatings on glass containers. The complexity of these systems, which involve chemical reactions in both the gas phase and on the deposition surface, as well as complex fluid dynamics, makes process optimization and design of new coating reactors a very difficult task. In 2001 the U.S. Dept. of Energy Industrial Technologies Program Glass Industry of the Future Team funded a project to address the need for more accuratemore » data concerning the tin oxide APCVD process. This report presents a case study of on-line APCVD using organometallic precursors, which are the primary reactants used in industrial coating processes. Research staff at Sandia National Laboratories in Livermore, CA, and the PPG Industries Glass Technology Center in Pittsburgh, PA collaborated to produce this work. In this report, we describe a detailed investigation of the factors controlling the growth of tin oxide films. The report begins with a discussion of the basic elements of the deposition chemistry, including gas-phase thermochemistry of tin species and mechanisms of chemical reactions involved in the decomposition of tin precursors. These results provide the basis for experimental investigations in which tin oxide growth rates were measured as a function of all major process variables. The experiments focused on growth from monobutyltintrichloride (MBTC) since this is one of the two primary precursors used industrially. There are almost no reliable growth-rate data available for this precursor. Robust models describing the growth rate as a function of these variables are derived from modeling of these data. Finally, the results are used to conduct computational fluid dynamic simulations of both pilot- and full-scale coating reactors. As a result, general conclusions are reached concerning the factors affecting the growth rate in on-line APCVD reactors. In addition, a substantial body of data was generated that can be used to model many different industrial tin oxide coating processes. These data include the most extensive compilation of thermochemistry for gas-phase tin-containing species as well as kinetic expressions describing tin oxide growth rates over a wide range of temperatures, pressures, and reactant concentrations.« less

Primary and Photochemically Aged Aerosol Emissions from Biomass Cookstoves: Chemical and Physical Characterization.

PubMed

Reece, Stephen M; Sinha, Aditya; Grieshop, Andrew P

2017-08-15

Secondary organic aerosol (SOA) formation during photo-oxidation of primary emissions from cookstoves used in developing countries may make important contributions to their climate and air quality impacts. We present results from laboratory experiments with a field portable oxidation flow reactor (F-OFR) to study the evolution of emissions over hours to weeks of equivalent atmospheric aging. Lab tests, using dry red oak, measured fresh and aged emissions from a 3 stone fire (TSF), a "rocket" natural draft stove (NDS), and a forced draft gasifier stove (FDGS), in order of increasing modified combustion efficiency (MCE) and decreasing particulate matter emission factors (EF). SOA production was observed for all stoves/tests; organic aerosol (OA) enhancement factor ranged from 1.2 to 3.1, decreasing with increased MCE. In primary emissions, OA mass spectral fragments associated with oxygenated species (primary biomass burning markers) increased (decreased) with MCE; fresh OA from FDGS combustion was especially oxygenated. OA oxygenation increased with further oxidation for all stove emissions, even where minimal enhancement was observed. More efficient stoves emit particles with greater net direct specific warming than TSFs, with the difference increasing with aging. Our results show that the properties and evolution of cookstove emissions are a strong function of combustion efficiency and atmospheric aging.

Pebble Fuel Handling and Reactivity Control for Salt-Cooled High Temperature Reactors

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

Peterson, Per; Greenspan, Ehud

2015-02-09

This report documents the work completed on the X-PREX facility under NEUP Project 11- 3172. This project seeks to demonstrate the viability of pebble fuel handling and reactivity control for fluoride salt-cooled high-temperature reactors (FHRs). The research results also improve the understanding of pebble motion in helium-cooled reactors, as well as the general, fundamental understanding of low-velocity granular flows. Successful use of pebble fuels in with salt coolants would bring major benefits for high-temperature reactor technology. Pebble fuels enable on-line refueling and operation with low excess reactivity, and thus simpler reactivity control and improved fuel utilization. If fixed fuel designsmore » are used, the power density of salt- cooled reactors is limited to 10 MW/m 3 to obtain adequate duration between refueling, but pebble fuels allow power densities in the range of 20 to 30 MW/m 3. This can be compared to the typical modular helium reactor power density of 5 MW/m3. Pebble fuels also permit radial zoning in annular cores and use of thorium or graphite pebble blankets to reduce neutron fluences to outer radial reflectors and increase total power production. Combined with high power conversion efficiency, compact low-pressure primary and containment systems, and unique safety characteristics including very large thermal margins (>500°C) to fuel damage during transients and accidents, salt-cooled pebble fuel cores offer the potential to meet the major goals of the Advanced Reactor Concepts Development program to provide electricity at lower cost than light water reactors with improved safety and system performance.This report presents the facility description, experimental results, and supporting simulation methods of the new X-Ray Pebble Recirculation Experiment (X-PREX), which is now operational and being used to collect data on the behavior of slow dense granular flows relevant to pebble bed reactor core designs. The X-PREX facility uses novel digital x-ray tomography methods to track both the translational and rotational motion of spherical pebbles, which provides unique experimental results that can be used to validate discrete element method (DEM) simulations of pebble motion. The validation effort supported by the X-PREX facility provides a means to build confidence in analysis of pebble bed configuration and residence time distributions that impact the neutronics, thermal hydraulics, and safety analysis of pebble bed reactor cores. Experimental and DEM simulation results are reported for silo drainage, a classical problem in the granular flow literature, at several hopper angles. These studies include conventional converging and novel diverging geometries that provide additional flexibility in the design of pebble bed reactor cores. Excellent agreement is found between the X-PREX experimental and DEM simulation results. This report also includes results for additional studies relevant to the design and analysis of pebble bed reactor cores including the study of forces on shut down blades inserted directly into a packed bed and pebble flow in a cylindrical hopper that is representative of a small test reactor.« less

Effect of Reactor Design on the Plasma Treatment of NOx

DTIC Science & Technology

1998-10-01

control parameter is the input energy density. Consequently, different reactor designs should yield basically the same plasma chemistry if the experiments are performed under identical gas composition and temperature conditions.

Supplement to Theory of Neutron Chain Reactions

DOE R&D Accomplishments Database

Weinberg, Alvin M.; Noderer, L. C.

1952-05-26

General discussions are given of the theory of neutron chain reactions. These include observations on exponential experiments, the general reactor with resonance fission, microscopic pile theory, and homogeneous slow neutron reactors. (B.J.H.)

Assessing the influence of reactor system design criteria on the performance of model colon fermentation units.

PubMed

Moorthy, Arun S; Eberl, Hermann J

2014-04-01

Fermentation reactor systems are a key platform in studying intestinal microflora, specifically with respect to questions surrounding the effects of diet. In this study, we develop computational representations of colon fermentation reactor systems as a way to assess the influence of three design elements (number of reactors, emptying mechanism, and inclusion of microbial immobilization) on three performance measures (total biomass density, biomass composition, and fibre digestion efficiency) using a fractional-factorial experimental design. It was determined that the choice of emptying mechanism showed no effect on any of the performance measures. Additionally, it was determined that none of the design criteria had any measurable effect on reactor performance with respect to biomass composition. It is recommended that model fermentation systems used in the experimenting of dietary effects on intestinal biomass composition be streamlined to only include necessary system design complexities, as the measured performance is not benefited by the addition of microbial immobilization mechanisms or semi-continuous emptying scheme. Additionally, the added complexities significantly increase computational time during simulation experiments. It was also noted that the same factorial experiment could be directly adapted using in vitro colon fermentation systems. Copyright © 2013 The Society for Biotechnology, Japan. Published by Elsevier B.V. All rights reserved.

High-Z plasma facing components in fusion devices: boundary conditions and operational experiences

NASA Astrophysics Data System (ADS)

Neu, R.

2006-04-01

In present day fusion devices optimization of the performance and experimental freedom motivates the use of low-Z plasma facing materials (PFMs). However, in a future fusion reactor, for economic reasons, a sufficient lifetime of the first wall components is essential. Additionally, tritium retention has to be small to meet safety requirements. Tungsten appears to be the most realistic material choice for reactor plasma facing components (PFCs) because it exhibits the lowest erosion. But besides this there are a lot of criteria which have to be fulfilled simultaneously in a reactor. Results from present day devices and from laboratory experiments confirm the advantages of high-Z PFMs but also point to operational restrictions, when using them as PFCs. These are associated with the central impurity concentration, which is determined by the sputtering yield, the penetration of the impurities and their transport within the confined plasma. The restrictions could exclude successful operation of a reactor, but concomitantly there exist remedies to ameliorate their impact. Obviously some price has to be paid in terms of reduced performance but lacking of materials or concepts which could substitute high-Z PFCs, emphasis has to be put on the development and optimization of reactor-relevant scenarios which incorporate the experiences and measures.

Criticality experiments and analysis of molybdenum reflected cylindrical uranyl fluoride water solution reactors

NASA Technical Reports Server (NTRS)

Fieno, D.; Fox, T.; Mueller, R.

1972-01-01

Clean criticality data were obtained from molybdenum-reflected cylindrical uranyl-fluoride-water solution reactors. Using ENDF/B molybdenum cross sections, a nine energy group two-dimensional transport calculation of a reflected reactor configuration predicted criticality to within 7 cents of the experimental value. For these reactors, it was necessary to compute the reflector resonance integral by a detailed transport calculation at the core-reflector interface volume in the energy region of the two dominant resonances of natural molybdenum.

Advanced reactors and novel reactions for the conversion of triglyceride based oils into high quality renewable transportation fuels

NASA Astrophysics Data System (ADS)

Linnen, Michael James

Sustainable energy continues to grow more important to all societies, leading to the research and development of a variety of alternative and renewable energy technologies. Of these, renewable liquid transportation fuels may be the most visible to consumers, and this visibility is further magnified by the long-term trend of increasingly expensive petroleum fuels that the public consumes. While first-generation biofuels such as biodiesel and fuel ethanol have been integrated into the existing fuel infrastructures of several countries, the chemical differences between them and their petroleum counterparts reduce their effectiveness. This gives rise to the development and commercialization of second generation biofuels, many of which are intended to have equivalent properties to those of their petroleum counterparts. In this dissertation, the primary reactions for a second-generation biofuel process, known herein as the University of North Dakota noncatalytic cracking process (NCP), have been studied at the fundamental level and improved. The NCP is capable of producing renewable fuels and chemicals that are virtually the same as their petroleum counterparts in performance and quality (i.e., petroleum-equivalent). In addition, a novel analytical method, FIMSDIST was developed which, within certain limitations, can increase the elution capabilities of GC analysis and decrease sample processing times compared to other high resolution methods. These advances are particularly useful for studies of highly heterogeneous fuel and/or organic chemical intermediates, such as those studied for the NCP. However the data from FIMSDIST must be supplemented with data from other methods such as for certain carboxylic acid, to provide accurate, comprehensive results, From a series of TAG cracking experiments that were performed, it was found that coke formation during cracking is most likely the result of excessive temperature and/or residence time in a cracking reactor. Based on this, a tubular cracking reactor was developed that could operate continuously without coke formation. The design also was proven to be scalable. Yields from the reactor were determined under a variety of conditions in order to predict the outputs from the NCP and to establish relationships/correlations between operating parameters and the product distribution. These studies led to the conclusion that the most severe operating conditions which do not induce coking are optimal over the experimental domain. In order to develop economical deoxygenation catalysts for use within the NCP, a series of experiments were performed using nickel catalysts, demonstrating that nickel catalysts could outperform their predecessor, a high cost palladium-based catalyst. A nickel catalyst was then tested in a packed bed reactor in order to determine suitable operating conditions for its commercial utilization in packed bed reactors.

Direct nn-Scattering Measurement With the Pulsed Reactor YAGUAR.

PubMed

Mitchell, G E; Furman, W I; Lychagin, E V; Muzichka, A Yu; Nekhaev, G V; Strelkov, A V; Sharapov, E I; Shvetsov, V N; Chernuhin, Yu I; Levakov, B G; Litvin, V I; Lyzhin, A E; Magda, E P; Crawford, B E; Stephenson, S L; Howell, C R; Tornow, W

2005-01-01

Although crucial for resolving the issue of charge symmetry in the nuclear force, direct measurement of nn-scattering by colliding free neutrons has never been performed. At present the Russian pulsed reactor YAGUAR is the best neutron source for performing such a measurement. It has a through channel where the neutron moderator is installed. The neutrons are counted by a neutron detector located 12 m from the reactor. In preliminary experiments an instantaneous value of 1.1 × 10(18)/cm(2)s was obtained for the thermal neutron flux density. The experiment will be performed by the DIANNA Collaboration as International Science & Technology Center (ISTC) project No. 2286.

Direct nn-Scattering Measurement With the Pulsed Reactor YAGUAR

PubMed Central

Mitchell, G. E.; Furman, W. I.; Lychagin, E. V.; Muzichka, A. Yu.; Nekhaev, G. V.; Strelkov, A. V.; Sharapov, E. I.; Shvetsov, V. N.; Chernuhin, Yu. I.; Levakov, B. G.; Litvin, V. I.; Lyzhin, A. E.; Magda, E. P.; Crawford, B. E.; Stephenson, S. L.; Howell, C. R.; Tornow, W

2005-01-01

Although crucial for resolving the issue of charge symmetry in the nuclear force, direct measurement of nn-scattering by colliding free neutrons has never been performed. At present the Russian pulsed reactor YAGUAR is the best neutron source for performing such a measurement. It has a through channel where the neutron moderator is installed. The neutrons are counted by a neutron detector located 12 m from the reactor. In preliminary experiments an instantaneous value of 1.1 × 1018/cm2s was obtained for the thermal neutron flux density. The experiment will be performed by the DIANNA Collaboration as International Science & Technology Center (ISTC) project No. 2286. PMID:27308126

Advances in modelling of condensation phenomena

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

Liu, W.S.; Zaltsgendler, E.; Hanna, B.

1997-07-01

The physical parameters in the modelling of condensation phenomena in the CANDU reactor system codes are discussed. The experimental programs used for thermal-hydraulic code validation in the Canadian nuclear industry are briefly described. The modelling of vapour generation and in particular condensation plays a key role in modelling of postulated reactor transients. The condensation models adopted in the current state-of-the-art two-fluid CANDU reactor thermal-hydraulic system codes (CATHENA and TUF) are described. As examples of the modelling challenges faced, the simulation of a cold water injection experiment by CATHENA and the simulation of a condensation induced water hammer experiment by TUFmore » are described.« less

Experience Gained during the Adaptation of Classical ChE Subjects to the Bologna Plan in Europe: The Case of Chemical Reactors

ERIC Educational Resources Information Center

Ponsa, Sergio; Sanchez, Antoni

2011-01-01

At present, due to the overall adaptation to the European Higher Education Area (EHEA), a new concept regarding the teaching methodology was thought to be essential for engineering subjects. In this paper we describe our experience teaching the altered content of the courses on two classical subjects; Chemical Reactors (Chemical Engineering) and…

How to Produce a Reactor Neutron Spectrum Using a Proton Accelerator

DOE PAGES

Burns, Kimberly A.; Wootan, David W.; Gates, Robert O.; ...

2015-06-18

A method for reproducing the neutron energy spectrum present in the core of an operating nuclear reactor using an engineered target in an accelerator proton beam is proposed. The protons interact with a target to create neutrons through various (p,n) type reactions. Spectral tailoring of the emitted neutrons can be used to modify the energy of the generated neutron spectrum to represent various reactor spectra. Through the use of moderators and reflectors, the neutron spectrum can be modified to reproduce many different spectra of interest including spectra in small thermal test reactors, large pressurized water reactors, and fast reactors. Themore » particular application of this methodology is the design of an experimental approach for using an accelerator to measure the betas produced during fission to be used to reduce uncertainties in the interpretation of reactor antineutrino measurements. This approach involves using a proton accelerator to produce a neutron field representative of a power reactor, and using this neutron field to irradiate fission foils of the primary isotopes contributing to fission in the reactor, creating unstable, neutron rich fission products that subsequently beta decay and emit electron antineutrinos. A major advantage of an accelerator neutron source over a neutron beam from a thermal reactor is that the fast neutrons can be slowed down or tailored to approximate various power reactor spectra. An accelerator based neutron source that can be tailored to match various reactor neutron spectra provides an advantage for control in studying how changes in the neutron spectra affect parameters such as the resulting fission product beta spectrum.« less

Pyrolysis of furan in a microreactor

NASA Astrophysics Data System (ADS)

Urness, Kimberly N.; Guan, Qi; Golan, Amir; Daily, John W.; Nimlos, Mark R.; Stanton, John F.; Ahmed, Musahid; Ellison, G. Barney

2013-09-01

A silicon carbide microtubular reactor has been used to measure branching ratios in the thermal decomposition of furan, C4H4O. The pyrolysis experiments are carried out by passing a dilute mixture of furan (approximately 0.01%) entrained in a stream of helium through the heated reactor. The SiC reactor (0.66 mm i.d., 2 mm o.d., 2.5 cm long) operates with continuous flow. Experiments were performed with a reactor inlet pressure of 100-300 Torr and a wall temperature between 1200 and 1600 K; characteristic residence times in the reactor are 60-150 μs. The unimolecular decomposition pathway of furan is confirmed to be: furan (+ M) rightleftharpoons α-carbene or β-carbene. The α-carbene fragments to CH2=C=O + HC≡CH while the β-carbene isomerizes to CH2=C=CHCHO. The formyl allene can isomerize to CO + CH3C≡CH or it can fragment to H + CO + HCCCH2. Tunable synchrotron radiation photoionization mass spectrometry is used to monitor the products and to measure the branching ratio of the two carbenes as well as the ratio of [HCCCH2]/[CH3C≡CH]. The results of these pyrolysis experiments demonstrate a preference for 80%-90% of furan decomposition to occur via the β-carbene. For reactor temperatures of 1200-1400 K, no propargyl radicals are formed. As the temperature rises to 1500-1600 K, at most 10% of the decomposition of CH2=C=CHCHO produces H + CO + HCCCH2 radicals. Thermodynamic conditions in the reactor have been modeled by computational fluid dynamics and the experimental results are compared to the predictions of three furan pyrolysis mechanisms. Uncertainty in the pressure-dependency of the initiation reaction rates is a possible a source of discrepancy between experimental results and theoretical predictions.

ETRCF, TRA654, INTERIOR. CAMERA IS ON MAIN FLOOR. NOTE CRANE ...

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

ETR-CF, TRA-654, INTERIOR. CAMERA IS ON MAIN FLOOR. NOTE CRANE HOOKS. ELECTRICAL EQUIPMENT IS PART OF PAST EXPERIMENT. DOOR AT LEFT EDGE OF VIEW LEADS TO REACTOR SERVICE BUILDING, TRA-635. INL NEGATIVE NO. HD24-1-2. Mike Crane, Photographer, ca. 2003 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Introduction to Chemical Engineering Reactor Analysis: A Web-Based Reactor Design Game

ERIC Educational Resources Information Center

Orbey, Nese; Clay, Molly; Russell, T.W. Fraser

2014-01-01

An approach to explain chemical engineering through a Web-based interactive game design was developed and used with college freshman and junior/senior high school students. The goal of this approach was to demonstrate how to model a lab-scale experiment, and use the results to design and operate a chemical reactor. The game incorporates both…

Development of advanced strain diagnostic techniques for reactor environments.

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

Fleming, Darryn D.; Holschuh, Thomas Vernon,; Miller, Timothy J.

2013-02-01

The following research is operated as a Laboratory Directed Research and Development (LDRD) initiative at Sandia National Laboratories. The long-term goals of the program include sophisticated diagnostics of advanced fuels testing for nuclear reactors for the Department of Energy (DOE) Gen IV program, with the future capability to provide real-time measurement of strain in fuel rod cladding during operation in situ at any research or power reactor in the United States. By quantifying the stress and strain in fuel rods, it is possible to significantly improve fuel rod design, and consequently, to improve the performance and lifetime of the cladding.more » During the past year of this program, two sets of experiments were performed: small-scale tests to ensure reliability of the gages, and reactor pulse experiments involving the most viable samples in the Annulated Core Research Reactor (ACRR), located onsite at Sandia. Strain measurement techniques that can provide useful data in the extreme environment of a nuclear reactor core are needed to characterize nuclear fuel rods. This report documents the progression of solutions to this issue that were explored for feasibility in FY12 at Sandia National Laboratories, Albuquerque, NM.« less

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. .

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

An analysis of decommissioning costs for the AFRRI TRIGA reactor facility

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

Forsbacka, Matt

1990-07-01

A decommissioning cost analysis for the AFRRI TRIGA Reactor Facility was made. AFRRI is not at this time suggesting that the AFRRI TRIGA Reactor Facility be decommissioned. This report was prepared to be in compliance with paragraph 50.33 of Title 10, Code of Federal Regulations which requires the assurance of availability of future decommissioning funding. The planned method of decommissioning is the immediate decontamination of the AFRRI TRIGA Reactor site to allow for restoration of the site to full public access - this is called DECON. The cost of DECON for the AFRRI TRIGA Reactor Facility in 1990 dollars ismore » estimated to be $3,200,000. The anticipated ancillary costs of facility site demobilization and spent fuel shipment is an additional $600,000. Thus the total cost of terminating reactor operations at AFRRI will be about $3,800,000. The primary basis for this cost estimate is a study of the decommissioning costs of a similar reactor facility that was performed by Battelle Pacific Northwest Laboratory (PNL) as provided in USNRC publication NUREG/CR-1756. The data in this study were adapted to reflect the decommissioning requirements of the AFRRI TRIGA. (author)« less

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.

Experimental and Computational Study of the Hydrodynamics of Trickle Bed Flow Reactor Operating Under Different Pressure Conditions

NASA Astrophysics Data System (ADS)

Rabbani, S.; Ben Salem, I.; Nadeem, H.; Kurnia, J. C.; Shamim, T.; Sassi, M.

2014-12-01

Pressure drop estimation and prediction of liquid holdup play a crucial role in design and operation of trickle bed reactors. Experiments are performed for Light Gas Oil (LGO)-nitrogen system in ambient temperature conditions in an industrial pilot plant with reactor height 0.79 m and diameter of 0.0183 m and pressure ranging from atmospheric to 10 bars. It was found that pressure drop increased with increase in system pressure, superficial gas velocity and superficial liquid velocity. It was demonstrated in the experiments that liquid holdup of the system increases with the increase in superficial liquid velocity and tends to decrease with increase in superficial gas velocity which is in good agreement with existing literature. Similar conditions were also simulated using CFD-software FLUENT. The Volume of Fluid (VoF) technique was employed in combination with "discrete particle approach" and results were compared with that of experiments. The overall pressure drop results were compared with the different available models and a new comprehensive model was proposed to predict the pressure drop in Trickle Bed Flow Reactor.

Optimization of food waste hydrolysis in leach bed coupled with methanogenic reactor: effect of pH and bulking agent.

PubMed

Xu, Su Yun; Lam, Hoi Pui; Karthikeyan, O Parthiba; Wong, Jonathan W C

2011-02-01

The effects of pH and bulking agents on hydrolysis/acidogenesis of food waste were studied using leach bed reactor (LBR) coupled with methanogenic up-flow anaerobic sludge blanket (UASB) reactor. The hydrolysis rate under regulated pH (6.0) was studied and compared with unregulated one during initial experiment. Then, the efficacies of five different bulking agents, i.e. plastic full particles, plastic hollow sphere, bottom ash, wood chip and saw dust were experimented under the regulated pH condition. Leachate recirculation with 50% water replacement was practiced throughout the experiment. Results proved that the daily leachate recirculation with pH control (6.0) accelerated the hydrolysis rate (59% higher volatile fatty acids) and methane production (up to 88%) compared to that of control without pH control. Furthermore, bottom ash improved the reactor alkalinity, which internally buffered the system that improved the methane production rate (0.182 l CH(4)/g VS(added)) than other bulking agents. Copyright © 2010 Elsevier Ltd. All rights reserved.

The νGeN experiment at the Kalinin Nuclear Power Plant

NASA Astrophysics Data System (ADS)

Belov, V.; Brudanin, V.; Egorov, V.; Filosofov, D.; Fomina, M.; Gurov, Yu.; Korotkova, L.; Lubashevskiy, A.; Medvedev, D.; Pritula, R.; Rozova, I.; Rozov, S.; Sandukovsky, V.; Timkin, V.; Yakushev, E.; Yurkowski, J.; Zhitnikov, I.

2015-12-01

The ν GeN is new experiment at the Kalinin Nuclear Power Plant (KNPP) for detection of coherent Neutrino-Ge Nucleus elastic scattering. Recent neutrino and Dark Matter search experiments have revolutionized the detection of rear events, and rear events with low energies, in particular. Experiments have achieved sensitivities on the level of several events per hundred kg of detector material per day with energy thresholds from few hundred eV. This opens up a new unique possibility for experimental detection of neutrino-nucleus coherent scattering that has been considered to be impossible so far. The νGeN project uses low threshold high-purity Ge-detectors (HPGe) developed by JINR (Dubna, Russia) in collaboration with BSI (Baltic Scientific Instruments, Riga, Latvia) for creation of a setup designated for first observation of neutrino coherent scattering on Ge. As a powerful neutrino source the experiment will use electron antineutrinos from one of the power-generating units (reactor unit #3) of the KNPP. The coherent neutrino scattering will be observed using a differential method that compares 1) the spectra measured at the reactor operation and shut-down periods; 2) the spectra measured at different distances from the reactor core during the reactor operation. For a setup placed at a 10 m distance from the center of reactor core and with an energy threshold of 350 eV up to tens of events corresponding to neutrino coherent scattering on Ge are expected to be detected per day in the constructed setup with four HPGe low-energy-threshold detectors (~ 400 grams each). The setup sensitivity will be even more increased by using new detectors with total mass up to 5 kg.

RELAP5-3D Modeling of Heat Transfer Components (Intermediate Heat Exchanger and Helical-Coil Steam Generator) for NGNP Application

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

N. A. Anderson; P. Sabharwall

2014-01-01

The Next Generation Nuclear Plant project is aimed at the research and development of a helium-cooled high-temperature gas reactor that could generate both electricity and process heat for the production of hydrogen. The heat from the high-temperature primary loop must be transferred via an intermediate heat exchanger to a secondary loop. Using RELAP5-3D, a model was developed for two of the heat exchanger options a printed-circuit heat exchanger and a helical-coil steam generator. The RELAP5-3D models were used to simulate an exponential decrease in pressure over a 20 second period. The results of this loss of coolant analysis indicate thatmore » heat is initially transferred from the primary loop to the secondary loop, but after the decrease in pressure in the primary loop the heat is transferred from the secondary loop to the primary loop. A high-temperature gas reactor model should be developed and connected to the heat transfer component to simulate other transients.« less

Oxidation of aluminum alloy cladding for research and test reactor fuel

NASA Astrophysics Data System (ADS)

Kim, Yeon Soo; Hofman, G. L.; Robinson, A. B.; Snelgrove, J. L.; Hanan, N.

2008-08-01

The oxide thicknesses on aluminum alloy cladding were measured for the test plates from irradiation tests RERTR-6 and 7A in the ATR (advanced test reactor). The measured thicknesses were substantially lower than those of test plates with similar power from other reactors available in the literature. The main reason is believed to be due to the lower pH (pH 5.1-5.3) of the primary coolant water in the ATR than in the other reactors (pH 5.9-6.5) for which we have data. An empirical model for oxide film thickness predictions on aluminum alloy used as fuel cladding in the test reactors was developed as a function of irradiation time, temperature, surface heat flux, pH, and coolant flow rate. The applicable ranges of pH and coolant flow rates cover most research and test reactors. The predictions by the new model are in good agreement with the in-pile test data available in the literature as well as with the RERTR test data measured in the ATR.

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

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

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

1991-01-01

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

Study of carbon dioxide gas treatment based on equations of kinetics in plasma discharge reactor

NASA Astrophysics Data System (ADS)

Abedi-Varaki, Mehdi

2017-08-01

Carbon dioxide (CO2) as the primary greenhouse gas, is the main pollutant that is warming earth. CO2 is widely emitted through the cars, planes, power plants and other human activities that involve the burning of fossil fuels (coal, natural gas and oil). Thus, there is a need to develop some method to reduce CO2 emission. To this end, this study investigates the behavior of CO2 in dielectric barrier discharge (DBD) plasma reactor. The behavior of different species and their reaction rates are studied using a zero-dimensional model based on equations of kinetics inside plasma reactor. The results show that the plasma reactor has an effective reduction on the CO2 density inside the reactor. As a result of reduction in the temporal variations of reaction rate, the speed of chemical reactions for CO2 decreases and very low concentration of CO2 molecules inside the plasma reactor is generated. The obtained results are compared with the existing experimental and simulation findings in the literature.

Radiation chemistry for modern nuclear energy development

NASA Astrophysics Data System (ADS)

Chmielewski, Andrzej G.; Szołucha, Monika M.

2016-07-01

Radiation chemistry plays a significant role in modern nuclear energy development. Pioneering research in nuclear science, for example the development of generation IV nuclear reactors, cannot be pursued without chemical solutions. Present issues related to light water reactors concern radiolysis of water in the primary circuit; long-term storage of spent nuclear fuel; radiation effects on cables and wire insulation, and on ion exchangers used for water purification; as well as the procedures of radioactive waste reprocessing and storage. Radiation effects on materials and enhanced corrosion are crucial in current (II/III/III+) and future (IV) generation reactors, and in waste management, deep geological disposal and spent fuel reprocessing. The new generation of reactors (III+ and IV) impose new challenges for radiation chemists due to their new conditions of operation and the usage of new types of coolant. In the case of the supercritical water-cooled reactor (SCWR), water chemistry control may be the key factor in preventing corrosion of reactor structural materials. This paper mainly focuses on radiation effects on long-term performance and safety in the development of nuclear power plants.

Emergency core cooling system

DOEpatents

Schenewerk, William E.; Glasgow, Lyle E.

1983-01-01

A liquid metal cooled fast breeder reactor provided with an emergency core cooling system includes a reactor vessel which contains a reactor core comprising an array of fuel assemblies and a plurality of blanket assemblies. The reactor core is immersed in a pool of liquid metal coolant. The reactor also includes a primary coolant system comprising a pump and conduits for circulating liquid metal coolant to the reactor core and through the fuel and blanket assemblies of the core. A converging-diverging venturi nozzle with an intermediate throat section is provided in between the assemblies and the pump. The intermediate throat section of the nozzle is provided with at least one opening which is in fluid communication with the pool of liquid sodium. In normal operation, coolant flows from the pump through the nozzle to the assemblies with very little fluid flowing through the opening in the throat. However, when the pump is not running, residual heat in the core causes fluid from the pool to flow through the opening in the throat of the nozzle and outwardly through the nozzle to the assemblies, thus providing a means of removing decay heat.

TRAC-PF1/MOD1 support calculations for the MIST/OTIS program

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

Fujita, R.K.; Knight, T.D.

1984-01-01

We are using the Transient Reactor Analysis Code (TRAC), specifically version TRAC-PF1/MOD1, to perform analyses in support of the MultiLoop Integral-System Test (MIST) and the Once-Through Integral-System (OTIS) experiment program. We have analyzed Geradrohr Dampferzeuger Anlage (GERDA) Test 1605AA to benchmark the TRAC-PF1/MOD1 code against phenomena expected to occur in a raised-loop B and W plant during a small-break loss-of-coolant accident (SBLOCA). These results show that the code can calculate both single- and two-phase natural circulation, flow interruption, boiler-condenser-mode (BCM) heat transfer, and primary-system refill in a B and W-type geometry with low-elevation auxiliary feedwater. 19 figures, 7 tables.

Eastern Europe Research Reactor Initiative nuclear education and training courses - Current activities and future challenges

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

Snoj, L.; Sklenka, L.; Rataj, J.

2012-07-01

The Eastern Europe Research Reactor Initiative was established in January 2008 to enhance cooperation between the Research Reactors in Eastern Europe. It covers three areas of research reactor utilisation: irradiation of materials and fuel, radioisotope production, neutron beam experiments, education and training. In the field of education and training an EERRI training course was developed. The training programme has been elaborated with the purpose to assist IAEA Member States, which consider building a research reactor (RR) as a first step to develop nuclear competence and infrastructure in the Country. The major strength of the reactor is utilisation of three differentmore » research reactors and a lot of practical exercises. Due to high level of adaptability, the course can be tailored to specific needs of institutions with limited or no access to research reactors. (authors)« less

Accumulation of radioactive corrosion products on steel surfaces of VVER-type nuclear reactors. II. 60Co

NASA Astrophysics Data System (ADS)

Varga, Kálmán; Hirschberg, Gábor; Németh, Zoltán; Myburg, Gerrit; Schunk, János; Tilky, Péter

2001-10-01

In the case of intact fuel claddings, the predominant source of radioactivity in the primary circuits of water-cooled nuclear reactors is the activation of corrosion products in the core. The most important corrosion product radionuclides in the primary coolant of pressurized water reactors (PWRs) are 60Co, 58Co, 51Cr, 54Mn, 59Fe (as well as 110mAg in some Soviet-made VVER-type reactor). The second part of this series is focused on the complex studies of the formation and build-up of 60Co-containing species on an austenitic stainless steel type 08X18H10T (GOST 5632-61) and magnetite-covered carbon steel often to be used in Soviet-planned VVERs. The kinetics and mechanism of the cobalt accumulation were studied by a combination (coupling) of an in situ radiotracer method and voltammetry in a model solution of the primary circuit coolant. In addition, independent techniques such as X-ray photoelectron spectroscopic (XPS) and ICP-OES are also used to analyze the chemical state of Co species in the passive layer formed on stainless steel as well as the chemical composition of model solution. The experimental results have revealed that: (i) The passive behavior of the austenitic stainless steel at open-circuit conditions, the slightly alkaline pH and the reducing water chemistry can be considered to be optimal to minimize the 60Co contamination. (ii) The highly potential dependent deposition of various Co-oxides at E>1.10 V (vs. RHE) offers a unique possibility to elaborate a novel electrochemical method for the decrease or removal of cobalt traces from borate-containing coolants contaminated with 60Co and/or 58Co radionuclides.

A comparison of BNR activated sludge systems with membrane and settling tank solid-liquid separation.

PubMed

Ramphao, M C; Wentzel, M C; Ekama, G A; Alexander, W V

2006-01-01

Installing membranes for solid-liquid separation into biological nutrient removal (BNR) activated sludge (AS) systems makes a profound difference not only to the design of the membrane bio-reactor (MBR) BNR system itself, but also to the design approach for the whole wastewater treatment plant (WWTP). In multi-zone BNR systems with membranes in the aerobic reactor and fixed volumes for the anaerobic, anoxic and aerobic zones (i.e. fixed volume fractions), the mass fractions can be controlled (within a range) with the inter-reactor recycle ratios. This zone mass fraction flexibility is a significant advantage of MBR BNR systems over BNR systems with secondary settling tanks (SSTs), because it allows changing the mass fractions to optimise biological N and P removal in conformity with influent wastewater characteristics and the effluent N and P concentrations required. For PWWF/ADWF ratios (fq) in the upper range (fq approximately 2.0), aerobic mass fractions in the lower range (f(maer) < 0.60) and high (usually raw) wastewater strengths, the indicated mode of operation of MBR BNR systems is as extended aeration WWTPs (no primary settling and long sludge age). However, the volume reduction compared with equivalent BNR systems with SSTs will not be large (40-60%), but the cost of the membranes can be offset against sludge thickening and stabilisation costs. Moving from a flow unbalanced raw wastewater system to a flow balanced (fq = 1) low (usually settled) wastewater strength system can double the ADWF capacity of the biological reactor, but the design approach of the WWTP changes away from extended aeration to include primary sludge stabilisation. The cost of primary sludge treatment then has to be offset against the savings of the increased WWTP capacity.

An exploratory study to determine applicability of nano-hardness and micro-compression measurements for yield stress estimation

NASA Astrophysics Data System (ADS)

Hosemann, P.; Swadener, J. G.; Kiener, D.; Was, G. S.; Maloy, S. A.; Li, N.

2008-03-01

The superior properties of ferritic/martensitic steels in a radiation environment (low swelling, low activation under irradiation and good corrosion resistance) make them good candidates for structural parts in future reactors and spallation sources. While it cannot substitute for true reactor experiments, irradiation by charged particles from accelerators can reduce the number of reactor experiments and support fundamental research for a better understanding of radiation effects in materials. Based on the nature of low energy accelerator experiments, only a small volume of material can be uniformly irradiated. Micro and nanoscale post irradiation tests thus have to be performed. We show here that nanoindentation and micro-compression testing on T91 and HT-9 stainless steel before and after ion irradiation are useful methods to evaluate the radiation induced hardening.

Corrosion of aluminum clad spent nuclear fuel in the 70 ton cask during transfer from L area to H-canyon

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

Mickalonis, J. I.

2015-08-31

Aluminum-clad spent nuclear fuel will be transported for processing in the 70-ton nuclear fuel element cask from L Basin to H-canyon. During transport these fuels would be expected to experience high temperature aqueous corrosion from the residual L Basin water that will be present in the cask. Cladding corrosion losses during transport were calculated for material test reactor (MTR) and high flux isotope reactors (HFIR) fuels using literature and site information on aqueous corrosion at a range of time/temperature conditions. Calculations of the cladding corrosion loss were based on Arrhenius relationships developed for aluminum alloys typical of cladding material withmore » the primary assumption that an adherent passive film does not form to retard the initial corrosion rate. For MTR fuels a cladding thickness loss of 33 % was found after 1 year in the cask with a maximum temperature of 263 °C. HFIR fuels showed a thickness loss of only 6% after 1 year at a maximum temperature of 180 °C. These losses are not expected to impact the overall confinement function of the aluminum cladding.« less

Corrosion of aluminum clad spent nuclear fuel in the 70 ton cask during transfer from L area to H-canyon

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

Mickalonis, J. I.

2015-08-01

Aluminum-clad spent nuclear fuel will be transported for processing in the 70-ton nuclear fuel element cask from L Basin to H-canyon. During transport these fuels would be expected to experience high temperature aqueous corrosion from the residual L Basin water that will be present in the cask. Cladding corrosion losses during transport were calculated for material test reactor (MTR) and high flux isotope reactors (HFIR) fuels using literature and site information on aqueous corrosion at a range of time/temperature conditions. Calculations of the cladding corrosion loss were based on Arrhenius relationships developed for aluminum alloys typical of cladding material withmore » the primary assumption that an adherent passive film does not form to retard the initial corrosion rate. For MTR fuels a cladding thickness loss of 33% was found after 1 year in the cask with a maximum temperature of 263 °C. HFIR fuels showed a thickness loss of only 6% after 1 year at a maximum temperature of 180 °C. These losses are not expected to impact the overall confinement function of the aluminum cladding.« less

Biogasification products of water hyacinth wastewater reclamation

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

Chynoweth, D.P.; Biljetina, R.; Srivastava, V.J.

1984-01-01

This paper describes the results of research in progress to evaluate the use of water hyacinth for wastewater treatment and subsequent conversion of hyacinth and sludge to methane by anaerobic digestion. Laboratory studies have been directed toward evaluating advanced biogasification concepts and establishing a data base for the design and operation of an experimental test unit (ETU) located at the water hyacinth wastewater treatment facility at Walt Disney World (WDW) located in Kissimmee, Florida. Laboratory-scale kinetic experiments have been conducted using continuously-stirred tank reactors (CSTR) and a novel non-mixed upflow solids reactor (USR) receiving a hyacinth/sludge blend at retention timesmore » of 15 down to 2.1 days. The data suggest that best performance is achieved in the USR which has longer solids and organism retention. A larger-scale ETU (160 cu ft) was designed and installed at WDW in 1983 and started up in 1984. The purpose of this unit is to validate laboratory experiments and to evaluate larger-scale equipment used for chopping, slurry preparation, mixing, and effluent dewatering. The ETU includes a front end designed for multiple feed processing and storage, a fully instrumented USR digester, and tanks for effluent and gas storage. The ETU is currently being operated on a 2:1 blend (dry wt basis) of water hyacinth and primary sludge. Performance is good without major operational problems. Results of laboratory studies and start-up and operation of the ETU will be presented. 7 references, 4 figures, 1 table.« less

Nucleation and growth of lead oxide particles in liquid lead-bismuth eutectic.

PubMed

Gladinez, Kristof; Rosseel, Kris; Lim, Jun; Marino, Alessandro; Heynderickx, Geraldine; Aerts, Alexander

2017-10-18

Liquid lead-bismuth eutectic (LBE) is an important candidate to become the primary coolant of future, generation IV, nuclear fast reactors and Accelerator Driven System (ADS) concepts. One of the main challenges with the use of LBE as a coolant is to avoid its oxidation which results in solid lead oxide (PbO) precipitation. The chemical equilibria governing PbO formation are well understood. However, insufficient kinetic information is currently available for the development of LBE-based nuclear technology. Here, we report the results of experiments in which the nucleation, growth and dissolution of PbO in LBE during temperature cycling are measured by monitoring dissolved oxygen using potentiometric oxygen sensors. The metastable region, above which PbO nucleation can occur, has been determined under conditions relevant for the operation of LBE cooled nuclear systems and was found to be independent of setup geometry and thus thought to be widely applicable. A kinetic model to describe formation and dissolution of PbO particles in LBE is proposed, based on Classical Nucleation Theory (CNT) combined with mass transfer limited growth and dissolution. This model can accurately predict the experimentally observed changes in oxygen concentration due to nucleation, growth and dissolution of PbO, using the effective interfacial energy of a PbO nucleus in LBE as a fitting parameter. The results are invaluable to evaluate the consequences of oxygen ingress in LBE cooled nuclear systems under normal operating and accidental conditions and form the basis for the development of cold trap technology to avoid PbO formation in the primary reactor circuit.

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.

Validation of CESAR Thermal-hydraulic Module of ASTEC V1.2 Code on BETHSY Experiments

NASA Astrophysics Data System (ADS)

Tregoures, Nicolas; Bandini, Giacomino; Foucher, Laurent; Fleurot, Joëlle; Meloni, Paride

The ASTEC V1 system code is being jointly developed by the French Institut de Radioprotection et Sûreté Nucléaire (IRSN) and the German Gesellschaft für Anlagen und ReaktorSicherheit (GRS) to address severe accident sequences in a nuclear power plant. Thermal-hydraulics in primary and secondary system is addressed by the CESAR module. The aim of this paper is to present the validation of the CESAR module, from the ASTEC V1.2 version, on the basis of well instrumented and qualified integral experiments carried out in the BETHSY facility (CEA, France), which simulates a French 900 MWe PWR reactor. Three tests have been thoroughly investigated with CESAR: the loss of coolant 9.1b test (OECD ISP N° 27), the loss of feedwater 5.2e test, and the multiple steam generator tube rupture 4.3b test. In the present paper, the results of the code for the three analyzed tests are presented in comparison with the experimental data. The thermal-hydraulic behavior of the BETHSY facility during the transient phase is well reproduced by CESAR: the occurrence of major events and the time evolution of main thermal-hydraulic parameters of both primary and secondary circuits are well predicted.

Final report of the decontamination and decommissioning of the BORAX-V facility turbine building

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

Arave, A.E.; Rodman, G.R.

1992-12-01

The Boiling Water Reactor Experiment (BORAX)-V Facility Turbine Building Decontamination and Decommissioning (D&D) Project is described in this report. The BORAX series of five National Reactor Testing Station (NRTS) reactors pioneered intensive work on boiling water reactor (BWR) experiments conducted between 1953 and 1964. Facility characterization, decision analyses, and D&D plans for the turbine building were prepared from 1979 through 1990. D&D activities of the turbine building systems were initiated in November of 1988 and completed with the demolition and backfill of the concrete foundation in March 1992. Due to the low levels of radioactivity and the absence of loosemore » contamination, the D&D activities were completed with no radiation exposure to the workers. The D&D activities were performed in a manner that no radiological health or safety hazard to the public or to personnel at the Idaho National Engineering Laboratory (INEL) remain.« less

Final report of the decontamination and decommissioning of the BORAX-V facility turbine building

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

Arave, A.E.; Rodman, G.R.

1992-12-01

The Boiling Water Reactor Experiment (BORAX)-V Facility Turbine Building Decontamination and Decommissioning (D D) Project is described in this report. The BORAX series of five National Reactor Testing Station (NRTS) reactors pioneered intensive work on boiling water reactor (BWR) experiments conducted between 1953 and 1964. Facility characterization, decision analyses, and D D plans for the turbine building were prepared from 1979 through 1990. D D activities of the turbine building systems were initiated in November of 1988 and completed with the demolition and backfill of the concrete foundation in March 1992. Due to the low levels of radioactivity and themore » absence of loose contamination, the D D activities were completed with no radiation exposure to the workers. The D D activities were performed in a manner that no radiological health or safety hazard to the public or to personnel at the Idaho National Engineering Laboratory (INEL) remain.« less

Investigation of a Possibility of Chromium-51 Accumulation in the SM-3 Reactor to Fabricate a Neutrino Source

NASA Astrophysics Data System (ADS)

Romanov, E. G.; Gavrin, V. N.; Tarasov, V. A.; Malkov, A. P.; Kupriyanov, A. V.; Danshin, S. N.; Veretenkin, E. P.

2017-01-01

Compact high intensity neutrino sources based on 51Cr isotope are demanded for very short baseline neutrino experiments. In particular, a 3 MCi 51Cr neutrino source is needed for the experiment BEST on search for transitions of electron neutrinos to sterile states. The paper presents the results of the analysis of options of the irradiation of highly enriched 50Cr in the existing trap of thermal neutrons of high-flux reactor SM-3, as well as using the most promising variants of the trap after upcoming reconstruction of the reactor. It is shown that it is possible to to obtain the intensity of 51Cr up to 3.85 MCi at the end of irradiation of 50Cr enriched to 97% in the high-flux reactor SM-3 of the JSC “SSC NIIAR”.

Operation of an aquatic worm reactor suitable for sludge reduction at large scale.

PubMed

Hendrickx, Tim L G; Elissen, Hellen H J; Temmink, Hardy; Buisman, Cees J N

2011-10-15

Treatment of domestic waste water results in the production of waste sludge, which requires costly further processing. A biological method to reduce the amount of waste sludge and its volume is treatment in an aquatic worm reactor. The potential of such a worm reactor with the oligochaete Lumbriculus variegatus has been shown at small scale. For scaling up purposes, a new configuration of the reactor was designed, in which the worms were positioned horizontally in the carrier material. This was tested in a continuous experiment of 8 weeks where it treated all the waste sludge from a lab-scale activated sludge process. The results showed a higher worm growth rate compared to previous experiments with the old configuration, whilst nutrient release was similar. The new configuration has a low footprint and allows for easy aeration and faeces collection, thereby making it suitable for full scale application. Copyright © 2011 Elsevier Ltd. All rights reserved.

New approach to control the methanogenic reactor of a two-phase anaerobic digestion system.

PubMed

von Sachs, Jürgen; Meyer, Ulrich; Rys, Paul; Feitkenhauer, Heiko

2003-03-01

A new control strategy for the methanogenic reactor of a two-phase anaerobic digestion system has been developed and successfully tested on the laboratory scale. The control strategy serves the purpose to detect inhibitory effects and to achieve good conversion. The concept is based on the idea that volatile fatty acids (VFA) can be measured in the influent of the methanogenic reactor by means of titration. Thus, information on the output (methane production) and input of the methanogenic reactor is available, and a (carbon) mass balance can be obtained. The control algorithm comprises a proportional/integral structure with the ratio of (a) the methane production rate measured online and (b) a maximum methane production rate expected (derived from the stoichiometry) as a control variable. The manipulated variable is the volumetric feed rate. Results are shown for an experiment with VFA (feed) concentration ramps and for experiments with sodium chloride as inhibitor.

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

Graslund, C.; Hellstrand, E.

Sweden benefits in many ways from the reactor safety research performed in other countries. Its own activity complements this effort, but a certain fraction is oriented toward safety issues that are intimately related to the special design of the ASEA-ATOM boiling-water reactor. Through the availability of the decommissioned Marviken reactor plant, Sweden has been able to play a leading role in integral containment experiments with international participation. Joint efforts with other countries are now devoted to defining new large-scale experiments to be performed in the unique Marviken facility. The largest portion of the safety research program in Sweden is performedmore » by Studsvik Energiteknik AB, but various universities, consultant firms, and research institutes are also involved. In addition, a substantial amount of work is done by the reactor vendor ASEA-ATOM. The overall annual budget is at present between $7 and $8 million, with three governmental authorities as the main financing bodies.« less

METHOD AND APPARATUS FOR HANDLING RADIOACTIVE PRODUCTS

DOEpatents

Nicoll, D.

1959-02-24

A device is described for handling fuel elements being discharged from a nuclear reactor. The device is adapted to be disposed beneath a reactor within the storage canal for spent fuel elements. The device is comprised essentially of a cylinder pivotally mounted to a base for rotational motion between a vertical position. where the mouth of the cylinder is in the top portion of the container for receiving a fuel element discharged from a reactor into the cylinder, and a horizontal position where the mouth of the cylinder is remote from the top portion of the container and the fuel element is discharged from the cylinder into the storage canal. The device is operated by hydraulic pressure means and is provided with a means to prevent contaminated primary liquid coolant in the reactor system from entering the storage canal with the spent fuel element.

Oak Ridge National Laboratory Support of Non-light Water Reactor Technologies: Capabilities Assessment for NRC Near-term Implementation Action Plans for Non-light Water Reactors

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

Belles, Randy; Jain, Prashant K.; Powers, Jeffrey J.

The Oak Ridge National Laboratory (ORNL) has a rich history of support for light water reactor (LWR) and non-LWR technologies. The ORNL history involves operation of 13 reactors at ORNL including the graphite reactor dating back to World War II, two aqueous homogeneous reactors, two molten salt reactors (MSRs), a fast-burst health physics reactor, and seven LWRs. Operation of the High Flux Isotope Reactor (HFIR) has been ongoing since 1965. Expertise exists amongst the ORNL staff to provide non-LWR training; support evaluation of non-LWR licensing and safety issues; perform modeling and simulation using advanced computational tools; run laboratory experiments usingmore » equipment such as the liquid salt component test facility; and perform in-depth fuel performance and thermal-hydraulic technology reviews using a vast suite of computer codes and tools. Summaries of this expertise are included in this paper.« less

SHIPPINGPORT OPERATIONS FROM POWER OPERATION AFTER FIRST REFUELING TO SECOND REFUELING, MAY 6, 1960 TO AUGUST 16, 1961

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

None

1963-10-31

A report of Shippingport operation during Seed 2 lifetime is presented. The information is primarily confined to the nuclear portion of the operation. A general review of station performance is given along with details of reactor physics, reactor thermal and hydraulic performance, reactor plant performance and modifications, operational chemistry, and radioactive contamination experience. (J.R.D.)

Tensile and Fatigue Testing and Material Hardening Model Development for 508 LAS Base Metal and 316 SS Similar Metal Weld under In-air and PWR Primary Loop Water Conditions

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

Mohanty, Subhasish; Soppet, William; Majumdar, Saurin

This report provides an update on an assessment of environmentally assisted fatigue for light water reactor components under extended service conditions. This report is a deliverable in September 2015 under the work package for environmentally assisted fatigue under DOE’s Light Water Reactor Sustainability program. In an April 2015 report we presented a baseline mechanistic finite element model of a two-loop pressurized water reactor (PWR) for systemlevel heat transfer analysis and subsequent thermal-mechanical stress analysis and fatigue life estimation under reactor thermal-mechanical cycles. In the present report, we provide tensile and fatigue test data for 508 low-alloy steel (LAS) base metal,more » 508 LAS heat-affected zone metal in 508 LAS–316 stainless steel (SS) dissimilar metal welds, and 316 SS-316 SS similar metal welds. The test was conducted under different conditions such as in air at room temperature, in air at 300 oC, and under PWR primary loop water conditions. Data are provided on materials properties related to time-independent tensile tests and time-dependent cyclic tests, such as elastic modulus, elastic and offset strain yield limit stress, and linear and nonlinear kinematic hardening model parameters. The overall objective of this report is to provide guidance to estimate tensile/fatigue hardening parameters from test data. Also, the material models and parameters reported here can directly be used in commercially available finite element codes for fatigue and ratcheting evaluation of reactor components under in-air and PWR water conditions.« less

Passive heat transfer means for nuclear reactors

DOEpatents

Burelbach, James P.

1984-01-01

An improved passive cooling arrangement is disclosed for maintaining adjacent or related components of a nuclear reactor within specified temperature differences. Specifically, heat pipes are operatively interposed between the components, with the vaporizing section of the heat pipe proximate the hot component operable to cool it and the primary condensing section of the heat pipe proximate the other and cooler component operable to heat it. Each heat pipe further has a secondary condensing section that is located outwardly beyond the reactor confinement and in a secondary heat sink, such as air ambient the containment, that is cooler than the other reactor component. Means such as shrouding normally isolated the secondary condensing section from effective heat transfer with the heat sink, but a sensor responds to overheat conditions of the reactor to open the shrouding, which thereby increases the cooling capacity of the heat pipe. By having many such heat pipes, an emergency passive cooling system is defined that is operative without electrical power.

The present situations and perspectives on utilization of research reactors in Thailand

NASA Astrophysics Data System (ADS)

Chongkum, Somporn

2002-01-01

The Thai Research Reactor 1/Modification 1, a TRIGA Mark III reactor, went critical on November 7, 1977. It has been playing a central role in the development of both Office of Atomic Energy for Peace (OAEP) and nuclear application in Thailand. It has a maximum power of 2 MW (thermal) at steady state and a pulsing capacity of 2000 MW. The highest thermal neutron flux at a central thimber is 1×10 13 n/cm 2/s, which is extensively utilized for radioisotope production, neutron activation analysis and neutron beam experiments, i.e. neutron scattering, prompt gamma analysis and neutron radiography. Following the nuclear technological development, the OAEP is in the process of establishing the Ongkharak Nuclear Research Center (ONRC). The center is being built in Nakhon Nayok province, 60 km northeast of Bangkok. The centerpiece of the ONRC is a multipurpose 10 MW TRIGA research reactor. Facilities are included for the production of radioisotopes for medicine, industry and agriculture, neutron transmutation doping of silicon, and neutron capture therapy. The neutron beam facilities will also be utilized for applied research and technology development as well as training in reactor operations, performance of experiments and reactor physics. This paper describes a recent program of utilization as well as a new research reactor for enlarging the perspectives of its utilization in the future.

Electronics for the STEREO experiment

NASA Astrophysics Data System (ADS)

HÉLAINE, Victor; STEREO Collaboration

2017-09-01

The STEREO experiment, aiming to probe short baseline neutrino oscillations by precisely measuring reactor anti-neutrino spectrum, is currently under installation. It is located at short distance from the compact research reactor core of the Institut Laue-Langevin, Grenoble, France. Dedicated electronics, hosted in a single µTCA crate, were designed for this experiment. In this article, the electronics requirements, architecture and the performances achieved are described. It is shown how intrinsic Pulse Shape Discrimination properties of the liquid scintillator are preserved and how custom adaptable logic is used to improve the muon veto efficiency.

Preparation of Gd Loaded Liquid Scintillator for Daya Bay Neutrino Experiment

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

Ding Yayun; Zhang Zhiyong

2010-05-12

Gadolinium loaded liquid scintillator (Gd-LS) is an excellent target material for reactor antineutrino experiments. Ideal Gd-LS should have long attenuation length, high light yield, long term stability, low toxicity, and should be compatible with the material used to build the detector. We have developed a new Gd-LS recipe in which carboxylic acid 3,5,5-trimethylhexanoic acid is used as the complexing ligand to gadolinium, 2,5-diphenyloxazole (PPO) and 1,4-bis[2-methylstyryl]benzene (bis-MSB) are used as primary fluor and wavelength shifter, respectively. The scintillator base is linear alkyl benzene (LAB). Eight hundred liters of Gd-LS has been synthesized and tested in a prototype detector. Results showmore » that the Gd-LS has high quality and is suitable for underground experiments in large quantity. Large scale production facility has been built. A full batch production of 4 t Gd-LS has been produced and monitored for several months. The production of 180 t Gd-LS will be carried out in the near future.« less

New results from RENO and future RENO-50 project

NASA Astrophysics Data System (ADS)

Kim, S. B.

2017-07-01

RENO (Reactor Experiment for Neutrino Oscillation) has obtained a more precise value of the smallest mixing angle θ_{13} and the first result on neutrino squared-mass difference \\vertΔ m_{ee}^2\\vert from an energy- and baseline-dependent disappearance of reactor electron antineutrinos (overline{ν}_e) using 500 days of data. Based on the ratio of inverse-beta-decay (IBD) prompt spectra measured between two identical far and near detectors, we obtain sin^2 2 θ_{13} = 0.082 ± 0.009({stat.}) ± 0.006({syst.}) and \\vertΔ m_{ee}^2\\vert =[2.62_{-0.23}^{+0.21}({stat.}) _{-0.13} ^{+0.12}({syst.})]× 10^{-3} eV2. An excess of reactor antineutrinos near 5MeV is observed in the measured prompt spectrum with respect to the most commonly used models. The excess is found to be consistent with coming from reactors. A future reactor experiment of RENO-50 is proposed to determine the neutrino mass hierarchy and to make highly precise measurements of θ_{12} , Δ m_{21}^2 , and \\vertΔ m_{ee}^2\\vert.

Yale High Energy Physics Research: Precision Studies of Reactor Antineutrinos

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

Heeger, Karsten M.

2014-09-13

This report presents experimental research at the intensity frontier of particle physics with particular focus on the study of reactor antineutrinos and the precision measurement of neutrino oscillations. The experimental neutrino physics group of Professor Heeger and Senior Scientist Band at Yale University has had leading responsibilities in the construction and operation of the Daya Bay Reactor Antineutrino Experiment and made critical contributions to the discovery of non-zeromore » $$\\theta_{13}$$. Heeger and Band led the Daya Bay detector management team and are now overseeing the operations of the antineutrino detectors. Postdoctoral researchers and students in this group have made leading contributions to the Daya Bay analysis including the prediction of the reactor antineutrino flux and spectrum, the analysis of the oscillation signal, and the precision determination of the target mass yielding unprecedented precision in the relative detector uncertainty. Heeger's group is now leading an R\\&D effort towards a short-baseline oscillation experiment, called PROSPECT, at a US research reactor and the development of antineutrino detectors with advanced background discrimination.« less

Performance of a composite membrane bioreactor treating toluene vapors: inocula selection, reactor performance and behavior under transient conditions.

PubMed

Kumar, Amit; Dewulf, Jo; Vercruyssen, Aline; Van Langenhove, Herman

2009-04-01

In this study, a membrane biofilm reactor performance for toluene as a model pollutant is presented. A composite membrane consisting of a porous polyacrylonitrile (PAN) support layer coated with a very thin (0.3 microm) dense polydimethylsiloxane (PDMS) top layer was used. Batch experiments were performed to select an appropriate inocula (slaughterhouse wastewater treatment sludge with a specific toluene consumption rate of 118+/-23 microg g(-1) VSS L(-1)) among the three available sources of inoculums. The maximum elimination capacity gas-side reactor volume based (EC)v and membrane based (EC)(m, max) obtained were 609 g m(-3) h(-1) and 1.2 g m(-2) h(-1) respectively, which is much higher than other membrane bioreactors. Further experiments involved the study of the membrane biofilm reactor flexibility when operational parameters as temperature, loading rate etc. were modified. In all cases, the membrane biofilm reactor showed a rapid adaptation and new steady-states were obtained within hours. Overall, the results illustrate that membrane bioreactors can potentially be a good option for treatment of air pollutants such as toluene.

Gas-Liquid Two-Phase Flows Through Packed Bed Reactors in Microgravity

NASA Technical Reports Server (NTRS)

Motil, Brian J.; Balakotaiah, Vemuri

2001-01-01

The simultaneous flow of gas and liquid through a fixed bed of particles occurs in many unit operations of interest to the designers of space-based as well as terrestrial equipment. Examples include separation columns, gas-liquid reactors, humidification, drying, extraction, and leaching. These operations are critical to a wide variety of industries such as petroleum, pharmaceutical, mining, biological, and chemical. NASA recognizes that similar operations will need to be performed in space and on planetary bodies such as Mars if we are to achieve our goals of human exploration and the development of space. The goal of this research is to understand how to apply our current understanding of two-phase fluid flow through fixed-bed reactors to zero- or partial-gravity environments. Previous experiments by NASA have shown that reactors designed to work on Earth do not necessarily function in a similar manner in space. Two experiments, the Water Processor Assembly and the Volatile Removal Assembly have encountered difficulties in predicting and controlling the distribution of the phases (a crucial element in the operation of this type of reactor) as well as the overall pressure drop.

Numerical and Experimental Thermal Responses of Single-cell and Differential Calorimeters: from Out-of-Pile Calibration to Irradiation Campaigns

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

Brun, J.; Reynard-Carette, C.; Carette, M.

2015-07-01

The nuclear radiation energy deposition rate (usually expressed in W.g{sup -1}) is a key parameter for the thermal design of experiments, on materials and nuclear fuel, carried out in experimental channels of irradiation reactors such as the French OSIRIS reactor in Saclay or inside the Polish MARIA reactor. In particular the quantification of the nuclear heating allows to predicting the heat and thermal conditions induced in the irradiation devices or/and structural materials. Various sensors are used to quantify this parameter, in particular radiometric calorimeters also called in-pile calorimeters. Two main kinds of in-pile calorimeter exist with in particular specific designs:more » single-cell calorimeter and differential calorimeter. The present work focuses on these two calorimeter kinds from their out-of-pile calibration step (transient and steady experiments respectively) to comparison between numerical and experimental results obtained from two irradiation campaigns (MARIA reactor and OSIRIS reactor respectively). The main aim of this paper is to propose a steady numerical approach to estimate the single-cell calorimeter response under irradiation conditions. (authors)« less

Naval Reactors Prime Contractor Team (NRPCT) Experiences and Considerations With Irradiation Test Performance in an International Environment

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

MH Lane

2006-02-15

This letter forwards a compilation of knowledge gained regarding international interactions and issues associated with Project Prometheus. The following topics are discussed herein: (1) Assessment of international fast reactor capability and availability; (2) Japanese fast reactor (JOYO) contracting strategy; (3) NRPCT/Program Office international contract follow; (4) Completion of the Japan Atomic Energy Agency (JAEA)/Pacific Northwest National Laboratory (PNNL) contract for manufacture of reactor test components; (5) US/Japanese Departmental interactions and required Treaties and Agreements; and (6) Non-technical details--interactions and considerations.

Benchmark tests of JENDL-3.2 for thermal and fast reactors

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

Takano, Hideki; Akie, Hiroshi; Kikuchi, Yasuyuki

1994-12-31

Benchmark calculations for a variety of thermal and fast reactors have been performed by using the newly evaluated JENDL-3 Version-2 (JENDL-3.2) file. In the thermal reactor calculations for the uranium and plutonium fueled cores of TRX and TCA, the k{sub eff} and lattice parameters were well predicted. The fast reactor calculations for ZPPR-9 and FCA assemblies showed that the k{sub eff} reactivity worths of Doppler, sodium void and control rod, and reaction rate distribution were in a very good agreement with the experiments.

SLSF in-reactor local fault safety experiment P4. Final report

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

Thompson, D. H.; Holland, J. W.; Braid, T. H.

The Sodium Loop Safety Facility (SLSF), a major facility in the US fast-reactor safety program, has been used to simulate a variety of sodium-cooled fast reactor accidents. SLSF experiment P4 was conducted to investigate the behavior of a "worse-than-case" local fault configuration. Objectives of this experiment were to eject molten fuel into a 37-pin bundle of full-length Fast-Test-Reactor-type fuel pins form heat-generating fuel canisters, to characterize the severity of any molten fuel-coolant interaction, and to demonstrate that any resulting blockage could either be tolerated during continued power operation or detected by global monitors to prevent fuel failure propagation. The designmore » goal for molten fuel release was 10 to 30 g. Explusion of molten fuel from fuel canisters caused failure of adjacent pins and a partial flow channel blockage in the fuel bundle during full-power operation. Molten fuel and fuel debris also lodged against the inner surface of the test subassembly hex-can wall. The total fuel disruption of 310 g evaluated from posttest examination data was in excellent agreement with results from the SLSF delayed neutron detection system, but exceeded the target molten fuel release by an order of magnitude. This report contains a summary description of the SLSF in-reactor loop and support systems and the experiment operations. results of the detailed macro- and microexamination of disrupted fuel and metal and results from the analysis of the on-line experimental data are described, as are the interpretations and conclusions drawn from the posttest evaluations. 60 refs., 74 figs.« less

Simulation of Oil Palm Shell Pyrolysis to Produce Bio-Oil with Self-Pyrolysis Reactor

NASA Astrophysics Data System (ADS)

Fika, R.; Nelwan, L. O.; Yulianto, M.

2018-05-01

A new self-pyrolysis reactor was designed to reduce the utilization of electric heater due to the energy saving for the production of bio-oil from oil palm shell. The yield of the bio- oil was then evaluated with the developed mathematical model by Sharma [1] with the characteristic of oil palm shell [2]. During the simulation, the temperature on the combustion chamber on the release of the bio-oil was utilized to determine the volatile composition from the combustion of the oil palm shell as fuel. The mass flow was assumed constant for three experiments. The model resulted in a significant difference between the simulated bio-oil and experiments. The bio-oil yields from the simulation were 22.01, 16.36, and 21.89 % (d.b.) meanwhile the experimental yields were 10.23, 9.82, and 8.41% (d.b.). The char yield varied from 30.7 % (d.b.) from the simulation to 40.9 % (d.b.) from the experiment. This phenomenon was due to the development of process temperature over time which was not considered as one of the influential factors in producing volatile matters on the simulation model. Meanwhile the real experiments highly relied on the process conditions (reactor type, temperature over time, gas flow). There was also possibilities of the occurrence of the gasification inside the reactor which caused the liquid yield was not as high as simulated. Further simulation model research on producing the bio-oil yield will be needed to predict the optimum condition and temperature development on the newly self-pyrolysis reactor.

Reactor ν̄ e disappearance in the Double Chooz experiment

DOE PAGES

Abe, Y.; Aberle, C.; dos Anjos, J. C.; ...

2012-09-18

The Double Chooz experiment has observed 8249 candidate electron antineutrino events in 227.93 live days with 33.71 GW-ton-years (reactor power×detector mass×live time) exposure using a 10.3 m³ fiducial volume detector located at 1050 m from the reactor cores of the Chooz nuclear power plant in France. The expectation in case of θ₁₃=0 is 8937 events. The deficit is interpreted as evidence of electron antineutrino disappearance. From a rate plus spectral shape analysis we find sin²2θ₁₃=0.109±0.030(stat)±0.025(syst). The data exclude the no-oscillation hypothesis at 99.8% CL (2.9σ).

Investigation of the Feasibility of Utilizing Gamma Emission Computed Tomography in Evaluating Fission Product Migration in Irradiated TRISO Fuel Experiments

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

Jason M. Harp; Paul A. Demkowicz

2014-10-01

In the High Temperature Gas-Cooled Reactor (HTGR) the TRISO particle fuel serves as the primary fission product containment. However the large number of TRISO particles present in proposed HTGRs dictates that there will be a small fraction (~10 -4 to 10 -5) of as manufactured and in-pile particle failures that will lead to some fission product release. The matrix material surrounding the TRISO particles in fuel compacts and the structural graphite holding the TRISO particles in place can also serve as sinks for containing any released fission products. However data on the migration of solid fission products through these materialsmore » is lacking. One of the primary goals of the AGR-3/4 experiment is to study fission product migration from failed TRISO particles in prototypic HTGR components such as structural graphite and compact matrix material. In this work, the potential for a Gamma Emission Computed Tomography (GECT) technique to non-destructively examine the fission product distribution in AGR-3/4 components and other irradiation experiments is explored. Specifically, the feasibility of using the Idaho National Laboratory (INL) Hot Fuels Examination Facility (HFEF) Precision Gamma Scanner (PGS) system for this GECT application is considered. To test the feasibility, the response of the PGS system to idealized fission product distributions has been simulated using Monte Carlo radiation transport simulations. Previous work that applied similar techniques during the AGR-1 experiment will also be discussed as well as planned uses for the GECT technique during the post irradiation examination of the AGR-2 experiment. The GECT technique has also been applied to other irradiated nuclear fuel systems that were currently available in the HFEF hot cell including oxide fuel pins, metallic fuel pins, and monolithic plate fuel.« less

Comparison of two-stage thermophilic (68 degrees C/55 degrees C) anaerobic digestion with one-stage thermophilic (55 degrees C) digestion of cattle manure.

PubMed

Nielsen, H B; Mladenovska, Z; Westermann, P; Ahring, B K

2004-05-05

A two-stage 68 degrees C/55 degrees C anaerobic degradation process for treatment of cattle manure was studied. In batch experiments, an increase of the specific methane yield, ranging from 24% to 56%, was obtained when cattle manure and its fractions (fibers and liquid) were pretreated at 68 degrees C for periods of 36, 108, and 168 h, and subsequently digested at 55 degrees C. In a lab-scale experiment, the performance of a two-stage reactor system, consisting of a digester operating at 68 degrees C with a hydraulic retention time (HRT) of 3 days, connected to a 55 degrees C reactor with 12-day HRT, was compared with a conventional single-stage reactor running at 55 degrees C with 15-days HRT. When an organic loading of 3 g volatile solids (VS) per liter per day was applied, the two-stage setup had a 6% to 8% higher specific methane yield and a 9% more effective VS-removal than the conventional single-stage reactor. The 68 degrees C reactor generated 7% to 9% of the total amount of methane of the two-stage system and maintained a volatile fatty acids (VFA) concentration of 4.0 to 4.4 g acetate per liter. Population size and activity of aceticlastic methanogens, syntrophic bacteria, and hydrolytic/fermentative bacteria were significantly lower in the 68 degrees C reactor than in the 55 degrees C reactors. The density levels of methanogens utilizing H2/CO2 or formate were, however, in the same range for all reactors, although the degradation of these substrates was significantly lower in the 68 degrees C reactor than in the 55 degrees C reactors. Temporal temperature gradient electrophoresis profiles (TTGE) of the 68 degrees C reactor demonstrated a stable bacterial community along with a less divergent community of archaeal species. Copyright 2004 Wiley Periodicals, Inc.

Monochloramine Cometabolism by Mixed-Culture Nitrifiers under Drinking Water Conditions

EPA Science Inventory

The current research investigated monochloramine cometabolism by nitrifying mixed cultures grown under drinking water relevant conditions and harvested from sand-packed reactors before conducting suspended growth batch kinetic experiments. Three batch reactors were used in each ...

MTR BASEMENT. GENERAL ELECTRIC CONTROL CONSOLE FOR AIRCRAFT NUCLEAR PROPULSION ...

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

MTR BASEMENT. GENERAL ELECTRIC CONTROL CONSOLE FOR AIRCRAFT NUCLEAR PROPULSION EXPERIMENT NO. 1. INL NEGATIVE NO. 6510. Unknown Photographer, 9/29/1959 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Performance of low smeared density sodium-cooled fast reactor metal fuel

NASA Astrophysics Data System (ADS)

Porter, D. L.; Chichester, H. J. M.; Medvedev, P. G.; Hayes, S. L.; Teague, M. C.

2015-10-01

An experiment was performed in the Experimental Breeder Rector-II (EBR-II) in the 1990s to show that metallic fast reactor fuel could be used in reactors with a single, once-through core. To prove the long duration, high burnup, high neutron exposure capability an experiment where the fuel pin was designed with a very large fission gas plenum and very low fuel smeared density (SD). The experiment, X496, operated to only 8.3 at.% burnup because the EBR-II reactor was scheduled for shut-down at that time. Many of the examinations of the fuel pins only funded recently with the resurgence of reactor designs using very high-burnup fuel. The results showed that, despite the low smeared density of 59% the fuel swelled radially to contact the cladding, fission gas release appeared to be slightly higher than demonstrated in conventional 75%SD fuel tests and axial growth was about the same as 75% SD fuel. There were axial positions in some of the fuel pins which showed evidence of fuel restructuring and an absence of fission products with low melting points and gaseous precursors (Cs and Rb). A model to investigate whether these areas may have overheated due to a loss of bond sodium indicates that it is a possible explanation for the fuel restructuring and something to be considered for fuel performance modeling of low SD fuel.

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

CRITICAL TESTS FOR PRT REACTOR

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

Triplett, J.R.; Anderson, J.K.; Dunn, R.E.

1960-07-01

Critical teste to be performed on the Plutonium Recycle Te st Heactor are described. Exponential, approach-tocritical, critical, and substitution experiments will be carried out. These experiments include: calibration of moderator level; determination of the wori of various fuel loadings; calibration of the shim system including determination of maximum control strength of the entire system; substitution experiments to determine reflector savings, void effects, effects of H/sub 2/O and degraded D/sub 2/O coolants, and effects of loop and other material intsllations; determination of fuel-plus-coolant and moderator temperature coefficients; and kinetic experiments to determine response of the reactor to reactivity changes. (M.C.G.)

Biodegradation of pharmaceuticals in hospital wastewater by a hybrid biofilm and activated sludge system (Hybas).

PubMed

Escolà Casas, Mònica; Chhetri, Ravi Kumar; Ooi, Gordon; Hansen, Kamilla M S; Litty, Klaus; Christensson, Magnus; Kragelund, Caroline; Andersen, Henrik R; Bester, Kai

2015-10-15

Hospital wastewater contributes a significant input of pharmaceuticals into municipal wastewater. The combination of suspended activated sludge and biofilm processes, as stand-alone or as hybrid process (hybrid biofilm and activated sludge system (Hybas™)) has been suggested as a possible solution for hospital wastewater treatment. To investigate the potential of such a hybrid system for the removal of pharmaceuticals in hospital wastewater a pilot plant consisting of a series of one activated sludge reactor, two Hybas™ reactors and one moving bed biofilm reactor (MBBR) has been established and adapted during 10 months of continuous operation. After this adaption phase batch and continuous experiments were performed for the determination of degradation of pharmaceuticals. Removal of organic matter and nitrification mainly occurred in the first reactor. Most pharmaceuticals were removed significantly. The removal of pharmaceuticals (including X-ray contrast media, β-blockers, analgesics and antibiotics) was fitted to a single first-order kinetics degradation function, giving degradation rate constants from 0 to 1.49 h(-1), from 0 to 7.78 × 10(-1)h(-1), from 0 to 7.86 × 10(-1)h(-1) and from 0 to 1.07 × 10(-1)h(-1) for first, second, third and fourth reactors respectively. Generally, the highest removal rate constants were found in the first and third reactors while the lowest were found in the second one. When the removal rate constants were normalized to biomass amount, the last reactor (biofilm only) appeared to have the most effective biomass in respect to removing pharmaceuticals. In the batch experiment, out of 26 compounds, 16 were assessed to degrade more than 20% of the respective pharmaceutical within the Hybas™ train. In the continuous flow experiments, the measured removals were similar to those estimated from the batch experiments, but the concentrations of a few pharmaceuticals appeared to increase during the first treatment step. Such increase could be attributed to de-conjugation or formation from other metabolites. Copyright © 2015 Elsevier B.V. All rights reserved.

Processes and systems for the production of propylene glycol from glycerol

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

Frye, John G; Oberg, Aaron A; Zacher, Alan H

2015-01-20

Processes and systems for converting glycerol to propylene glycol are disclosed. The glycerol feed is diluted with propylene glycol as the primary solvent, rather than water which is typically used. The diluted glycerol feed is sent to a reactor where the glycerol is converted to propylene glycol (as well as other byproducts) in the presence of a catalyst. The propylene glycol-containing product from the reactor is recycled as a solvent for the glycerol feed.

Treatment of fruit-juice industry wastewater in a two-stage anaerobic hybrid (AH) reactor system followed by a sequencing batch reactor (SBR).

PubMed

Tawfik, A; El-Kamah, H

2012-01-01

This study has been carried out to assess the performance of a combined system consisting of an anaerobic hybrid (AH) reactor followed by a sequencing batch reactor (SBR) for treatment of fruit-juice industry wastewater at a temperature of 26 degrees C. Three experimental runs were conducted in this investigation. In the first experiment, a single-stage AH reactor was operated at a hydraulic retention time (HRT) of 10.2 h and organic loading rate (OLR) of 11.8 kg COD m(-3) d(-1). The reactor achieved a removal efficiency of 42% for chemical oxygen demand (COD), 50.8% for biochemical oxygen demand (BOD5), 50.3% for volatile fatty acids (VFA) and 56.4% for total suspended solids (TSS). In the second experiment, two AH reactors connected in series achieved a higher removal efficiency for COD (67.4%), BOD5 (77%), and TSS (71.5%) at a total HRT of 20 h and an OLR of 5.9 kg COD m(-3) d(-1). For removal of the remaining portions of COD, BOD5 and TSS from the effluent of the two-stage AH system, a sequencing batch reactor (SBR) was investigated as a post-treatment unit. The reactor achieved a substantial reduction in total COD, resulting in an average effluent concentration of 50 mg L(-1) at an HRT of 11 h and OLR of 5.3 kg COD m(-3) d(-1). Almost complete removal of total BOD5 and oil and grease was achieved, i.e. 10 mg L(-1) and 1.2 mg L(-1), respectively, remained in the final effluent of the SBR.

Dismantlement of the TSF-SNAP Reactor Assembly

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

Peretz, Fred J

2009-01-01

This paper describes the dismantlement of the Tower Shielding Facility (TSF)?Systems for Nuclear Auxiliary Power (SNAP) reactor, a SNAP-10A reactor used to validate radiation source terms and shield performance models at Oak Ridge National Laboratory (ORNL) from 1967 through 1973. After shutdown, it was placed in storage at the Y-12 National Security Complex (Y-12), eventually falling under the auspices of the Highly Enriched Uranium (HEU) Disposition Program. To facilitate downblending of the HEU present in the fuel elements, the TSF-SNAP was moved to ORNL on June 24, 2006. The reactor assembly was removed from its packaging, inspected, and the sodium-potassiummore » (NaK) coolant was drained. A superheated steam process was used to chemically react the residual NaK inside the reactor assembly. The heat exchanger assembly was removed from the top of the reactor vessel, and the criticality safety sleeve was exchanged for a new safety sleeve that allowed for the removal of the vessel lid. A chain-mounted tubing cutter was used to separate the lid from the vessel, and the 36 fuel elements were removed and packaged in four U.S. Department of Transportation 2R/6M containers. The fuel elements were returned to Y-12 on July 13, 2006. The return of the fuel elements and disposal of all other reactor materials accomplished the formal objectives of the dismantlement project. In addition, a project model was established for the handling of a fully fueled liquid-metal?cooled reactor assembly. Current criticality safety codes have been benchmarked against experiments performed by Atomics International in the 1950s and 1960s. Execution of this project provides valuable experience applicable to future projects addressing space and liquid-metal-cooled reactors.« less

ORNL-TNS/PEPR overall heating requirements

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

Peng, Y. K.M.; Rome, J. A.

1977-01-01

The ORNL TNS/PEPR studies have the objectives of (1) leading to a system that demonstrates the fusion reactor core in the mid-to-late 1980's and extrapolates to an economic tokamak power reactor, and (2) providing a near-term focus for the scientific and technological programs toward the power reactor. This discussion of the overall heating requirements for the ORNL TNS/PEPR is concerned with the neutral beams as the primary heating method, the electron-cyclotron resonance (ECR) heating at a lower power level for profile control, and the upper hybrid resonance (UHR) initiation and preheating of currentless plasmas to reduce current start-up loop voltagemore » (V/sub l/) requirements.« less

Research on pressure control of pressurizer in pressurized water reactor nuclear power plant

NASA Astrophysics Data System (ADS)

Dai, Ling; Yang, Xuhong; Liu, Gang; Ye, Jianhua; Qian, Hong; Xue, Yang

2010-07-01

Pressurizer is one of the most important components in the nuclear reactor system. Its function is to keep the pressure of the primary circuit. It can prevent shutdown of the system from the reactor accident under the normal transient state while keeping the setting value in the normal run-time. This paper is mainly research on the pressure system which is running in the Daya Bay Nuclear Power Plant. A conventional PID controller and a fuzzy controller are designed through analyzing the dynamic characteristics and calculating the transfer function. Then a fuzzy PID controller is designed by analyzing the results of two controllers. The fuzzy PID controller achieves the optimal control system finally.

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.

Probing light sterile neutrino signatures at reactor and Spallation Neutron Source neutrino experiments

NASA Astrophysics Data System (ADS)

Kosmas, T. S.; Papoulias, D. K.; Tórtola, M.; Valle, J. W. F.

2017-09-01

We investigate the impact of a fourth sterile neutrino at reactor and Spallation Neutron Source neutrino detectors. Specifically, we explore the discovery potential of the TEXONO and COHERENT experiments to subleading sterile neutrino effects through the measurement of the coherent elastic neutrino-nucleus scattering event rate. Our dedicated χ2-sensitivity analysis employs realistic nuclear structure calculations adequate for high purity sub-keV threshold Germanium detectors.

Applying Bayesian Neural Network to determine neutrino incoming direction in reactor neutrino experiments and supernova explosion location by scintillator detectors

NASA Astrophysics Data System (ADS)

Xu, W. W.; Xu, Y.; Meng, Y. X.; Wu, B.

2009-01-01

In the paper, it is discussed by using Monte-Carlo simulation that the Bayesian Neural Network (BNN) is applied to determine neutrino incoming direction in reactor neutrino experiments and supernova explosion location by scintillator detectors. As a result, compared to the method in ref. [1], the uncertainty on the measurement of the neutrino direction using BNN is significantly improved. The uncertainty on the measurement of the reactor neutrino direction is about 1.0° at the 68.3% C.L., and the one in the case of supernova neutrino is about 0.6° at the 68.3% C.L. . Compared to the method in ref. [1], the uncertainty attainable by using BNN reduces by a factor of about 20. And compared to the Super-Kamiokande experiment (SK), it reduces by a factor of about 8.

Flat-plate solar array project. Task 1: Silicon material. Investigation of the hydrochlorination of SiC14

NASA Technical Reports Server (NTRS)

Mui, J. Y. P.

1982-01-01

A two inch diameter stainless steel reactor was designed and built to operate at pressures up to 500 psig for the experimental studies on the hydrochlorination of SiCl4 and metallurgical grade (m.g.) silicon metal to SiHCl3. In order to clearly see the effect of pressure, the experiments were carried out at low reactor pressures of 73 psig and 150 psig, respectively. A large pressure effect on the hydrochlorination reaction was observed between the results of the low pressure experiments and the results of the high pressure experiments. In general, higher pressure produces a higher conversion of SiHCl3, but at a lower reaction rate. The effect of temperature on the reaction rate was studied at 73 psig. Higher reaction temperature gave a higher conversion and a higher reaction rate. Samples of the materials used to construct the hydrochlorination reactor were prepared for corrosion tests.

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

NASA Astrophysics Data System (ADS)

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

2017-02-01

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

Supplemental Thermal-Hydraulic Transient Analyses of BR2 in Support of Conversion to LEU Fuel

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

Licht, J.; Dionne, B.; Sikik, E.

2016-01-01

Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The RELAP5/Mod 3.3 code has been used to perform transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. A RELAP5 model of BR2 has been validated against select transient BR2 reactor experiments performed in 1963 by showingmore » agreement with measured cladding temperatures. Following the validation, the RELAP5 model was then updated to represent the current use of the reactor; taking into account core configuration, neutronic parameters, trip settings, component changes, etc. Simulations of the 1963 experiments were repeated with this updated model to re-evaluate the boiling risks associated with the currently allowed maximum heat flux limit of 470 W/cm 2 and temporary heat flux limit of 600 W/cm 2. This document provides analysis of additional transient simulations that are required as part of a modern BR2 safety analysis report (SAR). The additional simulations included in this report are effect of pool temperature, reduced steady-state flow rate, in-pool loss of coolant accidents, and loss of external cooling. The simulations described in this document have been performed for both an HEU- and LEU-fueled core.« less

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.

Equipment for neutron measurements at VR-1 Sparrow training reactor.

PubMed

Kolros, Antonin; Huml, Ondrej; Kríz, Martin; Kos, Josef

2010-01-01

The VR-1 sparrow reactor is an experimental nuclear facility for training, student education and teaching purposes. The sparrow reactor is an educational platform for the basic experiments at the reactor physic and dosimetry. The aim of this article is to describe the new experimental equipment EMK310 features and possibilities for neutron detection by different gas filled detectors at VR-1 reactor. Among the EMK310 equipment typical attributes belong precise set-up, simple control, resistance to electromagnetic interference, high throughput (counting rate), versatility and remote controllability. The methods for non-linearity correction of pulse neutron detection system and reactimeter application are presented. Copyright 2009. Published by Elsevier Ltd.

A Measurement of the Absolute Reactor Antineutrino Flux and Spectrum at Daya Bay

NASA Astrophysics Data System (ADS)

An, Fengpeng

2017-12-01

The Daya Bay Reactor Neutrino Experiment uses an array of eight underground detectors to study antineutrinos from six reactor cores with different baselines. Since the start of data-taking from late 2011, Daya Bay has collected the largest sample of reactor antineutrino events to date, and has made the most precise measurement of the neutrino oscillation parameters sin22θ13 and Δm2ee. Using the data from the four detectors in the near experimental halls, Daya Bay has made a high statistics measurement of the absolute reactor antineutrino flux and spectrum. In this paper we will present this measurement and its comparison to predictions based on different flux models.

Preliminary CFD study of Pebble Size and its Effect on Heat Transfer in a Pebble Bed Reactor

NASA Astrophysics Data System (ADS)

Jones, Andrew; Enriquez, Christian; Spangler, Julian; Yee, Tein; Park, Jungkyu; Farfan, Eduardo

2017-11-01

In pebble bed reactors, the typical pebble diameter used is 6cm, and within each pebble is are thousands of nuclear fuel kernels. However, efficiency of the reactor does not solely depend on the number of kernels of fuel within each graphite sphere, but also depends on the type and motion of the coolant within the voids between the spheres and the reactor itself. In this work a physical analysis of the pebble bed nuclear reactor's fluid dynamics is undertaken using Computational Fluid Dynamics software. The primary goal of this work is to observe the relationship between the different pebble diameters in an idealized alignment and the thermal transport efficiency of the reactor. The model constructed of our idealized argument will consist on stacked 8 pebble columns that fixed at the inlet on the reactor. Two different pebble sizes 4 cm and 6 cm will be studied and helium will be supplied as coolant with a fixed flow rate of 96 kg/s, also a fixed pebble surface temperatures will be used. Comparison will then be made to evaluate the efficiency of coolant to transport heat due to the varying sizes of the pebbles. Assistant Professor for the Department of Civil and Construction Engineering PhD.

A preliminary assessment of the effects of heat flux distribution and penetration on the creep rupture of a reactor vessel lower head

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

Chu, T.Y.; Bentz, J.; Simpson, R.

1997-02-01

The objective of the Lower Head Failure (LHF) Experiment Program is to experimentally investigate and characterize the failure of the reactor vessel lower head due to thermal and pressure loads under severe accident conditions. The experiment is performed using 1/5-scale models of a typical PWR pressure vessel. Experiments are performed for various internal pressure and imposed heat flux distributions with and without instrumentation guide tube penetrations. The experimental program is complemented by a modest modeling program based on the application of vessel creep rupture codes developed in the TMI Vessel Investigation Project. The first three experiments under the LHF programmore » investigated the creep rupture of simulated reactor pressure vessels without penetrations. The heat flux distributions for the three experiments are uniform (LHF-1), center-peaked (LHF-2), and side-peaked (LHF-3), respectively. For all the experiments, appreciable vessel deformation was observed to initiate at vessel wall temperatures above 900K and the vessel typically failed at approximately 1000K. The size of failure was always observed to be smaller than the heated region. For experiments with non-uniform heat flux distributions, failure typically occurs in the region of peak temperature. A brief discussion of the effect of penetration is also presented.« less

In-air and pressurized water reactor environment fatigue experiments of 316 stainless steel to study the effect of environment on cyclic hardening

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

Mohanty, Subhasish; Soppet, William K.; Majumdar, Saurindranath

Argonne National Laboratory (ANL), under the sponsorship of Department of Energy’s Light Water Reactor Sustainability (LWRS) program, is trying to develop a mechanistic approach for more accurate life estimation of LWR components. In this context, ANL has conducted many fatigue experiments under different test and environment conditions on type 316 stainless steel (316SS) material which is widely used in the US reactors. Contrary to the conventional S~N curve based empirical fatigue life estimation approach, the aim of the present DOE sponsored work is to develop an understanding of the material ageing issues more mechanistically (e.g. time dependent hardening and softening)more » under different test and environmental conditions. Better mechanistic understanding will help develop computer-based advanced modeling tools to better extrapolate stress-strain evolution of reactor components under multi-axial stress states and hence help predict their fatigue life more accurately. In this paper (part-I) the fatigue experiments under different test and environment conditions and related stress-strain results for 316 SS are discussed. In a second paper (part-II) the related evolutionary cyclic plasticity material modeling techniques and results are discussed.« less

Physics capabilities of the SNO+ experiment

NASA Astrophysics Data System (ADS)

Arushanova, E.; Back, A. R.; SNO+ Collaboration

2017-09-01

SNO+ will soon enter its first phase of physics data-taking. The Canadian-based detector forms part of the SNOLAB underground facility, in a Sudbury nickel mine; its location providing more than two kilometres of rock overburden. We present an overview of the SNO+ experiment and its physics capabilities. Our primary goal is the search for neutrinoless double-beta decay, where our expected sensitivity would place an upper limit of 1.9 × 1026 y, at 90% CL, on the half-life of neutrinoless double-beta decay in 130Te. We also intend to build on the success of SNO by studying the solar neutrino spectrum. In the unloaded scintillator phase SNO+ has the ability to make precision measurements of the fluxes of low-energy pep neutrinos and neutrinos from the CNO cycle. Other physics goals include: determining the spectrum of reactor antineutrinos, to further constrain Δ {m}122; detecting neutrinos produced by a galactic supernova and investigating certain modes of nucleon decay.

Standardized reactors for the study of medical biofilms: a review of the principles and latest modifications.

PubMed

Gomes, Inês B; Meireles, Ana; Gonçalves, Ana L; Goeres, Darla M; Sjollema, Jelmer; Simões, Lúcia C; Simões, Manuel

2018-08-01

Biofilms can cause severe problems to human health due to the high tolerance to antimicrobials; consequently, biofilm science and technology constitutes an important research field. Growing a relevant biofilm in the laboratory provides insights into the basic understanding of the biofilm life cycle including responses to antibiotic therapies. Therefore, the selection of an appropriate biofilm reactor is a critical decision, necessary to obtain reproducible and reliable in vitro results. A reactor should be chosen based upon the study goals and a balance between the pros and cons associated with its use and operational conditions that are as similar as possible to the clinical setting. However, standardization in biofilm studies is rare. This review will focus on the four reactors (Calgary biofilm device, Center for Disease Control biofilm reactor, drip flow biofilm reactor, and rotating disk reactor) approved by a standard setting organization (ASTM International) for biofilm experiments and how researchers have modified these standardized reactors and associated protocols to improve the study and understanding of medical biofilms.

Initial conceptual design study of self-critical nuclear pumped laser systems

NASA Technical Reports Server (NTRS)

Rodgers, R. J.

1979-01-01

An analytical study of self-critical nuclear pumped laser system concepts was performed. Primary emphasis was placed on reactor concepts employing gaseous uranium hexafluoride (UF6) as the fissionable material. Relationships were developed between the key reactor design parameters including reactor power level, critical mass, neutron flux level, reactor size, operating pressure, and UF6 optical properties. The results were used to select a reference conceptual laser system configuration. In the reference configuration, the 3.2 m cubed lasing volume is surrounded by a graphite internal moderator and a region of heavy water. Results of neutronics calculations yield a critical mass of 4.9 U(235) in the form (235)UF6. The configuration appears capable of operating in a continuous steady-state mode. The average gas temperature in the core is 600 K and the UF6 partial pressure within the lasing volume is 0.34 atm.

RADON LEVELS AND ЕQUIVALENT DOSE RATES AT THE IRT-SOFIA RESEARCH REACTOR SITE.

PubMed

Krezhov, Kiril; Mladenov, Aleksander; Dimitrov, Dobromir

2018-06-11

Results from radon measurements by active sampling of indoor air in the buildings within the Nuclear Scientific Experimental and Educational Centre (NSEEC) protected site at the Institute for Nuclear Research and Nuclear Energy (INRNE) are presented. The inspected buildings included in this report are the IRT research reactor structure and several auxiliary formations wherein the laundry facilities and the gamma irradiator GOU-1 (60Co source) are installed as well as the Central Alarm Station (CAS) premises. Besides the reactor hall and the primary cooling loop area, special attention was given to the premises of the First Class Radiochemical Laboratory in the IRT reactor basement. Determination of radon concentration distribution in the premises of the constructions within the site is an important part of radiation surveillance during the operation and maintenance of the NSEEC facilities as well as for their involvement in the educational activities at INRNE.

Alternate Tritium Production Methods Using A Liquid Lithium Target

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

Wilson, J.

For over 60 years, the Savannah River Site’s primary mission has been the production of tritium. From the beginning, the Savannah River National Laboratory (SRNL) has provided the technical foundation to ensure the successful execution of this critical defense mission. SRNL has developed most of the processes used in the tritium mission and provides the research and development necessary to supply this critical component. This project was executed by first developing reactor models that could be used as a neutron source. In parallel to this development calculations were carried out testing the feasibility of accelerator technologies that could also bemore » used for tritium production. Targets were designed with internal moderating material and optimized target was calculated to be capable of 3000 grams using a 1400 MWt sodium fast reactor, 850 grams using a 400 MWt sodium fast reactor, and 100 grams using a 62 MWt reactor, annually.« less

Biaxial Creep Specimen Fabrication

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

JL Bump; RF Luther

This report documents the results of the weld development and abbreviated weld qualification efforts performed by Pacific Northwest National Laboratory (PNNL) for refractory metal and superalloy biaxial creep specimens. Biaxial creep specimens were to be assembled, electron beam welded, laser-seal welded, and pressurized at PNNL for both in-pile (JOYO reactor, O-arai, Japan) and out-of-pile creep testing. The objective of this test campaign was to evaluate the creep behavior of primary cladding and structural alloys under consideration for the Prometheus space reactor. PNNL successfully developed electron beam weld parameters for six of these materials prior to the termination of the Navalmore » Reactors program effort to deliver a space reactor for Project Prometheus. These materials were FS-85, ASTAR-811C, T-111, Alloy 617, Haynes 230, and Nirnonic PE16. Early termination of the NR space program precluded the development of laser welding parameters for post-pressurization seal weldments.« less

INDEPENDENT CONFIRMATORY SURVEY REPORT FOR THE REACTOR BUILDING, HOT LABORATORY, PRIMARY PUMP HOUSE, AND LAND AREAS AT THE PLUM BROOK REACTOR FACILITY, SANDUSKY, OHIO

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

Erika N. Bailey

2011-10-10

In 1941, the War Department acquired approximately 9,000 acres of land near Sandusky, Ohio and constructed a munitions plant. The Plum Brook Ordnance Works Plant produced munitions, such as TNT, until the end of World War II. Following the war, the land remained idle until the National Advisory Committee for Aeronautics later called the National Aeronautics and Space Administration (NASA) obtained 500 acres to construct a nuclear research reactor designed to study the effects of radiation on materials used in space flight. The research reactor was put into operation in 1961 and was the first of fifteen test facilities eventuallymore » built by NASA at the Plum Brook Station. By 1963, NASA had acquired the remaining land at Plum Brook for these additional test facilities« less

Effects of dissolved CO2 on Shallow Freshwater Microbial Communities simulating a CO2 Leakage Scenario

NASA Astrophysics Data System (ADS)

Gulliver, D. M.; Lowry, G. V.; Gregory, K.

2013-12-01

Geological carbon sequestration is likely to be part of a comprehensive strategy to minimize the atmospheric release of greenhouse gasses, establishing a concern of sequestered CO2 leakage into overlying potable aquifers. Leaking CO2 may affect existing biogeochemical processes and therefore water quality. There is a critical need to understand the evolution of CO2 exposed microbial communities that influence the biogeochemistry in these freshwater aquifers. The evolution of microbial ecology for different CO2 exposure concentrations was investigated using fluid-slurry samples obtained from a shallow freshwater aquifer (55 m depth, 0.5 MPa, 22 °C, Escatawpa, MS). The microbial community of well samples upstream and downstream of CO2 injection was characterized. In addition, batch vessel experiments were conducted with the upstream aquifer samples exposed to varying pCO2 from 0% to 100% under reservoir temperature and pressure for up to 56 days. The microbial community of the in situ experiment and the batch reactor experiment were analyzed with 16S rRNA clone libraries and qPCR. In both the in situ experiment and the batch reactor experiment, DNA concentration did not correlate with CO2 exposure. Both the in situ experiment and the batch reactors displayed a changing microbial community with increased CO2 exposure. The well water isolate, Curvibacter, appeared to be the most tolerant genus to high CO2 concentrations in the in situ experiments and to mid-CO2 concentrations in the batch reactors. In batch reactors with pCO2 concentrations higher than experienced in situ (pCO2 = 0.5 MPa), Pseudomonas appeared to be the most tolerant genus. Findings provide insight into a dynamic biogeochemical system that will alter with CO2 exposure. Adapted microbial populations will eventually give rise to the community that will impact the metal mobility and water quality. Knowledge of the surviving microbial populations will enable improved models for predicting the fate of CO2 following leakage and lead to better strategies for ensuring the quality of potable aquifer water.

MODELING THE AMBIENT CONDITION EFFECTS OF AN AIR-COOLED NATURAL CIRCULATION SYSTEM

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

Hu, Rui; Lisowski, Darius D.; Bucknor, Matthew

The Reactor Cavity Cooling System (RCCS) is a passive safety concept under consideration for the overall safety strategy of advanced reactors such as the High Temperature Gas-Cooled Reactor (HTGR). One such variant, air-cooled RCCS, uses natural convection to drive the flow of air from outside the reactor building to remove decay heat during normal operation and accident scenarios. The Natural convection Shutdown heat removal Test Facility (NSTF) at Argonne National Laboratory (“Argonne”) is a half-scale model of the primary features of one conceptual air-cooled RCCS design. The facility was constructed to carry out highly instrumented experiments to study the performancemore » of the RCCS concept for reactor decay heat removal that relies on natural convection cooling. Parallel modeling and simulation efforts were performed to support the design, operation, and analysis of the natural convection system. Throughout the testing program, strong influences of ambient conditions were observed in the experimental data when baseline tests were repeated under the same test procedures. Thus, significant analysis efforts were devoted to gaining a better understanding of these influences and the subsequent response of the NSTF to ambient conditions. It was determined that air humidity had negligible impacts on NSTF system performance and therefore did not warrant consideration in the models. However, temperature differences between the building exterior and interior air, along with the outside wind speed, were shown to be dominant factors. Combining the stack and wind effects together, an empirical model was developed based on theoretical considerations and using experimental data to correlate zero-power system flow rates with ambient meteorological conditions. Some coefficients in the model were obtained based on best fitting the experimental data. The predictive capability of the empirical model was demonstrated by applying it to the new set of experimental data. The empirical model was also implemented in the computational models of the NSTF using both RELAP5-3D and STARCCM+ codes. Accounting for the effects of ambient conditions, simulations from both codes predicted the natural circulation flow rates very well.« less

A study on using fireclay as a biomass carrier in an activated sludge system.

PubMed

Tilaki, Ramazan Ali Dianati

2011-01-01

By adding a biomass carrier to an activated sludge system, the biomass concentration will increase, and subsequently the organic removal efficiency will be enhanced. In this study, the possibility of using excess sludge from ceramic and tile manufacturing plants as a biomass carrier was investigated. The aim of this study was to determine the effect of using fireclay as a biomass carrier on biomass concentration, organic removal and nitrification efficiency in an activated sludge system. Experiments were conducted by using a bench scale activated sludge system operating in batch and continuous modes. Artificial simulated wastewater was made by using recirculated water in a ceramic manufacturing plant. In the continuous mode, hydraulic detention time in the aeration reactor was 8 and 22 h. In the batch mode, aeration time was 8 and 16 h. Fireclay doses were 500, 1,400 and 2,250 mg l(-1), and were added to the reactors in each experiment separately. The reactor with added fireclay was called a Hybrid Biological Reactor (HBR). A reactor without added fireclay was used as a control. Efficiency parameters such as COD, MLVSS and nitrate were measured in the control and HBR reactors according to standard methods. The average concentration of biomass in the HBR reactor was greater than in the control reactor. The total biomass concentration in the HBR reactor (2.25 g l(-1) fireclay) in the continuous mode was 3,000 mg l(-1) and in the batch mode was 2,400 mg l(-1). The attached biomass concentration in the HBR reactor (2.25 g l(-1) fireclay) in the continuous mode was 1,500 mg l(-1) and in the batch mode was 980 mg l(-1). Efficiency for COD removal in the HBR and control reactor was 95 and 55%, respectively. In the HBR reactor, nitrification was enhanced, and the concentration of nitrate was increased by 80%. By increasing the fireclay dose, total and attached biomass was increased. By adding fireclay as a biomass carrier, the efficiency of an activated sludge system to treat wastewater from ceramic manufacturing plants was increased.

Impact of intermittent aerations on leachate quality and greenhouse gas reduction in the aerobic-anaerobic landfill method.

PubMed

Nag, Mitali; Shimaoka, Takayuki; Komiya, Teppei

2016-09-01

The aerobic-anaerobic landfill method (AALM) is a novel approach in solid waste management that could shorten the landfill post-closure period and minimize the environmental loads. In this study, the aerobic-anaerobic landfill method was evaluated by using intermittent aeration. In addition, the nitrification-denitrification process was assessed as a means of reducing the emission of greenhouse gases (GHGs) and improving the leachate quality during the degradation of the organic solid waste. The leachate quality and the gas composition in each of the reactors were measured during the experimental period (408days). The aeration process entailed the injection of air into plexiglass cylinders (200cm height×10 cm diameter), filled with fresh organic solid waste collected from a composting plant. Different aeration routines were applied, namely, continuous aeration (aerobic reactor A), aeration for three days/week (aerobic-anaerobic reactor B), aeration for 6h/day (aerobic-anaerobic reactor C), and no aeration (non-aerated reactor D). It was found that aerobic reactor A produced the best results in terms of reduction of GHGs and improvement of the leachate quality. The aerobic-anaerobic reactor C was found to be more effective than reactor B in respect of both the emission of GHGs and the leachate quality; moreover, compared with aerobic reactor A, energy costs were reduced by operating this reactor. The transition period phenomenon was investigated during an intensive seven-day experiment conducted on the discharged leachate obtained from aerobic-anaerobic reactors B and C. The experiment concerned the differences in the composition of the gas during the aeration and the non-aeration periods. It was found that the transition period between the aeration and non-aeration cycles, which followed the simultaneous nitrification-denitrification had a considerable effect on the leachate quality of both the reactors. The results indicated that AALM has the potential to reduce leachate pollutants and the emission of GHGs. Furthermore, the occurrence of simultaneous nitrification-denitrification presents the prospect that intermittent aeration could reduce landfill aftercare and energy costs. Copyright © 2015 Elsevier Ltd. All rights reserved.

Catalytic Tar Reduction for Assistance in Thermal Conversion of Space Waste for Energy Production

NASA Technical Reports Server (NTRS)

Caraccio, Anne Joan; Devor, Robert William; Hintze, Paul E.; Muscatello, Anthony C.; Nur, Mononita

2014-01-01

The Trash to Gas (TtG) project investigates technologies for converting waste generated during spaceflight into various resources. One of these technologies was gasification, which employed a downdraft reactor designed and manufactured at NASA's Kennedy Space Center (KSC) for the conversion of simulated space trash to carbon dioxide. The carbon dioxide would then be converted to methane for propulsion and water for life support systems. A minor byproduct of gasification includes large hydrocarbons, also known as tars. Tars are unwanted byproducts that add contamination to the product stream, clog the reactor and cause complications in analysis instrumentation. The objective of this research was to perform reduction studies of a mock tar using select catalysts and choose the most effective for primary treatment within the KSC downdraft gasification reactor. Because the KSC reactor is operated at temperatures below typical gasification reactors, this study evaluates catalyst performance below recommended catalytic operating temperatures. The tar reduction experimentation was observed by passing a model tar vapor stream over the catalysts at similar conditions to that of the KSC reactor. Reduction in tar was determined using gas chromatography. Tar reduction efficiency and catalyst performances were evaluated at different temperatures.

Biodegradation of 2,4,6-trichlorophenol in a packed-bed biofilm reactor equipped with an internal net draft tube riser for aeration and liquid circulation.

PubMed

Jesús, A Gómez-De; Romano-Baez, F J; Leyva-Amezcua, L; Juárez-Ramírez, C; Ruiz-Ordaz, N; Galíndez-Mayer, J

2009-01-30

For the aerobic biodegradation of the fungicide and defoliant 2,4,6-trichlorophenol (2,4,6-TCP), a bench-scale packed-bed bioreactor equipped with a net draft tube riser for liquid circulation and oxygenation (PB-ALR) was constructed. To obtain a high packed-bed volume relative to the whole bioreactor volume, a high A(D)/A(R) ratio was used. Reactor's downcomer was packed with a porous support of volcanic stone fragments. PB-ALR hydrodynamics and oxygen mass transfer behavior was evaluated and compared to the observed behavior of the unpacked reactor operating as an internal airlift reactor (ALR). Overall gas holdup values epsilon(G), and zonal oxygen mass transfer coefficients determined at various airflow rates in the PB-ALR, were higher than those obtained with the ALR. When comparing mixing time values obtained in both cases, a slight increment in mixing time was observed when reactor was operated as a PB-ALR. By using a mixed microbial community, the biofilm reactor was used to evaluate the aerobic biodegradation of 2,4,6-TCP. Three bacterial strains identified as Burkholderia sp., Burkholderia kururiensis and Stenotrophomonas sp. constituted the microbial consortium able to cometabolically degrade the 2,4,6-TCP, using phenol as primary substrate. This consortium removed 100% of phenol and near 99% of 2,4,6-TCP. Mineralization and dehalogenation of 2,4,6-TCP was evidenced by high COD removal efficiencies ( approximately 95%), and by the stoichiometric release of chloride ions from the halogenated compound ( approximately 80%). Finally, it was observed that the microbial consortium was also capable to metabolize 2,4,6-TCP without phenol as primary substrate, with high removal efficiencies (near 100% for 2,4,6-TCP, 92% for COD and 88% for chloride ions).

New results from RENO & prospects with RENO-50

NASA Astrophysics Data System (ADS)

Joo, K. K.

2017-09-01

This paper briefly describes recent progress of RENO and next generation future prospect of the reactor neutrino oscillation experiment, RENO-50. Recently the RENO experiment has updated its latest value on sin22θ 13 and provided new results on 5 MeV excess, Δm2 ee, θ 13 with n-H analysis, absolute antineutrino flux measurement, and sterile neutrino search. It gives rich programs of neutrino properties, detector development, nuclear monitoring and application. Using reactor neutrinos, the future RENO-50 experiment will search for more precise measurement of θ 12, Δm 2 12 and mass hierarchy.

A Simple Tubular Reactor Experiment.

ERIC Educational Resources Information Center

Hudgins, Robert R.; Cayrol, Bertrand

1981-01-01

Using the hydrolysis of crystal violet dye by sodium hydroxide as an example, the theory, apparatus, and procedure for a laboratory demonstration of tubular reactor behavior are described. The reaction presented can occur at room temperature and features a color change to reinforce measured results. (WB)

Liquid fuel molten salt reactors for thorium utilization

DOE PAGES

Gehin, Jess C.; Powers, Jeffrey J.

2016-04-08

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

Temperature measuring analysis of the nuclear reactor fuel assembly

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

Urban, F., E-mail: jozef.bereznai@stuba.sk, E-mail: zdenko.zavodny@stuba.sk; Kučák, L., E-mail: jozef.bereznai@stuba.sk, E-mail: zdenko.zavodny@stuba.sk; Bereznai, J., E-mail: jozef.bereznai@stuba.sk, E-mail: zdenko.zavodny@stuba.sk

2014-08-06

Study was based on rapid changes of measured temperature values from the thermocouple in the VVER 440 nuclear reactor fuel assembly. Task was to determine origin of fluctuations of the temperature values by experiments on physical model of the fuel assembly. During an experiment, heated water was circulating in the system and cold water inlet through central tube to record sensitivity of the temperature sensor. Two positions of the sensor was used. First, just above the central tube in the physical model fuel assembly axis and second at the position of the thermocouple in the VVER 440 nuclear reactor fuelmore » assembly. Dependency of the temperature values on time are presented in the diagram form in the paper.« less

Evaluation of catalytic combustion of actual coal-derived gas

NASA Technical Reports Server (NTRS)

Blanton, J. C.; Shisler, R. A.

1982-01-01

The combustion characteristics of a Pt-Pl catalytic reactor burning coal-derived, low-Btu gas were investigated. A large matrix of test conditions was explored involving variations in fuel/air inlet temperature and velocity, reactor pressure, and combustor exit temperature. Other data recorded included fuel gas composition, reactor temperatures, and exhaust emissions. Operating experience with the reactor was satisfactory. Combustion efficiencies were quite high (over 95 percent) over most of the operating range. Emissions of NOx were quite high (up to 500 ppm V and greater), owing to the high ammonia content of the fuel gas.

Development of RF plasma simulations of in-reactor tests of small models of the nuclear light bulb fuel region

NASA Technical Reports Server (NTRS)

Roman, W. C.; Jaminet, J. F.

1972-01-01

Experiments were conducted to develop test configurations and technology necessary to simulate the thermal environment and fuel region expected to exist in in-reactor tests of small models of nuclear light bulb configurations. Particular emphasis was directed at rf plasma tests of approximately full-scale models of an in-reactor cell suitable for tests in Los Alamos Scientific Laboratory's Nuclear Furnace. The in-reactor tests will involve vortex-stabilized fissioning uranium plasmas of approximately 200-kW power, 500-atm pressure and equivalent black-body radiating temperatures between 3220 and 3510 K.

Heat exchanger with intermediate evaporating and condensing fluid

DOEpatents

Fraas, Arthur P.

1978-01-01

A shell and tube-type heat exchanger, such as a liquid sodium-operated steam generator for use in nuclear reactors, comprises a shell containing a primary fluid tube bundle, a secondary fluid tube bundle at higher elevation, and an intermediate fluid vaporizing at the surface of the primary fluid tubes and condensing at the surface of the secondary fluid tubes.

Calculation of natural convection test at Phenix using the NETFLOW++ code

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

Mochizuki, H.; Kikuchi, N.; Li, S.

2012-07-01

The present paper describes modeling and analyses of a natural convection of the pool-type fast breeder reactor Phenix. The natural convection test was carried out as one of the End of Life Tests of the Phenix. Objective of the present study is to assess the applicability of the NETFLOW++ code which has been verified thus far using various water facilities and validated using the plant data of the loop-type FBR 'Monju' and the loop-type experimental fast reactor 'Joyo'. The Phenix primary heat transport system is modeled based on the benchmark documents available from IAEA. The calculational model consists of onlymore » the primary heat transport system with boundary conditions on the secondary-side of IHX. The coolant temperature at the primary pump inlet, the primary coolant temperature at the IHX inlet and outlet, the secondary coolant temperatures and other parameters are calculated by the code where the heat transfer between the hot and cold pools is explicitly taken into account. A model including the secondary and tertiary systems was prepared, and the calculated results also agree well with the measured data in general. (authors)« less

A Methodology for the Integration of a Mechanistic Source Term Analysis in a Probabilistic Framework for Advanced Reactors

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

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

GE Hitachi Nuclear Energy (GEH) and Argonne National Laboratory are currently engaged in a joint effort to modernize and develop probabilistic risk assessment (PRA) techniques for advanced non-light water reactors. At a high level, the primary outcome of this project will be the development of next-generation PRA methodologies that will enable risk-informed prioritization of safety- and reliability-focused research and development, while also identifying gaps that may be resolved through additional research. A subset of this effort is the development of PRA methodologies to conduct a mechanistic source term (MST) analysis for event sequences that could result in the release ofmore » radionuclides. The MST analysis seeks to realistically model and assess the transport, retention, and release of radionuclides from the reactor to the environment. The MST methods developed during this project seek to satisfy the requirements of the Mechanistic Source Term element of the ASME/ANS Non-LWR PRA standard. The MST methodology consists of separate analysis approaches for risk-significant and non-risk significant event sequences that may result in the release of radionuclides from the reactor. For risk-significant event sequences, the methodology focuses on a detailed assessment, using mechanistic models, of radionuclide release from the fuel, transport through and release from the primary system, transport in the containment, and finally release to the environment. The analysis approach for non-risk significant event sequences examines the possibility of large radionuclide releases due to events such as re-criticality or the complete loss of radionuclide barriers. This paper provides details on the MST methodology, including the interface between the MST analysis and other elements of the PRA, and provides a simplified example MST calculation for a sodium fast reactor.« less

Demonstration of fully coupled simplified extended station black-out accident simulation with RELAP-7

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

Zhao, Haihua; Zhang, Hongbin; Zou, Ling

2014-10-01

The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at the Idaho National Laboratory (INL). The RELAP-7 code develop-ment effort started in October of 2011 and by the end of the second development year, a number of physical components with simplified two phase flow capability have been de-veloped to support the simplified boiling water reactor (BWR) extended station blackout (SBO) analyses. The demonstration case includes the major components for the primary system of a BWR, as well as the safety system components for the safety relief valve (SRV), the reactor core isolation cooling (RCIC)more » system, and the wet well. Three scenar-ios for the SBO simulations have been considered. Since RELAP-7 is not a severe acci-dent analysis code, the simulation stops when fuel clad temperature reaches damage point. Scenario I represents an extreme station blackout accident without any external cooling and cooling water injection. The system pressure is controlled by automatically releasing steam through SRVs. Scenario II includes the RCIC system but without SRV. The RCIC system is fully coupled with the reactor primary system and all the major components are dynamically simulated. The third scenario includes both the RCIC system and the SRV to provide a more realistic simulation. This paper will describe the major models and dis-cuss the results for the three scenarios. The RELAP-7 simulations for the three simplified SBO scenarios show the importance of dynamically simulating the SRVs, the RCIC sys-tem, and the wet well system to the reactor safety during extended SBO accidents.« less

PWR-related integral safety experiments in the PKL 111 test facility SBLOCA under beyond-design-basis accident conditions

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

Weber, P.; Umminger, K.J.; Schoen, B.

1995-09-01

The thermal hydraulic behavior of a PWR during beyond-design-basis accident scenarios is of vital interest for the verification and optimization of accident management procedures. Within the scope of the German reactor safety research program experiments were performed in the volumetrically scaled PKL 111 test facility by Siemens/KWU. This highly instrumented test rig simulates a KWU-design PWR (1300 MWe). In particular, the latest tests performed related to a SBLOCA with additional system failures, e.g. nitrogen entering the primary system. In the case of a SBLOCA, it is the goal of the operator to put the plant in a condition where themore » decay heat can be removed first using the low pressure emergency core cooling system and then the residual heat removal system. The experimental investigation presented assumed the following beyond-design-basis accident conditions: 0.5% break in a cold leg, 2 of 4 steam generators (SGs) isolated on the secondary side (feedwater- and steam line-valves closed), filled with steam on the primary side, cooldown of the primary system using the remaining two steam generators, high pressure injection system only in the two loops with intact steam generators, if possible no operator actions to reach the conditions for residual heat removal system activation. Furthermore, it was postulated that 2 of the 4 hot leg accumulators had a reduced initial water inventory (increased nitrogen inventory), allowing nitrogen to enter the primary system at a pressure of 15 bar and nearly preventing the heat transfer in the SGs ({open_quotes}passivating{close_quotes} U-tubes). Due to this the heat transfer regime in the intact steam generators changed remarkably. The primary system showed self-regulating system effects and heat transfer improved again (reflux-condenser mode in the U-tube inlet region).« less

The status of the positron beam facility at NEPOMUC

NASA Astrophysics Data System (ADS)

Hugenschmidt, C.

2011-01-01

The NEutron induced POsitron source MUniCh NEPOMUC provides a high intensity positron beam with 9·108 moderated positrons per second with a primary beam energy of 1keV. After remoderation, the positron beam is magnetically guided to five experimental setups: a coincident Doppler-broadening spectrometer (CDBS), a positron annihilation induced Auger-electron spectrometer (PAES), a pulsed low-energy positron system (PLEPS) as well as an interface for providing a pulsed beam with further improved brightness. An apparatus for the production of the negatively charged positronium ion Ps- is currently in operation at the open multi-purpose beam port, where additional experiments can be realized. Within this contribution, an overview of the positron beam facility NEPOMUC with its instrumentation at the research reactor FRMII is given.

Technical Information on the Carbonation of the EBR-II Reactor, Summary Report Part 2: Application to EBR-II Primary Sodium System and Related Systems

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

Steven R. Sherman; Collin J. Knight

Residual sodium is defined as sodium metal that remains behind in pipes, vessels, and tanks after the bulk sodium metal has been melted and drained from such components. The residual sodium has the same chemical properties as bulk sodium, and differs from bulk sodium only in the thickness of the sodium deposit. Typically, sodium is considered residual when the thickness of the deposit is less than 5-6 cm. This residual sodium must be removed or deactivated when a pipe, vessel, system, or entire reactor is permanently taken out of service, in order to make the component or system safer and/ormore » to comply with decontamination and decomissioning regulations. As an alternative to the established residual sodium deactivation techniques (steam-and-nitrogen, wet vapor nitrogen, etc.), a technique involving the use of moisture and carbon dioxide has been developed. With this technique, sodium metal is converted into sodium bicarbonate by reacting it with humid carbon dioxide. Hydrogen is emitted as a by-product. This technique was first developed in the laboratory by exposing sodium samples to humidifed carbon dioxide under controlled conditions, and then demonstrated on a larger scale by treating residual sodium within the Experimental Breeder Reactor II (EBR-II) secondary cooling system, followed by the primary cooling system, respectively. The EBR-II facility is located at the Idaho National Laboratory (INL) in southeastern Idaho, USA. This report is Part 2 of a two-part report. This second report provides a supplement to the first report and describes the application of the humdidified carbon dioxide technique ("carbonation") to the EBR-II primary tank, primary cover gas systems, and the intermediate heat exchanger. Future treatment plans are also provided.« less

Experiment on search for neutron-antineutron oscillations using a projected UCN source at the WWR-M reactor

NASA Astrophysics Data System (ADS)

Fomin, A. K.; Serebrov, A. P.; Zherebtsov, O. M.; Leonova, E. N.; Chaikovskii, M. E.

2017-01-01

We propose an experiment on search for neutron-antineutron oscillations based on the storage of ultracold neutrons (UCN) in a material trap. The sensitivity of the experiment mostly depends on the trap size and the amount of UCN in it. In Petersburg Nuclear Physics Institute (PNPI) a high-intensity UCN source is projected at the WWR-M reactor, which must provide UCN density 2-3 orders of magnitude higher than existing sources. The results of simulations of the designed experimental scheme show that the sensitivity can be increased by ˜ 10-40 times compared to sensitivity of previous experiment depending on the model of neutron reflection from walls.

Low-temperature catalytic gasification of food processing wastes. 1995 topical report

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

Elliott, D.C.; Hart, T.R.

The catalytic gasification system described in this report has undergone continuing development and refining work at Pacific Northwest National Laboratory (PNNL) for over 16 years. The original experiments, performed for the Gas Research Institute, were aimed at developing kinetics information for steam gasification of biomass in the presence of catalysts. From the fundamental research evolved the concept of a pressurized, catalytic gasification system for converting wet biomass feedstocks to fuel gas. Extensive batch reactor testing and limited continuous stirred-tank reactor tests provided useful design information for evaluating the preliminary economics of the process. This report is a follow-on to previousmore » interim reports which reviewed the results of the studies conducted with batch and continuous-feed reactor systems from 1989 to 1994, including much work with food processing wastes. The discussion here provides details of experiments on food processing waste feedstock materials, exclusively, that were conducted in batch and continuous- flow reactors.« less

Degradation of 2,4,6-trichlorophenol and determination of bacterial community structure by micro-electrical stimulation with or without external organic carbon source.

PubMed

Xu, Hao; Tong, Na; Huang, Shaobin; Zhou, Shaofeng; Li, Shuang; Li, Jianjun; Zhang, Yongqing

2018-05-03

This study aimed to investigate the degradation efficiency of 2,4,6-trichlorophenol through a batch of potentiostatic experiments (0.2 V vs. Ag/AgCl). Efficiencies in the presence and absence of acetate and glucose were compared through open-circuit reference experiments. Significant differences in degradation efficiency were observed in six reactors. The highest and lowest degradation efficiencies were observed in the closed-circuit reactor fed with glucose and in the open-circuit reactor, respectively. This finding was due to the enhanced bacterial metabolism caused by the application of micro-electrical field and degradable organics as co-substrates. The different treatment efficiencies were also caused by the distinct bacterial communities. The composition of bacterial community was affected by adding different organics as co-substrates. At the phylum level, the most dominant bacteria in the reactor with the added acetate and glucose were Proteobacteria and Firmicutes, respectively. Copyright © 2018 Elsevier Ltd. All rights reserved.

IN-PILE CORROSION TEST LOOPS FOR AQUEOUS HOMOGENEOUS REACTOR SOLUTIONS

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

Savage, H.C.; Jenks, G.H.; Bohlmann, E.G.

1960-12-21

An in-pile corrosion test loop is described which is used to study the effect of reactor radiation on the corrosion of materials of construction and the chemical stability of fuel solutions of interest to the Aqueous Homogeneous Reactor Program at ORNL. Aqueous solutions of uranyl sulfate are circulated in the loop by means of a 5-gpm canned-rotor pump, and the pump loop is designed for operation at temperatures to 300 ts C and pressures to 2000 psia while exposed to reactor radiation in beam-hole facilities of the LITR and ORR. Operation of the first loop in-pile was begun in Octobermore » 1954, and since that time 17 other in-pile loop experiments were completed. Design criteria of the pump loop and its associated auxiliary equipment and instrumentation are described. In-pile operating procedures, safety features, and operating experience are presented. A cost summary of the design, fabrication, and installation of the loop and experimental facillties is also included. (auth)« less

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.

Mass transfer study on the electrochemical removal of copper ions from synthetic effluents using reticulated vitreous carbon.

PubMed

Britto-Costa, Pedro H; Ruotolo, Luís Augusto M

2013-01-01

Porous electrodes have been successfully used for metal electrodeposition from diluted aqueous solution due to their high porosity and specific surface area, which lead to high mass transfer rates. This work studies the mass transfer of copper electrodeposition on reticulated vitreous carbon in a flow reactor without membrane. The flow configuration, otherwise the filter-press electrochemical reactors, was designed in order to minimize the pressure drop. The mass transfer coefficient was determined by voltammetric and galvanostatic electrodeposition. In the voltammetric experiments a Luggin capillary was used to measure the current-potential curves and to determine the limiting current (and, consequently, the mass transfer coefficient). In the galvanostatic experiments the concentration-time curves were obtained and considering a limiting current kinetics model, the mass transfer coefficient (k(m)) was determined for different flow velocities. The results showed that both methods give similar values of k(m), thus the voltammetric method can be recommended because it is faster and simpler. Finally, the reactor performance was compared with others from literature, and it was observed that the proposed reactor design has high Sherwood numbers similar to other reactor configurations using membranes and reticulated vitreous carbon electrodes.

Carbofuran removal in continuous-photocatalytic reactor: Reactor optimization, rate-constant determination and carbofuran degradation pathway analysis.

PubMed

Vishnuganth, M A; Remya, Neelancherry; Kumar, Mathava; Selvaraju, N

2017-05-04

Carbofuran (CBF) removal in a continuous-flow photocatalytic reactor with granular activated carbon supported titanium dioxide (GAC-TiO 2 ) catalyst was investigated. The effects of feed flow rate, TiO 2 concentration and addition of supplementary oxidants on CBF removal were investigated. The central composite design (CCD) was used to design the experiments and to estimate the effects of feed flow rate and TiO 2 concentration on CBF removal. The outcome of CCD experiments demonstrated that reactor performance was influenced mainly by feed flow rate compared to TiO 2 concentration. A second-order polynomial model developed based on CCD experiments fitted the experimental data with good correlation (R 2 ∼ 0.964). The addition of 1 mL min -1 hydrogen peroxide has shown complete CBF degradation and 76% chemical oxygen demand removal under the following operating conditions of CBF ∼50 mg L -1 , TiO 2 ∼5 mg L -1 and feed flow rate ∼82.5 mL min -1 . Rate constant of the photodegradation process was also calculated by applying the kinetic data in pseudo-first-order kinetics. Four major degradation intermediates of CBF were identified using GC-MS analysis. As a whole, the reactor system and GAC-TiO 2 catalyst used could be constructive in cost-effective CBF removal with no impact to receiving environment through getaway of photocatalyst.

Recent results of Daya Bay reactor neutrino experiment

NASA Astrophysics Data System (ADS)

Leitner, R.; Daya Bay Collaboration

2017-04-01

The Daya Bay reactor neutrino experiment has been designed to precisely measure the least known neutrino mixing angle θ13. In March 2012, Daya Bay collaboration announced [Daya Bay Collaboration (F. P. An et al.), Observation of electron-antineutrino disappearance at Daya Bay, Phys. Rev. Lett. 108 (2012) 171803] the observation of non-zero value of sin2 ⁡ 2θ13. Because of large statistics of detected antineutrinos and excellent performance of the experiment, Daya Bay continuously improves the precision of world best measurement of sin2 ⁡ 2θ13. In addition it provides results on neutrino mass splitting Δ mee2 competitive with measurements of other experiments, results on precise measurement of reactor fluxes and on limits of the existence of hypothetical fourth neutrino. In this paper, we report the results available by the time of the 6th Capri workshop: the measurement of oscillation parameters sin2 ⁡ (2θ13) = 0.084 ± 0.005 and | Δmee2 | = (2.42 ± 0.11) ×10-3eV2 [Daya Bay Collaboration (F. P. An et al.), New Measurement of Antineutrino Oscillation with the Full Detector Configuration at Daya Bay, Phys. Rev. Lett. 115 (2015) no. 11, 111802], searches for sterile neutrinos [Daya Bay Collaboration (F. P. An et al.) Search for a Light Sterile Neutrino at Daya Bay, Phys. Rev. Lett. 113 (2014) 141802] and precise measurement of reactor neutrino flux [Daya Bay Collaboration (F. P. An et al.), Measurement of the Reactor Anti-neutrino Flux and Spectrum at Daya Bay, Phys. Rev. Lett. 116 (2016) no. 6, 061801]. These are based on 621 days of measurement, most of the data has been taken in full detector configuration. More precise results [Daya Bay Collaboration (F. P. An et al.), Measurement of electron antineutrino oscillation based on 1230 days of operation of the Daya Bay experiment, arxiv:arXiv:1610.04802] with 1230 days of operation have been presented few weeks later at the Neutrino 2016 conference.

Biological CO2 conversion to acetate in subsurface coal-sand formation using a high-pressure reactor system.

PubMed

Ohtomo, Yoko; Ijiri, Akira; Ikegawa, Yojiro; Tsutsumi, Masazumi; Imachi, Hiroyuki; Uramoto, Go-Ichiro; Hoshino, Tatsuhiko; Morono, Yuki; Sakai, Sanae; Saito, Yumi; Tanikawa, Wataru; Hirose, Takehiro; Inagaki, Fumio

2013-01-01

Geological CO2 sequestration in unmineable subsurface oil/gas fields and coal formations has been proposed as a means of reducing anthropogenic greenhouse gasses in the atmosphere. However, the feasibility of injecting CO2 into subsurface depends upon a variety of geological and economic conditions, and the ecological consequences are largely unpredictable. In this study, we developed a new flow-through-type reactor system to examine potential geophysical, geochemical and microbiological impacts associated with CO2 injection by simulating in-situ pressure (0-100 MPa) and temperature (0-70°C) conditions. Using the reactor system, anaerobic artificial fluid and CO2 (flow rate: 0.002 and 0.00001 ml/min, respectively) were continuously supplemented into a column comprised of bituminous coal and sand under a pore pressure of 40 MPa (confined pressure: 41 MPa) at 40°C for 56 days. 16S rRNA gene analysis of the bacterial components showed distinct spatial separation of the predominant taxa in the coal and sand over the course of the experiment. Cultivation experiments using sub-sampled fluids revealed that some microbes survived, or were metabolically active, under CO2-rich conditions. However, no methanogens were activated during the experiment, even though hydrogenotrophic and methylotrophic methanogens were obtained from conventional batch-type cultivation at 20°C. During the reactor experiment, the acetate and methanol concentration in the fluids increased while the δ(13)Cacetate, H2 and CO2 concentrations decreased, indicating the occurrence of homo-acetogenesis. 16S rRNA genes of homo-acetogenic spore-forming bacteria related to the genus Sporomusa were consistently detected from the sandstone after the reactor experiment. Our results suggest that the injection of CO2 into a natural coal-sand formation preferentially stimulates homo-acetogenesis rather than methanogenesis, and that this process is accompanied by biogenic CO2 conversion to acetate.

Biological CO2 conversion to acetate in subsurface coal-sand formation using a high-pressure reactor system

PubMed Central

Ohtomo, Yoko; Ijiri, Akira; Ikegawa, Yojiro; Tsutsumi, Masazumi; Imachi, Hiroyuki; Uramoto, Go-Ichiro; Hoshino, Tatsuhiko; Morono, Yuki; Sakai, Sanae; Saito, Yumi; Tanikawa, Wataru; Hirose, Takehiro; Inagaki, Fumio

2013-01-01

Geological CO2 sequestration in unmineable subsurface oil/gas fields and coal formations has been proposed as a means of reducing anthropogenic greenhouse gasses in the atmosphere. However, the feasibility of injecting CO2 into subsurface depends upon a variety of geological and economic conditions, and the ecological consequences are largely unpredictable. In this study, we developed a new flow-through-type reactor system to examine potential geophysical, geochemical and microbiological impacts associated with CO2 injection by simulating in-situ pressure (0–100 MPa) and temperature (0–70°C) conditions. Using the reactor system, anaerobic artificial fluid and CO2 (flow rate: 0.002 and 0.00001 ml/min, respectively) were continuously supplemented into a column comprised of bituminous coal and sand under a pore pressure of 40 MPa (confined pressure: 41 MPa) at 40°C for 56 days. 16S rRNA gene analysis of the bacterial components showed distinct spatial separation of the predominant taxa in the coal and sand over the course of the experiment. Cultivation experiments using sub-sampled fluids revealed that some microbes survived, or were metabolically active, under CO2-rich conditions. However, no methanogens were activated during the experiment, even though hydrogenotrophic and methylotrophic methanogens were obtained from conventional batch-type cultivation at 20°C. During the reactor experiment, the acetate and methanol concentration in the fluids increased while the δ13Cacetate, H2 and CO2 concentrations decreased, indicating the occurrence of homo-acetogenesis. 16S rRNA genes of homo-acetogenic spore-forming bacteria related to the genus Sporomusa were consistently detected from the sandstone after the reactor experiment. Our results suggest that the injection of CO2 into a natural coal-sand formation preferentially stimulates homo-acetogenesis rather than methanogenesis, and that this process is accompanied by biogenic CO2 conversion to acetate. PMID:24348470

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.

Initiating Event Analysis of a Lithium Fluoride Thorium Reactor

NASA Astrophysics Data System (ADS)

Geraci, Nicholas Charles

The primary purpose of this study is to perform an Initiating Event Analysis for a Lithium Fluoride Thorium Reactor (LFTR) as the first step of a Probabilistic Safety Assessment (PSA). The major objective of the research is to compile a list of key initiating events capable of resulting in failure of safety systems and release of radioactive material from the LFTR. Due to the complex interactions between engineering design, component reliability and human reliability, probabilistic safety assessments are most useful when the scope is limited to a single reactor plant. Thus, this thesis will study the LFTR design proposed by Flibe Energy. An October 2015 Electric Power Research Institute report on the Flibe Energy LFTR asked "what-if?" questions of subject matter experts and compiled a list of key hazards with the most significant consequences to the safety or integrity of the LFTR. The potential exists for unforeseen hazards to pose additional risk for the LFTR, but the scope of this thesis is limited to evaluation of those key hazards already identified by Flibe Energy. These key hazards are the starting point for the Initiating Event Analysis performed in this thesis. Engineering evaluation and technical study of the plant using a literature review and comparison to reference technology revealed four hazards with high potential to cause reactor core damage. To determine the initiating events resulting in realization of these four hazards, reference was made to previous PSAs and existing NRC and EPRI initiating event lists. Finally, fault tree and event tree analyses were conducted, completing the logical classification of initiating events. Results are qualitative as opposed to quantitative due to the early stages of system design descriptions and lack of operating experience or data for the LFTR. In summary, this thesis analyzes initiating events using previous research and inductive and deductive reasoning through traditional risk management techniques to arrive at a list of key initiating events that can be used to address vulnerabilities during the design phases of LFTR development.

Bioreactor design studies for a hydrogen-producing bacterium.

PubMed

Wolfrum, Edward J; Watt, Andrew S

2002-01-01

Carbon monoxide (CO) can be metabolized by a number of microorganisms along with water to produce hydrogen (H2) and carbon dioxide. National Renewable Energy Laboratory researchers have isolated a number of bacteria that perform this so-called water-gas shift reaction at ambient temperatures. We performed experiments to measure the rate of CO conversion and H2 production in a trickle-bed reactor (TBR). The liquid recirculation rate and the reactor support material both affected the mass transfer coefficient, which controls the overall performance of the reactor. A simple reactor model taken from the literature was used to quantitatively compare the performance of the TBR geometry at two different size scales. Good agreement between the two reactor scales was obtained.

Bioconversion of waste office paper to hydrogen using pretreated rumen fluid inoculum.

PubMed

Botta, Lívia Silva; Ratti, Regiane Priscila; Sakamoto, Isabel Kimiko; Ramos, Lucas Rodrigues; Silva, Edson Luiz; Varesche, Maria Bernadete Amâncio

2016-12-01

In this study, a microbial consortium from an acid-treated rumen fluid was used to improve the yields of H 2 production from paper residues in batch reactors. The anaerobic batch reactors, which contained paper and cellulose, were operated under three conditions: (1) 0.5 g paper/L, (2) 2 g paper/L, and (3) 4 g paper/L. Cellulase was added to promote the hydrolysis of paper to soluble sugars. The H 2 yields were 5.51, 4.65, and 3.96 mmol H 2 /g COD, respectively, with substrate degradation ranging from 56 to 65.4 %. Butyric acid was the primary soluble metabolite in the three reactors, but pronounced solventogenesis was detected in the reactors incubated with increased paper concentrations (2.0 and 4.0 g/L). A substantial prevalence of Clostridium acetobutylicum (99 % similarity) was observed in the acid-treated rumen fluid, which has been recognized as an efficient H 2 -producing strain in addition to ethanol and n-butanol which were also detected in the reactors.

A comparison of neutron and gamma damage effects on silica glass in a nuclear reactor radiation environment

NASA Astrophysics Data System (ADS)

Holcomb, David E.; Miller, Don W.

1993-08-01

A study of the relative damage effects of neutrons and gamma rays on silica glass in a nuclear reactor radiation environment is reported. The neutron and gamma energy spectra of the Ohio State University Research Reactor beam port #1 were applied to silica glass to obtain primary knock-on charged particle energy spectra. The resultant charged particle spectra were then applied to the polyatomic forms of the Lindhard et al. integrodifferential equation for damage energy and the Parkin and Coulter integrodifferential equation for net atomic displacement. The results show that near a nuclear reactor core the vast majority of the dose to silica is due to gamma rays (factor of roughly 40) and that neutrons cause much more displacement damage than gamma rays (35 times the oxygen displacement rate and 500 times the silicon displacement rate). However, pure silica core optical fibers irradiated in a nuclear reactor's mixed neutron/gamma environment exhibit little difference in transmission loss on an equal dose basis compared to fibers irradiated in a gamma only environment, indicating that atomic displacement is not a significant damage mechanism.

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

Gerczak, Tyler J.; Smith, Kurt R.; Petrie, Christian M.

Tristructural-isotropic (TRISO)–coated particle fuel is a promising advanced fuel concept consisting of a spherical fuel kernel made of uranium oxide and uranium carbide, surrounded by a porous carbonaceous buffer layer and successive layers of dense inner pyrolytic carbon (IPyC), silicon carbide (SiC) deposited by chemical vapor , and dense outer pyrolytic carbon (OPyC). This fuel concept is being considered for advanced reactor applications such as high temperature gas-cooled reactors (HTGRs) and molten salt reactors (MSRs), as well as for accident-tolerant fuel for light water reactors (LWRs). Development and implementation of TRISO fuel for these reactor concepts support the US Departmentmore » of Energy (DOE) Office of Nuclear Energy mission to promote safe, reliable nuclear energy that is sustainable and environmentally friendly. During operation, the SiC layer serves as the primary barrier to metallic fission products and actinides not retained in the kernel. It has been observed that certain fission products are released from TRISO fuel during operation, notably, Ag, Eu, and Sr [1]. Release of these radioisotopes causes safety and maintenance concerns.« less

Hydrodynamic effects on cells in agitated tissue culture reactors

NASA Technical Reports Server (NTRS)

Cherry, R. S.; Papoutsakis, E. T.

1986-01-01

The mechanisms by which hydrodynamic forces can affect cells grown on microcarrier beads in agitated cell culture reactors were investigated by analyzing the motion of microcarriers relative to the surrounding fluid, to each other, and to moving or stationary solid surfaces. It was found that harmful effects on cell cultures that have been previously attributed to shear can be better explained as the effects of turbulence (of a size scale comparable to the microcarriers or the spacing between them) or collisions. The primary mechanisms of cell damage involve direct interaction between microcarriers and turbulent eddies, collisions between microcarriers in turbulent flow, and collisions against the impeller or other solid surfaces. The implications of these analytical results for the design of tissue culture reactors are discussed.

Isotopic evidence for nitrous oxide production pathways in a partial nitritation-anammox reactor.

PubMed

Harris, Eliza; Joss, Adriano; Emmenegger, Lukas; Kipf, Marco; Wolf, Benjamin; Mohn, Joachim; Wunderlin, Pascal

2015-10-15

Nitrous oxide (N2O) production pathways in a single stage, continuously fed partial nitritation-anammox reactor were investigated using online isotopic analysis of offgas N2O with quantum cascade laser absorption spectroscopy (QCLAS). N2O emissions increased when reactor operating conditions were not optimal, for example, high dissolved oxygen concentration. SP measurements indicated that the increase in N2O was due to enhanced nitrifier denitrification, generally related to nitrite build-up in the reactor. The results of this study confirm that process control via online N2O monitoring is an ideal method to detect imbalances in reactor operation and regulate aeration, to ensure optimal reactor conditions and minimise N2O emissions. Under normal operating conditions, the N2O isotopic site preference (SP) was much higher than expected - up to 40‰ - which could not be explained within the current understanding of N2O production pathways. Various targeted experiments were conducted to investigate the characteristics of N2O formation in the reactor. The high SP measurements during both normal operating and experimental conditions could potentially be explained by a number of hypotheses: i) unexpectedly strong heterotrophic N2O reduction, ii) unknown inorganic or anammox-associated N2O production pathway, iii) previous underestimation of SP fractionation during N2O production from NH2OH, or strong variations in SP from this pathway depending on reactor conditions. The second hypothesis - an unknown or incompletely characterised production pathway - was most consistent with results, however the other possibilities cannot be discounted. Further experiments are needed to distinguish between these hypotheses and fully resolve N2O production pathways in PN-anammox systems. Copyright © 2015 Elsevier Ltd. All rights reserved.

Pressurized-water reactor internals aging degradation study. Phase 1

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

Luk, K.H.

1993-09-01

This report documents the results of a Phase I study on the effects of aging degradations on pr internals. Primary stressers for internals an generated by the primary coolant flow in the they include unsteady hydrodynamic forces and pump-generated pressure pulsations. Other stressors are applied loads, manufacturing processes, impurities in the coolant and exposures to fast neutron fluxes. A survey of reported aging-related failure information indicates that fatigue, stress corrosion cracking (SCC) and mechanical wear are the three major aging-related degradation mechanisms for PWR internals. Significant reported failures include thermal shield flow-induced vibration problems, SCC in guide tube support pinsmore » and core support structure bolts, fatigue-induced core baffle water-jet impingement problems and excess wear in flux thimbles. Many of the reported problems have been resolved by accepted engineering practices. Uncertainties remain in the assessment of long-term neutron irradiation effects and environmental factors in high-cycle fatigue failures. Reactor internals are examined by visual inspections and the technique is access limited. Improved inspection methods, especially one with an early failure detection capability, can enhance the safety and efficiency of reactor operations.« less