Northrop Triga facility decommissioning plan versus actual results

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

Gardner, F.W.

1986-01-01

This paper compares the Triga facility decontamination and decommissioning plan to the actual results and discusses key areas where operational activities were impacted upon by the final US Nuclear Regulatory Commission (NRC)-approved decontamination and decommissioning plan. Total exposures for fuel transfer were a factor of 4 less than planned. The design of the Triga reactor components allowed the majority of the components to be unconditionally released.

Experimental power density distribution benchmark in the TRIGA Mark II reactor

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

Snoj, L.; Stancar, Z.; Radulovic, V.

2012-07-01

In order to improve the power calibration process and to benchmark the existing computational model of the TRIGA Mark II reactor at the Josef Stefan Inst. (JSI), a bilateral project was started as part of the agreement between the French Commissariat a l'energie atomique et aux energies alternatives (CEA) and the Ministry of higher education, science and technology of Slovenia. One of the objectives of the project was to analyze and improve the power calibration process of the JSI TRIGA reactor (procedural improvement and uncertainty reduction) by using absolutely calibrated CEA fission chambers (FCs). This is one of the fewmore » available power density distribution benchmarks for testing not only the fission rate distribution but also the absolute values of the fission rates. Our preliminary calculations indicate that the total experimental uncertainty of the measured reaction rate is sufficiently low that the experiments could be considered as benchmark experiments. (authors)« less

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.

Characterization of gamma field in the JSI TRIGA reactor

NASA Astrophysics Data System (ADS)

Ambrožič, Klemen; Radulović, Vladimir; Snoj, Luka; Gruel, Adrien; Guillou, Mael Le; Blaise, Patrick; Destouches, Christophe; Barbot, Loïc

2018-01-01

Research reactors such as the "Jožzef Stefan" Institute TRIGA reactor have primarily been designed for experimentation and sample irradiation with neutrons. However recent developments in incorporating additional instrumentation for nuclear power plant support and with novel high flux material testing reactor designs, γ field characterization has become of great interest for the characterization of the changes in operational parameters of electronic devices and for the evaluation of γ heating of MTR's structural materials in a representative reactor Γ spectrum. In this paper, we present ongoing work on γ field characterization both experimentally, by performing γ field measurements, and by simulations, using Monte Carlo particle transport codes in conjunction with R2S methodology for delayed γ field characterization.

Monte Carlo modelling of TRIGA research reactor

NASA Astrophysics Data System (ADS)

El Bakkari, B.; Nacir, B.; El Bardouni, T.; El Younoussi, C.; Merroun, O.; Htet, A.; Boulaich, Y.; Zoubair, M.; Boukhal, H.; Chakir, M.

2010-10-01

The Moroccan 2 MW TRIGA MARK II research reactor at Centre des Etudes Nucléaires de la Maâmora (CENM) achieved initial criticality on May 2, 2007. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes for their use in agriculture, industry, and medicine. This study deals with the neutronic analysis of the 2-MW TRIGA MARK II research reactor at CENM and validation of the results by comparisons with the experimental, operational, and available final safety analysis report (FSAR) values. The study was prepared in collaboration between the Laboratory of Radiation and Nuclear Systems (ERSN-LMR) from Faculty of Sciences of Tetuan (Morocco) and CENM. The 3-D continuous energy Monte Carlo code MCNP (version 5) was used to develop a versatile and accurate full model of the TRIGA core. The model represents in detailed all components of the core with literally no physical approximation. Continuous energy cross-section data from the more recent nuclear data evaluations (ENDF/B-VI.8, ENDF/B-VII.0, JEFF-3.1, and JENDL-3.3) as well as S( α, β) thermal neutron scattering functions distributed with the MCNP code were used. The cross-section libraries were generated by using the NJOY99 system updated to its more recent patch file "up259". The consistency and accuracy of both the Monte Carlo simulation and neutron transport physics were established by benchmarking the TRIGA experiments. Core excess reactivity, total and integral control rods worth as well as power peaking factors were used in the validation process. Results of calculations are analysed and discussed.

The SANS facility at the Pitesti 14MW TRIGA reactor

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

Ionita, I.; Grabcev, B.; Todireanu, S.

2006-12-15

The SANS facility existing at the Pitesti 14MW TRIGA reactor is presented. The main characteristics and the preliminary evaluation of the installation performances are given. A monochromatic neutron beam with 1.5 A {<=} {lambda} {<=} 5 A is produced by a mechanical velocity selector with helical slots. A fruitful partnership was established between INR Pitesti (Romania) and JINR Dubna (Russia). The first step in this cooperation consists in the manufacturing in Dubna of a battery of gas-filled positional detectors devoted to the SANS instrument.

Development of the ageing management database of PUSPATI TRIGA reactor

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

Ramli, Nurhayati, E-mail: nurhayati@nm.gov.my; Tom, Phongsakorn Prak; Husain, Nurfazila

Since its first criticality in 1982, PUSPATI TRIGA Reactor (RTP) has been operated for more than 30 years. As RTP become older, ageing problems have been seen to be the prominent issues. In addressing the ageing issues, an Ageing Management (AgeM) database for managing related ageing matters was systematically developed. This paper presents the development of AgeM database taking into account all RTP major Systems, Structures and Components (SSCs) and ageing mechanism of these SSCs through the system surveillance program.

Vibro-acoustic Imaging at the Breazeale Reactor

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

Smith, James Arthur; Jewell, James Keith; Lee, James Edwin

2016-09-01

The INL is developing Vibro-acoustic imaging technology to characterize microstructure in fuels and materials in spent fuel pools and within reactor vessels. A vibro-acoustic development laboratory has been established at the INL. The progress in developing the vibro-acoustic technology at the INL is the focus of this report. A successful technology demonstration was performed in a working TRIGA research reactor. Vibro-acoustic imaging was performed in the reactor pool of the Breazeale reactor in late September of 2015. A confocal transducer driven at a nominal 3 MHz was used to collect the 60 kHz differential beat frequency induced in a spentmore » TRIGA fuel rod and empty gamma tube located in the main reactor water pool. Data was collected and analyzed with the INLDAS data acquisition software using a short time Fourier transform.« less

Research reactor decommissioning experience - concrete removal and disposal -

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

Manning, Mark R.; Gardner, Frederick W.

1990-07-01

Removal and disposal of neutron activated concrete from biological shields is the most significant operational task associated with research reactor decommissioning. During the period of 1985 thru 1989 Chem-Nuclear Systems, Inc. was the prime contractor for complete dismantlement and decommissioning of the Northrop TRIGA Mark F, the Virginia Tech Argonaut, and the Michigan State University TRIGA Mark I Reactor Facilities. This paper discusses operational requirements, methods employed, and results of the concrete removal, packaging, transport and disposal operations for these (3) research reactor decommissioning projects. Methods employed for each are compared. Disposal of concrete above and below regulatory release limitsmore » for unrestricted use are discussed. This study concludes that activated reactor biological shield concrete can be safely removed and buried under current regulations.« less

Validation of absolute axial neutron flux distribution calculations with MCNP with 197Au(n,γ)198Au reaction rate distribution measurements at the JSI TRIGA Mark II reactor.

PubMed

Radulović, Vladimir; Štancar, Žiga; Snoj, Luka; Trkov, Andrej

2014-02-01

The calculation of axial neutron flux distributions with the MCNP code at the JSI TRIGA Mark II reactor has been validated with experimental measurements of the (197)Au(n,γ)(198)Au reaction rate. The calculated absolute reaction rate values, scaled according to the reactor power and corrected for the flux redistribution effect, are in good agreement with the experimental results. The effect of different cross-section libraries on the calculations has been investigated and shown to be minor. Copyright © 2013 Elsevier Ltd. All rights reserved.

Implementation of k0-INAA standardisation at ITU TRIGA Mark II research reactor, Turkey based on k0-IAEA software

NASA Astrophysics Data System (ADS)

Esen, Ayse Nur; Haciyakupoglu, Sevilay

2016-02-01

The purpose of this study is to test the applicability of k0-INAA method at the Istanbul Technical University TRIGA Mark II research reactor. The neutron spectrum parameters such as epithermal neutron flux distribution parameter (α), thermal to epithermal neutron flux ratio (f) and thermal neutron flux (φth) were determined at the central irradiation channel of the ITU TRIGA Mark II research reactor using bare triple-monitor method. HPGe detector calibrations and calculations were carried out by k0-IAEA software. The α, f and φth values were calculated to be -0.009, 15.4 and 7.92·1012 cm-2 s-1, respectively. NIST SRM 1633b coal fly ash and intercomparison samples consisting of clay and sandy soil samples were used to evaluate the validity of the method. For selected elements, the statistical evaluation of the analysis results was carried out by z-score test. A good agreement between certified/reported and experimental values was obtained.

An investigation into the feasibility of thorium fuels utilization in seed-blanket configurations for TRIGA PUSPATI Reactor (RTP)

NASA Astrophysics Data System (ADS)

Damahuri, Abdul Hannan Bin; Mohamed, Hassan; Aziz Mohamed, Abdul; Idris, Faridah

2018-01-01

Thorium is one of the elements that needs to be explored for nuclear fuel research and development. One of the popular core configurations of thorium fuel is seed-blanket configuration or also known as Radkowsky Thorium Fuel concept. The seed will act as a supplier of neutrons, which will be placed inside of the core. The blanket, on the other hand, is the consumer of neutrons that is located at outermost of the core. In this work, a neutronic analysis of seed-blanket configuration for the TRIGA PUSPATI Reactor (RTP) is carried out using Monte Carlo method. The reactor, which has been operated since 1982 use uranium zirconium hydride (U-ZrH1.6) as the fuel and have multiple uranium weight which are 8.5, 12 and 20 wt.%. The pool type reactor is one and only research reactor that located in Malaysia. The design of core included the Uranium Zirconium Hydride located at the centre of the core that will act as the seed to supply neutron. The thorium oxide that will act as blanket situated outside of seed region will receive neutron to transmute 232Th to 233U. The neutron multiplication factor or criticality of each configuration is estimated. Results show that the highest initial criticality achieved is 1.30153.

Analysis of JSI TRIGA MARK II reactor physical parameters calculated with TRIPOLI and MCNP.

PubMed

Henry, R; Tiselj, I; Snoj, L

2015-03-01

New computational model of the JSI TRIGA Mark II research reactor was built for TRIPOLI computer code and compared with existing MCNP code model. The same modelling assumptions were used in order to check the differences of the mathematical models of both Monte Carlo codes. Differences between the TRIPOLI and MCNP predictions of keff were up to 100pcm. Further validation was performed with analyses of the normalized reaction rates and computations of kinetic parameters for various core configurations. Copyright © 2014 Elsevier Ltd. All rights reserved.

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

NASA Astrophysics Data System (ADS)

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

2018-01-01

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

Basic requirements and parameter optimization for boron neutron capture therapy of extracorporeal irradiated and auto-transplanted organs.

PubMed

Wortmann, Birgit; Knorr, Jürgen

2012-08-01

In 2001 and 2003, at the University of Pavia, Italy, boron neutron capture therapy (BNCT) has been successfully used in the treatment of hepatic colorectal metastases (Pinelli et al., 2002; Zonta et al., 2006). The treatment procedure (TAOrMINA protocol) is characterised by the auto-transplantation and extracorporeal irradiation of the liver using a thermal neutron beam. The clinical use of this approach requires well founded data and an optimized irradiation facility. In order to start with this work and to decide upon its feasibility at the research reactor TRIGA Mainz, basic data and requirements have been considered (Wortmann, 2008). Computer calculations using the ATTILA (Transpire Inc. 2006) and MCNP (LANL, 2005) codes have been performed, including data from conventional radiation therapy, from the TAOrMINA approach, resulting in reasonable estimations. Basic data and requirements and optimal parameters have been worked out, especially for use at an optimized TRIGA irradiation facility (Wortmann, 2008). Advantages of the extracorporeal irradiation with auto-transplantation and the potential of an optimized irradiation facility could be identified. Within the requirements, turning the explanted organ over by 180° appears preferable to a whole side source, similar to a permanent rotation of the organ. The design study and the parameter optimization confirm the potential of this approach to treat metastases in explanted organs. The results do not represent actual treatment data but a first estimation. Although all specific values refer to the TRIGA Mainz, they may act as a useful guide for other types of neutron sources. The recommended modifications (Wortmann, 2008) show the suitability of TRIGA reactors as a radiation source for BNCT of extracorporeal irradiated and auto-transplanted organs. Copyright © 2012 Elsevier Ltd. All rights reserved.

Conceptual Design of a Clinical BNCT Beam in an Adjacent Dry Cell of the Jozef Stefan Institute TRIGA Reactor

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

Maucec, Marko

2000-11-15

The MCNP4B Monte Carlo transport code is used in a feasibility study of the epithermal neutron boron neutron capture therapy facility in the thermalizing column of the 250-kW TRIGA Mark II reactor at the Jozef Stefan Institute (JSI). To boost the epithermal neutron flux at the reference irradiation point, the efficiency of a fission plate with almost 1.5 kg of 20% enriched uranium and 2.3 kW of thermal power is investigated. With the same purpose in mind, the TRIGA reactor core setup is optimized, and standard fresh fuel elements are concentrated partly in the outermost ring of the core. Further,more » a detailed parametric study of the materials and dimensions for all the relevant parts of the irradiation facility is carried out. Some of the standard epithermal neutron filter/moderator materials, as well as 'pressed-only' low-density Al{sub 2}O{sub 3} and AlF{sub 3}, are considered. The proposed version of the BNCT facility, with PbF{sub 2} as the epithermal neutron filter/moderator, provides an epithermal neutron flux of {approx}1.1 x 10{sup 9} n/cm{sup 2}.s, thus enabling patient irradiation times of <60 min. With reasonably low fast neutron and photon contamination ([overdot]D{sub nfast}/{phi}{sub epi} < 5 x 10{sup -13} Gy.cm{sup 2}/n and [overdot]D{sub {gamma}} /{phi}{sub epi} < 3 x 10{sup -13} Gy.cm{sup 2}/n), the in-air performances of the proposed beam are comparable to all existing epithermal BNCT facilities. The design presents an equally efficient alternative to the BNCT beams in TRIGA reactor thermal columns that are more commonly applied. The cavity of the dry cell, a former JSI TRIGA reactor spent-fuel storage facility, adjacent to the thermalizing column, could rather easily be rearranged into a suitable patient treatment room, which would substantially decrease the overall developmental costs.« less

Technical Application of Nuclear Fission

NASA Astrophysics Data System (ADS)

Denschlag, J. O.

The chapter is devoted to the practical application of the fission process, mainly in nuclear reactors. After a historical discussion covering the natural reactors at Oklo and the first attempts to build artificial reactors, the fundamental principles of chain reactions are discussed. In this context chain reactions with fast and thermal neutrons are covered as well as the process of neutron moderation. Criticality concepts (fission factor η, criticality factor k) are discussed as well as reactor kinetics and the role of delayed neutrons. Examples of specific nuclear reactor types are presented briefly: research reactors (TRIGA and ILL High Flux Reactor), and some reactor types used to drive nuclear power stations (pressurized water reactor [PWR], boiling water reactor [BWR], Reaktor Bolshoi Moshchnosti Kanalny [RBMK], fast breeder reactor [FBR]). The new concept of the accelerator-driven systems (ADS) is presented. The principle of fission weapons is outlined. Finally, the nuclear fuel cycle is briefly covered from mining, chemical isolation of the fuel and preparation of the fuel elements to reprocessing the spent fuel and conditioning for deposit in a final repository.

Fabrication and testing of a 4-node micro-pocket fission detector array for the Kansas State University TRIGA Mk. II research nuclear reactor

NASA Astrophysics Data System (ADS)

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

2017-08-01

Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Micro-Pocket Fission Detectors (MPFDs) have been fabricated and tested previously, but successful testing of these prior detectors was limited to single-node operation with specialized designs. Described in this work is a modular, four-node MPFD array fabricated and tested at Kansas State University (KSU). The four sensor nodes were equally spaced to span the length of the fuel-region of the KSU TRIGA Mk. II research nuclear reactor core. The encapsulated array was filled with argon gas, serving as an ionization medium in the small cavities of the MPFDs. The unified design improved device ruggedness and simplified construction over previous designs. A 0.315-in. (8-mm) penetration in the upper grid plate of the KSU TRIGA Mk. II research nuclear reactor was used to deploy the array between fuel elements in the core. The MPFD array was coupled to an electronic support system which has been developed to support pulse-mode operation. Neutron-induced pulses were observed on all four sensor channels. Stable device operation was confirmed by testing under steady-state reactor conditions. Each of the four sensors in the array responded to changes in reactor power between 10 kWth and full power (750 kWth). Reactor power transients were observed in real-time including positive transients with periods of 5, 15, and 30 s. Finally, manual reactor power oscillations were observed in real-time.

Qualification of heavy water based irradiation device in the JSI TRIGA reactor for irradiations of FT-TIMS samples for nuclear safeguards

NASA Astrophysics Data System (ADS)

Radulović, Vladimir; Kolšek, Aljaž; Fauré, Anne-Laure; Pottin, Anne-Claire; Pointurier, Fabien; Snoj, Luka

2018-03-01

The Fission Track Thermal Ionization Mass Spectrometry (FT-TIMS) method is considered as the reference method for particle analysis in the field of nuclear Safeguards for measurements of isotopic compositions (fissile material enrichment levels) in micrometer-sized uranium particles collected in nuclear facilities. An integral phase in the method is the irradiation of samples in a very well thermalized neutron spectrum. A bilateral collaboration project was carried out between the Jožef Stefan Institute (JSI, Slovenia) and the Commissariat à l'Énergie Atomique et aux Énergies Alternatives (CEA, France) to determine whether the JSI TRIGA reactor could be used for irradiations of samples for the FT-TIMS method. This paper describes Monte Carlo simulations, experimental activation measurements and test irradiations performed in the JSI TRIGA reactor, firstly to determine the feasibility, and secondly to design and qualify a purpose-built heavy water based irradiation device for FT-TIMS samples. The final device design has been shown experimentally to meet all the required performance specifications.

NATCRCTR: One-dimensional thermal-hydraulics analysis code for natural-circulation TRIGA reactors

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

Feltus, M.A.; Rubinaccio, G.

1996-12-31

The Pennsylvania State University nuclear engineering department is evaluating the upgrade of the Reed College (Portland, Oregon) TRIGA reactor from 250 kW to 1 MW in two areas: thermal-hydraulics and steady-state neutronics analysis. This analysis was initiated as a cooperative effort between Penn State and Reed College as a training project for two International Atomic Energy Agency (IAEA) fellows from Ghana. The two Ghanaian IAEA fellows were assisted by G. Rubinaccio, an undergraduate, who undertook the task of writing the new computer programs for the thermal-hydraulic and physics evaluation as a three-credit special design project course. The Reed College TRIGA,more » which has a fixed graphite radial reflector, is cooled by natural circulation, without external cross-flow; whereas, the Penn State Breazeale Reactor has significant crossflow into its sides. To model the Reed TRIGA, the NATCRCTR program has been developed from first principles using the following assumptions: 1. The core is surrounded by the fixed reflector structure, which acts as a one-dimensional channel. 2. The core inlet temperature distribution is constant at the core bottom. 3. The axial heat flux distribution is a chopped cosine shape. 4. The heat transfer in the fuel is primarily in the radial directions. 5. A small gap between the fuel and cladding exists. The NATCRCTR code is used to find the peak centerline fuel, gap, and cladding surface temperatures, based on assumed flux and engineering peaking factors.« less

Transition from HEU to LEU fuel in Romania`s 14-MW TRIGA reactor

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

Bretscher, M.M.; Snelgrove, J.L.

1991-12-31

The 14-MW TRIGA steady state reactor (SSR) located in Pitesti, Romania, first went critical in the fall of 1979. Initially, the core configuration for full power operation used 29 fuel clusters each containing a 5 {times} 5 square array of HEU (10 wt%) -- ZrH -- Er (2.8 wt%) fuel-moderator rods (1.295 cm o.d.) clad in Incology. With a total inventory of 35 HEU fuel clusters, burnup considerations required a gradual expansion of the core from 29 to 32 and finally to 35 clusters before the reactor was shut down because of insufficient excess reactivity. At this time each ofmore » the original 29 fuel clusters had an overage {sup 235}U burnup in the range from 50 to 62%. Because of the US policy regarding the export of highly enriched uranium, fresh HEU TRIGA replacement fuel is not available. After a number of safety-related measurements, the SSR is expected to resume full power operation in the near future using a mixed core containing five LEU TRIGA clusters of the same geometry as the original fuel but with fuel-moderator rods containing 45 wt% U (19.7% {sup 235}U enrichment) and 1.1 wt% Er. Rods for 14 additional LEU fuel clusters will be fabricated by General Atomics. In support of the SSR mixed core operation numerous neutronic calculations have been performed. This paper presents some of the results of those calculations.« less

Assessment and mitigation of power quality problems for PUSPATI TRIGA Reactor (RTP)

NASA Astrophysics Data System (ADS)

Zakaria, Mohd Fazli; Ramachandaramurthy, Vigna K.

2017-01-01

An electrical power systems are exposed to different types of power quality disturbances. Investigation and monitoring of power quality are necessary to maintain accurate operation of sensitive equipment especially for nuclear installations. This paper will discuss the power quality problems observed at the electrical sources of PUSPATI TRIGA Reactor (RTP). Assessment of power quality requires the identification of any anomalous behavior on a power system, which adversely affects the normal operation of electrical or electronic equipment. A power quality assessment involves gathering data resources; analyzing the data (with reference to power quality standards) then, if problems exist, recommendation of mitigation techniques must be considered. Field power quality data is collected by power quality recorder and analyzed with reference to power quality standards. Normally the electrical power is supplied to the RTP via two sources in order to keep a good reliability where each of them is designed to carry the full load. The assessment of power quality during reactor operation was performed for both electrical sources. There were several disturbances such as voltage harmonics and flicker that exceeded the thresholds. To reduce these disturbances, mitigation techniques have been proposed, such as to install passive harmonic filters to reduce harmonic distortion, dynamic voltage restorer (DVR) to reduce voltage disturbances and isolate all sensitive and critical loads.

Characterization of fast neutron spectrum in the TRIGA for hardness testing of electronic components

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

Nelson, George W.

1986-07-01

Argonne National Laboratory-West, operated by the University of Chicago, is located near Idaho Falls, ID, on the Idaho National Engineering Laboratory Site. ANL-West performs work in support of the Liquid Metal Fast Breeder Reactor Program (LMFBR) sponsored by the United States Department of Energy. The NRAD reactor is located at the Argonne Site within the Hot Fuel Examination Facility/North, a large hot cell facility where both non-destructive and destructive examinations are performed on highly irradiated reactor fuels and materials in support of the LMFBR program. The NRAD facility utilizes a 250-kW TRIGA reactor and is completely dedicated to neutron radiographymore » and the development of radiography techniques. Criticality was first achieved at the NRAD reactor in October of 1977. Since that time, a number of modifications have been implemented to improve operational efficiency and radiography production. This paper describes the modifications and changes that significantly improved operational efficiency and reliability of the reactor and the essential auxiliary reactor systems. (author)« less

Radioactivity of spent TRIGA fuel

NASA Astrophysics Data System (ADS)

Usang, M. D.; Nabil, A. R. A.; Alfred, S. L.; Hamzah, N. S.; Abi, M. J. B.; Rawi, M. Z. M.; Abu, M. P.

2015-04-01

Some of the oldest TRIGA fuel in the Malaysian Reaktor TRIGA PUSPATI (RTP) is approaching the limit of its end of life with burn-up of around 20%. Hence it is prudent for us to start planning on the replacement of the fuel in the reactor and other derivative activities associated with it. In this regard, we need to understand all of the risk associated with such operation and one of them is to predict the radioactivity of the fuel, so as to estimate the safety of our working conditions. The radioactivity of several fuels are measured and compared with simulation results to confirm the burnup levels of the selected fuels. The radioactivity measurement are conducted inside the water tank to reduce the risk of exposure and in this case the detector wrapped in plastics are lowered under water. In nuclear power plant, the general practice was to continuously burn the fuel. In research reactor, most operations are based on the immediate needs of the reactor and our RTP for example operate periodically. By integrating the burnup contribution for each core configuration, we simplify the simulation of burn up for each core configuration. Our results for two (2) fuel however indicates that the dose from simulation underestimate the actual dose from our measurements. Several postulates are investigated but the underlying reason remain inconclusive.

Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

NASA Astrophysics Data System (ADS)

Mohammed, Abdul Aziz; Pauzi, Anas Muhamad; Rahman, Shaik Mohmmed Haikhal Abdul; Zin, Muhamad Rawi Muhammad; Jamro, Rafhayudi; Idris, Faridah Mohamad

2016-01-01

In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 (233U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintaining the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.

Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

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

Mohammed, Abdul Aziz, E-mail: azizM@uniten.edu.my; Rahman, Shaik Mohmmed Haikhal Abdul; Pauzi, Anas Muhamad, E-mail: anas@uniten.edu.my

2016-01-22

In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 ({sup 233}U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintainingmore » the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.« less

Fission products detection in irradiated TRIGA fuel by means of gamma spectroscopy and MCNP calculation.

PubMed

Cagnazzo, M; Borio di Tigliole, A; Böck, H; Villa, M

2018-05-01

Aim of this work was the detection of fission products activity distribution along the axial dimension of irradiated fuel elements (FEs) at the TRIGA Mark II research reactor of the Technische Universität (TU) Wien. The activity distribution was measured by means of a customized fuel gamma scanning device, which includes a vertical lifting system to move the fuel rod along its vertical axis. For each investigated FE, a gamma spectrum measurement was performed along the vertical axis, with steps of 1 cm, in order to determine the axial distribution of the fission products. After the fuel elements underwent a relatively short cooling down period, different fission products were detected. The activity concentration was determined by calibrating the gamma detector with a standard calibration source of known activity and by MCNP6 simulations for the evaluation of self-absorption and geometric effects. Given the specific TRIGA fuel composition, a correction procedure is developed and used in this work for the measurement of the fission product Zr 95 . This measurement campaign is part of a more extended project aiming at the modelling of the TU Wien TRIGA reactor by means of different calculation codes (MCNP6, Serpent): the experimental results presented in this paper will be subsequently used for the benchmark of the models developed with the calculation codes. Copyright © 2018 Elsevier Ltd. All rights reserved.

Automated power control system for reactor TRIGA PUSPATI

NASA Astrophysics Data System (ADS)

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

2017-01-01

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

Decommissioning of the TRIGA mark II and III and radioactive waste management

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

Doo Seong Hwang; Yoon Ji Lee; Gyeong Hwan Chung

2013-07-01

KAERI has carried out decommissioning projects for two research reactors (KRR-1 and 2). The decommissioning project of KRR-1 (TRIGA Mark II) and 2 (TRIGA Mark III) was launched in 1997 with a total budget of 23.25 million US dollars. KRR-2 and all auxiliary facilities were already decommissioned, and KRR-1 is being decommissioned now. Much more dismantled waste is generated than in any other operations of nuclear facilities. Thus, the waste needs to be reduced and stabilized through decontamination or treatment before disposal. This paper introduces the current status of the decommissioning projects and describes the volume reduction and conditioning ofmore » decommissioning waste for final disposal. (authors)« less

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...

Verification of MCNP simulation of neutron flux parameters at TRIGA MK II reactor of Malaysia.

PubMed

Yavar, A R; Khalafi, H; Kasesaz, Y; Sarmani, S; Yahaya, R; Wood, A K; Khoo, K S

2012-10-01

A 3-D model for 1 MW TRIGA Mark II research reactor was simulated. Neutron flux parameters were calculated using MCNP-4C code and were compared with experimental results obtained by k(0)-INAA and absolute method. The average values of φ(th),φ(epi), and φ(fast) by MCNP code were (2.19±0.03)×10(12) cm(-2)s(-1), (1.26±0.02)×10(11) cm(-2)s(-1) and (3.33±0.02)×10(10) cm(-2)s(-1), respectively. These average values were consistent with the experimental results obtained by k(0)-INAA. The findings show a good agreement between MCNP code results and experimental results. Copyright © 2012 Elsevier Ltd. All rights reserved.

Neutron detection of the Triga Mark III reactor, using nuclear track methodology

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

Espinosa, G., E-mail: espinosa@fisica.unam.mx; Golzarri, J. I.; Raya-Arredondo, R.

Nuclear Track Methodology (NTM), based on the neutron-proton interaction is one often employed alternative for neutron detection. In this paper we apply NTM to determine the Triga Mark III reactor operating power and neutron flux. The facility nuclear core, loaded with 85 Highly Enriched Uranium as fuel with control rods in a demineralized water pool, provide a neutron flux around 2 × 10{sup 12} n cm{sup −2} s{sup −1}, at the irradiation channel TO-2. The neutron field is measured at this channel, using Landauer{sup ®} PADC as neutron detection material, covered by 3 mm Plexiglas{sup ®} as converter. After exposure, plasticmore » detectors were chemically etched to make observable the formed latent tracks induced by proton recoils. The track density was determined by a custom made Digital Image Analysis System. The resulting average nuclear track density shows a direct proportionality response for reactor power in the range 0.1-7 kW. We indicate several advantages of the technique including the possibility to calibrate the neutron flux density measured at low reactor power.« less

Neutron measurement at the thermal column of the Malaysian Triga Mark II reactor using gold foil activation method and TLD

NASA Astrophysics Data System (ADS)

Shalbi, Safwan; Salleh, Wan Norhayati Wan; Mohamad Idris, Faridah; Aliff Ashraff Rosdi, Muhammad; Syahir Sarkawi, Muhammad; Liyana Jamsari, Nur; Nasir, Nur Aishah Mohd

2018-01-01

In order to design facilities for boron neutron capture therapy (BNCT), the neutron measurement must be considered to obtain the optimal design of BNCT facility such as collimator and shielding. The previous feasibility study showed that the thermal column could generate higher thermal neutrons yield for BNCT application at the TRIGA MARK II reactor. Currently, the facility for BNCT are planned to be developed at thermal column. Thus, the main objective was focused on the thermal neutron and epithermal neutron flux measurement at the thermal column. In this measurement, pure gold and cadmium were used as a filter to obtain the thermal and epithermal neutron fluxes from inside and outside of the thermal column door of the 200kW reactor power using a gold foil activation method. The results were compared with neutron fluxes using TLD 600 and TLD 700. The outcome of this work will become the benchmark for the design of BNCT collimator and the shielding

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.

Operation and reactivity measurements of an accelerator driven subcritical TRIGA reactor

NASA Astrophysics Data System (ADS)

O'Kelly, David Sean

Experiments were performed at the Nuclear Engineering Teaching Laboratory (NETL) in 2005 and 2006 in which a 20 MeV linear electron accelerator operating as a photoneutron source was coupled to the TRIGA (Training, Research, Isotope production, General Atomics) Mark II research reactor at the University of Texas at Austin (UT) to simulate the operation and characteristics of a full-scale accelerator driven subcritical system (ADSS). The experimental program provided a relatively low-cost substitute for the higher power and complexity of internationally proposed systems utilizing proton accelerators and spallation neutron sources for an advanced ADSS that may be used for the burning of high-level radioactive waste. Various instrumentation methods that permitted ADSS neutron flux monitoring in high gamma radiation fields were successfully explored and the data was used to evaluate the Stochastic Pulsed Feynman method for reactivity monitoring.

Design and characterization of an irradiation facility with real-time monitoring

NASA Astrophysics Data System (ADS)

Braisted, Jonathan David

Radiation causes performance degradation in electronics by inducing atomic displacements and ionizations. While radiation hardened components are available, non-radiation hardened electronics can be preferable because they are generally more compact, require less power, and less expensive than radiation tolerant equivalents. It is therefore important to characterize the performance of electronics, both hardened and non-hardened, to prevent costly system or mission failures. Radiation effects tests for electronics generally involve a handful of step irradiations, leading to poorly-resolved data. Step irradiations also introduce uncertainties in electrical measurements due to temperature annealing effects. This effect may be intensified if the time between exposure and measurement is significant. Induced activity in test samples also complicates data collection of step irradiated test samples. The University of Texas at Austin operates a 1.1 MW Mark II TRIGA research reactor. An in-core irradiation facility for radiation effects testing with a real-time monitoring capability has been designed for the UT TRIGA reactor. The facility is larger than any currently available non-central location in a TRIGA, supporting testing of larger electronic components as well as other in-core irradiation applications requiring significant volume such as isotope production or neutron transmutation doping of silicon. This dissertation describes the design and testing of the large in-core irradiation facility and the experimental campaign developed to test the real-time monitoring capability. This irradiation campaign was performed to test the real-time monitoring capability at various reactor power levels. The device chosen for characterization was the 4N25 general-purpose optocoupler. The current transfer ratio, which is an important electrical parameter for optocouplers, was calculated as a function of neutron fluence and gamma dose from the real-time voltage measurements. The resultant radiation effects data was seen to be repeatable and exceptionally finely-resolved. Therefore, the capability at UT TRIGA has been proven competitive with world-class effects characterization facilities.

The TRIGA Reactor Facility at the Armed Forces Radiobiology Research Institute: A Simplified Technical Description.

DTIC Science & Technology

1986-05-01

COUNT Technical FROM_ TO May 1986 20 16. SUPPLEMENTARY NOTATION 17. COSATI CODES 18. SUBJECT TERMS iConitinue on reverse if neceasary and identify by...Reactor, Modes of Operation, The AFRRI Reactor, Exposure Facilities, and Cerenkov Radiation. I- 20 DISTRISUTIONIAVAILABILITY OF ABSTRACT 21. ABSTRACT...6 Exposure Facilities 12 Cerenkov Radiation 17 Acoessiofl For NTIS GRA&I DT.C TABUnamnnounced [] UusnriOfltond -. By IZ Distribution/ Availability

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.

Analytical analyses of startup measurements associated with the first use of LEU fuel in Romania`s 14-MW TRIGA reactor

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

Bretscher, M.M.; Snelgrove, J.L.; Ciocanescu, M.

1992-12-01

The 14-MW TRIGA steady state reactor (SSR) is located in Pitesti, Romania. Beginning with an HEU core (10 wt% U), the reactor first went critical in November 1979 but was shut down ten years later because of insufficient excess reactivity. Last November the Institute for Nuclear Research (INR), which operates the SSR, received from the ANL RERTR program a shipment of 125 LEU pins fabricated by General Atomics and of the same geometry as the original fuel but with an enrichment of 19.7% 235U and a loading of 45 wt% U. Using 100 of these pins, four LEU clusters, eachmore » containing a 5 x 5 square array of fuel rods, were assembled. These four LEU clusters replaced the four most highly burned HEU elements in the SSR. The reactor resumed operations last February with a 35-element mixed HEU/LEU core configuration. In preparation for full power operation of the SSR with this mixed HEU/LEU core, a number of measurements were made. These included control rod calibrations, excess reactivity determinations, worths of experiment facilities, reaction rate distributions, and themocouple measurements of fuel temperatures as a function of reactor power. This paper deals with a comparison of some of these measured reactor parameters with corresponding analytical calculations.« less

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

PubMed

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

2017-12-01

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

MCNP study for epithermal neutron irradiation of an isolated liver at the Finnish BNCT facility.

PubMed

Kotiluoto, P; Auterinen, I

2004-11-01

A successful boron neutron capture treatment (BNCT) of a patient with multiple liver metastases has been first given in Italy, by placing the removed organ into the thermal neutron column of the Triga research reactor of the University of Pavia. In Finland, FiR 1 Triga reactor with an epithermal neutron beam well suited for BNCT has been extensively used to irradiate patients with brain tumors such as glioblastoma and recently also head and neck tumors. In this work we have studied by MCNP Monte Carlo simulations, whether it would be beneficial to treat an isolated liver with epithermal neutrons instead of thermal ones. The results show, that the epithermal field penetrates deeper into the liver and creates a build-up distribution of the boron dose. Our results strongly encourage further studying of irradiation arrangement of an isolated liver with epithermal neutron fields.

Preliminary study on new configuration with LEU fuel assemblies for the Dalat nuclear research reactor

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

Van Lam Pham; Vinh Vinh Le; Ton Nghiem Huynh

2008-07-15

The fuel conversion of the Dalat Nuclear Research Reactor (DNRR) is being realized. The DNRR is a pool type research reactor which was reconstructed from the 250 kW TRIGA- MARK II reactor. The reconstructed reactor attained its nominal power of 500 kW in February 1984. According to the results of design and safety analyses performed by the joint study between RERTR Program at Argonne National Laboratory (ANL) and Vietnam Atomic Energy Commission (VAEC) the mixed core of irradiated HEU and new LEU WWR-M2 fuel assemblies will be created soon. This paper presents the results of preliminary study on new configurationmore » with only LEU fuel assemblies for the DNRR. The codes MCNP, REBUS and VARI3D are used to calculate neutron flux performance in irradiation positions and kinetics parameters. The idea of change of Beryllium rod reloading enables to get working configuration assured shutdown margin, thermal-hydraulic safety and increase in thermal neutron flux in neutron trap at the center of DNRR active core. (author)« less

Relative fission product yield determination in the USGS TRIGA Mark I reactor

NASA Astrophysics Data System (ADS)

Koehl, Michael A.

Fission product yield data sets are one of the most important and fundamental compilations of basic information in the nuclear industry. This data has a wide range of applications which include nuclear fuel burnup and nonproliferation safeguards. Relative fission yields constitute a major fraction of the reported yield data and reduce the number of required absolute measurements. Radiochemical separations of fission products reduce interferences, facilitate the measurement of low level radionuclides, and are instrumental in the analysis of low-yielding symmetrical fission products. It is especially useful in the measurement of the valley nuclides and those on the extreme wings of the mass yield curve, including lanthanides, where absolute yields have high errors. This overall project was conducted in three stages: characterization of the neutron flux in irradiation positions within the U.S. Geological Survey TRIGA Mark I Reactor (GSTR), determining the mass attenuation coefficients of precipitates used in radiochemical separations, and measuring the relative fission products in the GSTR. Using the Westcott convention, the Westcott flux, modified spectral index, neutron temperature, and gold-based cadmium ratios were determined for various sampling positions in the USGS TRIGA Mark I reactor. The differential neutron energy spectrum measurement was obtained using the computer iterative code SAND-II-SNL. The mass attenuation coefficients for molecular precipitates were determined through experiment and compared to results using the EGS5 Monte Carlo computer code. Difficulties associated with sufficient production of fission product isotopes in research reactors limits the ability to complete a direct, experimental assessment of mass attenuation coefficients for these isotopes. Experimental attenuation coefficients of radioisotopes produced through neutron activation agree well with the EGS5 calculated results. This suggests mass attenuation coefficients of molecular precipitates can be approximated using EGS5, especially in the instance of radioisotopes produced predominantly through uranium fission. Relative fission product yields were determined for three sampling positions in the USGS TRIGA Mark I reactor through radiochemical analysis. The relative mass yield distribution for valley nuclides decreases with epithermal neutrons compared to thermal neutrons. Additionally, a proportionality constant which related the measured beta activity of a fission product to the number of fissions that occur in a sample of irradiated uranium was determined for the detector used in this study and used to determine the thermal and epithermal flux. These values agree well with a previous study which used activation foils to determine the flux. The results of this project clearly demonstrate that R-values can be measured in the GSTR.

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

Edgue, E.

The point kinetics approach is a classical useful method for a reactor transient analysis. It is helpful to known, however, when a more elaborate transient analysis, involving the space-dependence change of the flux through a given transient, should be considered. In this paper, the authors present a rather elegant and quick method to check the need for a space-dependent flux analysis through a control rod transient in a given nuclear reactor. The method is applied to a series of rod ejection experiments in the TRIGA MARK-II reactor of Istanbul Technical University (ITU).

An approach to model reactor core nodalization for deterministic safety analysis

NASA Astrophysics Data System (ADS)

Salim, Mohd Faiz; Samsudin, Mohd Rafie; Mamat @ Ibrahim, Mohd Rizal; Roslan, Ridha; Sadri, Abd Aziz; Farid, Mohd Fairus Abd

2016-01-01

Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH1.6, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D® computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.

An approach to model reactor core nodalization for deterministic safety analysis

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

Salim, Mohd Faiz, E-mail: mohdfaizs@tnb.com.my; Samsudin, Mohd Rafie, E-mail: rafies@tnb.com.my; Mamat Ibrahim, Mohd Rizal, E-mail: m-rizal@nuclearmalaysia.gov.my

Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to bemore » employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH{sub 1.6}, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D{sup ®} computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.« less

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

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

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

2014-01-01

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

Utilization of thorium and U-ZrH1.6 fuels in various heterogeneous cores for TRIGA PUSPATI Reactor (RTP)

NASA Astrophysics Data System (ADS)

Damahuri, Abdul Hannan Bin; Mohamed, Hassan; Aziz Mohamed, Abdul; Idris, Faridah

2018-01-01

The use of thorium as nuclear fuel has been an appealing prospect for many years and will be great significance to nuclear power generation. There is an increasing need for more research on thorium as Malaysian government is currently active in the national Thorium Flagship Project, which was launched in 2014. The thorium project, which is still in phase 1, focuses on the research and development of the thorium extraction from mineral processing ore. Thus, the aim of the study is to investigate other alternative TRIGA PUSPATI Reactor (RTP) core designs that can fully utilize thorium. Currently, the RTP reactor has an average neutron flux of 2.797 x 1012 cm-2/s-1 and an effective multiplication factor, k eff, of 1.001. The RTP core has a circular array core configuration with six circular rings. Each ring consists of 6, 12, 18, 24, 30 or 36 U-ZrH1.6 fuel rods. There are three main type of uranium weight, namely 8.5, 12 and 20 wt.%. For this research, uranium zirconium hydride (U-ZrH1.6) fuel rods in the RTP core were replaced by thorium (ThO2) fuel rods. Seven core configurations with different thorium fuel rods placements were modelled in a 2D structure and simulated using Monte Carlo n-particle (MCNPX) code. Results show that the highest initial criticality obtained is around 1.35101. Additionally there is a significant discrepancy between results from previous study and the work because of the large estimated leakage probability of approximately 21.7% and 2D model simplification.

Cryostat system for investigation on new neutron moderator materials at reactor TRIGA PUSPATI

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

Dris, Zakaria bin, E-mail: zakariadris@gmail.com; Centre for Nuclear Energy, Universiti Tenaga Nasional; Mohamed, Abdul Aziz bin

2016-01-22

A simple continuous flow (SCF) cryostat was designed to investigate the neutron moderation of alumina in high temperature co-ceramic (HTCC) and polymeric materials such as Teflon under TRIGA neutron environment using a reflected neutron beam from a monochromator. Cooling of the cryostat will be carried out using liquid nitrogen. The cryostat will be built with an aluminum holder for moderator within stainless steel cylinder pipe. A copper thermocouple will be used as the temperature sensor to monitor the moderator temperature inside the cryostat holder. Initial measurements of neutron spectrum after neutron passing through the moderating materials have been carried outmore » using a neutron spectrometer.« less

Confirmation of a realistic reactor model for BNCT dosimetry at the TRIGA Mainz

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

Ziegner, Markus, E-mail: Markus.Ziegner.fl@ait.ac.at; Schmitz, Tobias; Hampel, Gabriele

2014-11-01

Purpose: In order to build up a reliable dose monitoring system for boron neutron capture therapy (BNCT) applications at the TRIGA reactor in Mainz, a computer model for the entire reactor was established, simulating the radiation field by means of the Monte Carlo method. The impact of different source definition techniques was compared and the model was validated by experimental fluence and dose determinations. Methods: The depletion calculation code ORIGEN2 was used to compute the burn-up and relevant material composition of each burned fuel element from the day of first reactor operation to its current core. The material composition ofmore » the current core was used in a MCNP5 model of the initial core developed earlier. To perform calculations for the region outside the reactor core, the model was expanded to include the thermal column and compared with the previously established ATTILA model. Subsequently, the computational model is simplified in order to reduce the calculation time. Both simulation models are validated by experiments with different setups using alanine dosimetry and gold activation measurements with two different types of phantoms. Results: The MCNP5 simulated neutron spectrum and source strength are found to be in good agreement with the previous ATTILA model whereas the photon production is much lower. Both MCNP5 simulation models predict all experimental dose values with an accuracy of about 5%. The simulations reveal that a Teflon environment favorably reduces the gamma dose component as compared to a polymethyl methacrylate phantom. Conclusions: A computer model for BNCT dosimetry was established, allowing the prediction of dosimetric quantities without further calibration and within a reasonable computation time for clinical applications. The good agreement between the MCNP5 simulations and experiments demonstrates that the ATTILA model overestimates the gamma dose contribution. The detailed model can be used for the planning of structural modifications in the thermal column irradiation channel or the use of different irradiation sites than the thermal column, e.g., the beam tubes.« less

Neutron flux and gamma dose measurement in the BNCT irradiation facility at the TRIGA reactor of the University of Pavia

NASA Astrophysics Data System (ADS)

Bortolussi, S.; Protti, N.; Ferrari, M.; Postuma, I.; Fatemi, S.; Prata, M.; Ballarini, F.; Carante, M. P.; Farias, R.; González, S. J.; Marrale, M.; Gallo, S.; Bartolotta, A.; Iacoviello, G.; Nigg, D.; Altieri, S.

2018-01-01

University of Pavia is equipped with a TRIGA Mark II research nuclear reactor, operating at a maximum steady state power of 250 kW. It has been used for many years to support Boron Neutron Capture Therapy (BNCT) research. An irradiation facility was constructed inside the thermal column of the reactor to produce a sufficient thermal neutron flux with low epithermal and fast neutron components, and low gamma dose. In this irradiation position, the liver of two patients affected by hepatic metastases from colon carcinoma were irradiated after borated drug administration. The facility is currently used for cell cultures and small animal irradiation. Measurements campaigns have been carried out, aimed at characterizing the neutron spectrum and the gamma dose component. The neutron spectrum has been measured by means of multifoil neutron activation spectrometry and a least squares unfolding algorithm; gamma dose was measured using alanine dosimeters. Results show that in a reference position the thermal neutron flux is (1.20 ± 0.03) ×1010 cm-2 s-1 when the reactor is working at the maximum power of 250 kW, with the epithermal and fast components, respectively, 2 and 3 orders of magnitude lower than the thermal component. The ratio of the gamma dose with respect to the thermal neutron fluence is 1.2 ×10-13 Gy/(n/cm2).

NDI (Nondestructive Inspection) Oriented Corrosion Control for Army Aircraft. Phase 1. Inspection Methods

DTIC Science & Technology

1989-07-01

Appendices A and B and are provided as cover sheets from each item rather than completc packages. The Pamplet Series materials were furnished as camera-ready...34Stational Neutron Radiography System for Aircraft Reliability and Maintainability." G. A. Technologies Brochure , Triga Reactor Division, San Diego

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

PubMed

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

2015-10-09

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

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

PubMed Central

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

2015-01-01

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

Testing of a Transport Cask for Research Reactor Spent Fuel - 13003

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

Mourao, Rogerio P.; Leite da Silva, Luiz; Miranda, Carlos A.

2013-07-01

Since the beginning of the last decade three Latin American countries that operate research reactors - Argentina, Brazil and Chile - have been joining efforts to improve the regional capability in the management of spent fuel elements from the TRIGA and MTR reactors operated in the region. A main drive in this initiative, sponsored by the International Atomic Energy Agency, is the fact that no definite solution regarding the back end of the research reactor fuel cycle has been taken by any of the participating country. However, any long-term solution - either disposition in a repository or storage away frommore » reactor - will involve at some stage the transportation of the spent fuel through public roads. Therefore, a licensed cask that provides adequate shielding, assurance of subcriticality, and conformance to internationally accepted safety, security and safeguards regimes is considered a strategic part of any future solution to be adopted at a regional level. As a step in this direction, a packaging for the transport of irradiated fuel for MTR and TRIGA research reactors was designed by the tri-national team and a half-scale model equipped with the MTR version of the internal basket was constructed in Argentina and Brazil and tested in Brazil. Three test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions. After failing the tests in the first two test series, the specimen successfully underwent the last test sequence. A second specimen, incorporating the structural improvements in view of the previous tests results, will be tested in the near future. Numerical simulations of the free drop and thermal tests are being carried out in parallel, in order to validate the computational modeling that is going to be used as a support for the package certification. (authors)« less

77 FR 68155 - The Armed Forces Radiobiology Research Institute TRIGA Reactor: Facility Operating License No. R-84

Federal Register 2010, 2011, 2012, 2013, 2014

2012-11-15

... a determination by the presiding officer that the filing demonstrates good cause by satisfying the... electronic storage media. Participants may not submit paper copies of their filings unless they seek an... through Friday, excluding government holidays. Participants who believe that they have a good cause for...

Calculations to Support On-line Neutron Spectrum Adjustment by Measurements with Miniature Fission Chambers in the JSI TRIGA Reactor

NASA Astrophysics Data System (ADS)

Kaiba, Tanja; Radulović, Vladimir; Žerovnik, Gašper; Snoj, Luka; Fourmentel, Damien; Barbot, LoÏc; Destouches, Christophe AE(; )

2018-01-01

Preliminary calculations were performed with the aim to establish optimal experimental conditions for the measurement campaign within the collaboration between the Jožef Stefan Institute (JSI) and Commissariat à l'Énergie Atomique et aux Énergies Alternatives (CEA Cadarache). The goal of the project is to additionally characterize the neutron spectruminside the JSI TRIGA reactor core with focus on the measurement epi-thermal and fast part of the spectrum. Measurements will be performed with fission chambers containing different fissile materials (235U, 237Np and 242Pu) covered with thermal neutron filters (Cd and Gd). The changes in the detected signal and neutron flux spectrum with and without transmission filter were studied. Additional effort was put into evaluation of the effect of the filter geometry (e.g. opening on the top end of the filter) on the detector signal. After the analysis of the scoping calculations it was concluded to position the experiment in the outside core ring inside one of the empty fuel element positions.

Testing the applicability of the k0-NAA method at the MINT's TRIGA MARK II reactor

NASA Astrophysics Data System (ADS)

Siong, Wee Boon; Dung, Ho Manh; Wood, Ab. Khalik; Salim, Nazaratul Ashifa Abd.; Elias, Md. Suhaimi

2006-08-01

The Analytical Chemistry Laboratory at MINT is using the NAA technique since 1980s and is the only laboratory in Malaysia equipped with a research reactor, namely the TRIGA MARK II. Throughout the years the development of NAA technique has been very encouraging and was made applicable to a wide range of samples. At present, the k0 method has become the preferred standardization method of NAA ( k0-NAA) due to its multi-elemental analysis capability without using standards. Additionally, the k0 method describes NAA in physically and mathematically understandable definitions and is very suitable for computer evaluation. Eventually, the k0-NAA method has been adopted by MINT in 2003, in collaboration with the Nuclear Research Institute (NRI), Vietnam. The reactor neutron parameters ( α and f) for the pneumatic transfer system and for the rotary rack at various locations, as well as the detector efficiencies were determined. After calibration of the reactor and the detectors, the implemented k0 method was validated by analyzing some certified reference materials (including IAEA Soil 7, NIST 1633a, NIST 1632c, NIST 1646a and IAEA 140/TM). The analysis results of the CRMs showed an average u score well below the threshold value of 2 with a precision of better than ±10% for most of the elemental concentrations obtained, validating herewith the introduction of the k0-NAA method at the MINT.

Computational analysis of the dose rates at JSI TRIGA reactor irradiation facilities.

PubMed

Ambrožič, K; Žerovnik, G; Snoj, L

2017-12-01

The JSI TRIGA Mark II, IJS research reactor is equipped with numerous irradiation positions, where samples can be irradiated by neutrons and γ-rays. Irradiation position selection is based on its properties, such as physical size and accessibility, as well as neutron and γ-ray spectra, flux and dose intensities. This paper presents an overview on the neutron and γ-ray fluxes, spectra and dose intensities calculations using Monte Carlo MCNP software and ENDF/B-VII.0 nuclear data libraries. The dose-rates are presented in terms of ambient dose equivalents, air kerma, and silicon dose equivalent. At full reactor power the neutron ambient dose equivalent ranges from 5.5×10 3 Svh -1 to 6×10 6 Svh -1 , silicon dose equivalent from 6×10 2 Gy/h si to 3×10 5 Gy/h si , and neutron air kerma from 4.3×10 3 Gyh -1 to 2×10 5 Gyh -1 . Ratio of fast (1MeV

High Temperature Fuel Cladding Chemical Interactions Between TRIGA Fuels and 304 Stainless Steel

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

Perez, Emmanuel; Keiser, Jr., Dennis D.; Forsmann, Bryan

High-temperature fuel-cladding chemical interactions (FCCI) between TRIGA (Training, Research, Isotopes, General Atomics) fuel elements and the 304 stainless steel (304SS) are of interest to develop an understanding of the fuel behavior during transient reactor scenarios. TRIGA fuels are composed of uranium (U) particles dispersed in a zirconium-hydride (Zr-H) matrix. In reactor, the fuel is encased in 304-stainless-steel (304SS) or Incoloy 800 clad tubes. At high temperatures, the fuel can readily interact with the cladding, resulting in FCCI. A number of FCCI can take place in this system. Interactions can be expected between the cladding and the Zr-H matrix, and/or betweenmore » the cladding and the U-particles. Other interactions may be expected between the Zr-H matrix and the U-particles. Furthermore, the fuel contains erbium-oxide (Er-O) additions. Interactions can also be expected between the Er-O, the cladding, the Zr-H and the U-particles. The overall result is that very complex interactions may take place as a result of fuel and cladding exposures to high temperatures. This report discusses the characterization of the baseline fuel microstructure in the as-received state (prior to exposure to high temperature), characterization of the fuel after annealing at 950C for 24 hours and the results from diffusion couple experiments carries out at 1000C for 5 and 24 hours. Characterization was carried out via scanning electron microscopy (SEM) and transmission electron microscopy (TEM) with sample preparation via focused ion beam in situ-liftout-technique.« less

Performance evaluation and post-irradiation examination of a novel LWR fuel composed of U0.17ZrH1.6 fuel pellets bonded to Zircaloy-2 cladding by lead bismuth eutectic

NASA Astrophysics Data System (ADS)

Balooch, Mehdi; Olander, Donald R.; Terrani, Kurt A.; Hosemann, Peter; Casella, Andrew M.; Senor, David J.; Buck, Edgar C.

2017-04-01

A novel light water reactor fuel has been designed and fabricated at the University of California, Berkeley; irradiated at the Massachusetts Institute of Technology Reactor; and examined within the Radiochemical Processing Laboratory at the Pacific Northwest National Laboratory. This fuel consists of U0.17ZrH1.6 fuel pellets core-drilled from TRIGA reactor fuel elements that are clad in Zircaloy-2 and bonded with lead-bismuth eutectic. The performance evaluation and post irradiation examination of this fuel are presented here.

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

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

Little, G.A.

For better than ten years there was little public notice of the TRIGA reactor at UC-Berkeley. Then: a) A non-student persuaded the Student and Senate to pass a resolution to request Campus Administration to stop operation of the reactor and remove it from campus. b) Presence of the reactor became a campaign-issue in a City Mayoral election. c) Two local residents reported adverse physical reactions before, during, and after a routine tour of the reactor facility. d) The Berkeley City Council began a study of problems associated with radioactive material within the city. e) Friends Of The Earth formally petitionedmore » the NRC to terminate the reactor's license. Campus personnel have expended many man-hours and many pounds of paper in responding to these happenings. Some of the details are of interest, and may be of use to other reactor facilities. (author)« 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.

TRIGA MARK-II source term

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

Usang, M. D., E-mail: mark-dennis@nuclearmalaysia.gov.my; Hamzah, N. S., E-mail: mark-dennis@nuclearmalaysia.gov.my; Abi, M. J. B., E-mail: mark-dennis@nuclearmalaysia.gov.my

ORIGEN 2.2 are employed to obtain data regarding γ source term and the radio-activity of irradiated TRIGA fuel. The fuel composition are specified in grams for use as input data. Three types of fuel are irradiated in the reactor, each differs from the other in terms of the amount of Uranium compared to the total weight. Each fuel are irradiated for 365 days with 50 days time step. We obtain results on the total radioactivity of the fuel, the composition of activated materials, composition of fission products and the photon spectrum of the burned fuel. We investigate the differences ofmore » results using BWR and PWR library for ORIGEN. Finally, we compare the composition of major nuclides after 1 year irradiation of both ORIGEN library with results from WIMS. We found only minor disagreements between the yields of PWR and BWR libraries. In comparison with WIMS, the errors are a little bit more pronounced. To overcome this errors, the irradiation power used in ORIGEN could be increased a little, so that the differences in the yield of ORIGEN and WIMS could be reduced. A more permanent solution is to use a different code altogether to simulate burnup such as DRAGON and ORIGEN-S. The result of this study are essential for the design of radiation shielding from the fuel.« less

Modifications to the NRAD Reactor, 1977 to present

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

Weeks, A.A.; Pruett, D.P.; Heidel, C.C.

1986-01-01

Argonne National Laboratory-West, operated by the University of Chicago, is located near Idaho Falls, ID, on the Idaho National Engineering laboratory Site. ANL-West performs work in support of the Liquid Metal Fast Breeder Reactor Program (LMFBR) sponsored by the United States Department of Energy. The NRAD reactor is located at the Argonne Site within the Hot Fuel Examination Facility/North, a large hot cell facility where both non-destructive and destructive examinations are performed on highly irradiated reactor fuels and materials in support of the LMFBR program. The NRAD facility utilizes a 250-kW TRIGA reactor and is completely dedicated to neutron radiographymore » and the development of radiography techniques. Criticality was first achieved at the NRAD reactor in October of 1977. Since that time, a number of modifications have been implemented to improve operational efficiency and radiography production. This paper describes the modifications and changes that significantly improved operational efficiency and reliability of the reactor and the essential auxiliary reactor systems.« less

Delayed Gamma Measurements in Different Nuclear Research Reactors Bringing Out the Importance of the Delayed Contribution in Gamma Flux Calculations

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

Fourmentel, D.; Radulovic, V.; Barbot, L.

Neutron and gamma flux levels are key parameters in nuclear research reactors. In Material Testing Reactors, such as the future Jules Horowitz Reactor, under construction at the French Alternative Energies and Atomic Energy Commission (CEA Cadarache, France), the expected gamma flux levels are very high (nuclear heating is of the order of 20 W/g at 100 MWth). As gamma rays deposit their energy in the reactor structures and structural materials it is important to take them into account when designing irradiation devices. There are only a few sensors which allow measurements of the nuclear heating ; a recent development atmore » the CEA Cadarache allows measurements of the gamma flux using a miniature ionization chamber (MIC). The measured MIC response is often compared with calculation using modern Monte Carlo (MC) neutron and photon transport codes, such as TRIPOLI-4 and MCNP6. In these calculations only the production of prompt gamma rays in the reactor is usually modelled thus neglecting the delayed gamma rays. Hence calculations and measurements are usually in better accordance for the neutron flux than for the gamma flux. In this paper we study the contribution of delayed gamma rays to the total MIC signal in order to estimate the systematic error in gamma flux MC calculations. In order to experimentally determine the delayed gamma flux contributions to the MIC response, we performed gamma flux measurements with CEA developed MIC at three different research reactors: the OSIRIS reactor (MTR - 70 MWth at CEA Saclay, France), the TRIGA MARK II reactor (TRIGA - 250 kWth at the Jozef Stefan Institute, Slovenia) and the MARIA reactor (MTR - 30 MWth at the National Center for Nuclear Research, Poland). In order to experimentally assess the delayed gamma flux contribution to the total gamma flux, several reactor shut down (scram) experiments were performed specifically for the purpose of the measurements. Results show that on average about 30 % of the MIC signal is due to the delayed gamma rays. In this paper we describe experiments in each of the three reactors and how we estimate delayed gamma rays with MIC measurements. The results and perspectives are discussed. (authors)« less

Fundamental approaches for analysis thermal hydraulic parameter for Puspati Research Reactor

NASA Astrophysics Data System (ADS)

Hashim, Zaredah; Lanyau, Tonny Anak; Farid, Mohamad Fairus Abdul; Kassim, Mohammad Suhaimi; Azhar, Noraishah Syahirah

2016-01-01

The 1-MW PUSPATI Research Reactor (RTP) is the one and only nuclear pool type research reactor developed by General Atomic (GA) in Malaysia. It was installed at Malaysian Nuclear Agency and has reached the first criticality on 8 June 1982. Based on the initial core which comprised of 80 standard TRIGA fuel elements, the very fundamental thermal hydraulic model was investigated during steady state operation using the PARET-code. The main objective of this paper is to determine the variation of temperature profiles and Departure of Nucleate Boiling Ratio (DNBR) of RTP at full power operation. The second objective is to confirm that the values obtained from PARET-code are in agreement with Safety Analysis Report (SAR) for RTP. The code was employed for the hot and average channels in the core in order to calculate of fuel's center and surface, cladding, coolant temperatures as well as DNBR's values. In this study, it was found that the results obtained from the PARET-code showed that the thermal hydraulic parameters related to safety for initial core which was cooled by natural convection was in agreement with the designed values and safety limit in SAR.

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

NONE

Digital instrumentation and controls system technique is being introduced in new constructed research reactor or life extension of older research reactor. Digital systems are easy to change and optimize but the validated process for them is required. Also, to reduce project risk or cost, we have to make it sure that configuration and control functions is right before the commissioning phase on research reactor. For this purpose, simulators have been widely used in developing control systems in automotive and aerospace industries. In these literatures, however, very few of these can be found regarding test on the control system of researchmore » reactor with simulator. Therefore, this paper proposes a simulation platform to verify the performance of RRS (Reactor Regulating System) for research reactor. This simulation platform consists of the reactor simulation model and the interface module. This simulation platform is applied to I and C upgrade project of TRIGA reactor, and many problems of RRS configuration were found and solved. And it proved that the dynamic performance testing based on simulator enables significant time saving and improves economics and quality for RRS in the system test phase. (authors)« less

Benchmarking of Neutron Flux Parameters at the USGS TRIGA Reactor in Lakewood, Colorado

NASA Astrophysics Data System (ADS)

Alzaabi, Osama E.

The USGS TRIGA Reactor (GSTR) located at the Denver Federal Center in Lakewood Colorado provides opportunities to Colorado School of Mines students to do experimental research in the field of neutron activation analysis. The scope of this thesis is to obtain precise knowledge of neutron flux parameters at the GSTR. The Colorado School of Mines Nuclear Physics group intends to develop several research projects at the GSTR, which requires the precise knowledge of neutron fluxes and energy distributions in several irradiation locations. The fuel burn-up of the new GSTR fuel configuration and the thermal neutron flux of the core were recalculated since the GSTR core configuration had been changed with the addition of two new fuel elements. Therefore, a MCNP software package was used to incorporate the burn up of reactor fuel and to determine the neutron flux at different irradiation locations and at flux monitoring bores. These simulation results were compared with neutron activation analysis results using activated diluted gold wires. A well calibrated and stable germanium detector setup as well as fourteen samplers were designed and built to achieve accuracy in the measurement of the neutron flux. Furthermore, the flux monitoring bores of the GSTR core were used for the first time to measure neutron flux experimentally and to compare to MCNP simulation. In addition, International Atomic Energy Agency (IAEA) standard materials were used along with USGS national standard materials in a previously well calibrated irradiation location to benchmark simulation, germanium detector calibration and sample measurements to international standards.

Experimental parameters optimization of instrumental neutron activation analysis in order to determine selected elements in some industrial soils in Turkey

NASA Astrophysics Data System (ADS)

Haciyakupoglu, Sevilay; Nur Esen, Ayse; Erenturk, Sema

2014-08-01

The purpose of this study is optimization of the experimental parameters for analysis of soil matrix by instrumental neutron activation analysis and quantitative determination of barium, cerium, lanthanum, rubidium, scandium and thorium in soil samples collected from industrialized urban areas near Istanbul. Samples were irradiated in TRIGA MARK II Research Reactor of Istanbul Technical University. Two types of reference materials were used to check the accuracy of the applied method. The achieved results were found to be in compliance with certified values of the reference materials. The calculated En numbers for mentioned elements were found to be less than 1. The presented data of element concentrations in soil samples will help to trace the pollution as an impact of urbanization and industrialization, as well as providing database for future studies.

Neutronics calculation of RTP core

NASA Astrophysics Data System (ADS)

Rabir, Mohamad Hairie B.; Zin, Muhammad Rawi B. Mohamed; Karim, Julia Bt. Abdul; Bayar, Abi Muttaqin B. Jalal; Usang, Mark Dennis Anak; Mustafa, Muhammad Khairul Ariff B.; Hamzah, Na'im Syauqi B.; Said, Norfarizan Bt. Mohd; Jalil, Muhammad Husamuddin B.

2017-01-01

Reactor calculation and simulation are significantly important to ensure safety and better utilization of a research reactor. The Malaysian's PUSPATI TRIGA Reactor (RTP) achieved initial criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes. Since early 90s, neutronics modelling were used as part of its routine in-core fuel management activities. The are several computer codes have been used in RTP since then, based on 1D neutron diffusion, 2D neutron diffusion and 3D Monte Carlo neutron transport method. This paper describes current progress and overview on neutronics modelling development in RTP. Several important parameters were analysed such as keff, reactivity, neutron flux, power distribution and fission product build-up for the latest core configuration. The developed core neutronics model was validated by means of comparison with experimental and measurement data. Along with the RTP core model, the calculation procedure also developed to establish better prediction capability of RTP's behaviour.

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

NASA Astrophysics Data System (ADS)

Ciftcioglu, Özer

1987-05-01

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

Trace elements study of high purity nanocrystalline silicon carbide (3C-SiC) using k0-INAA method

NASA Astrophysics Data System (ADS)

Huseynov, Elchin; Jazbec, Anze

2017-07-01

Silicon carbide (3C-SiC) nanoparticles have been irradiated by neutron flux (2×1013 n·cm-2·s-1) at TRIGA Mark II type research reactor. After neutron irradiation, the radioisotopes of trace elements in the nanocrystalline 3C-SiC were studied as time functions. The identification of isotopes which significantly increased the activity of the samples as a result of neutron radiation was carried out. Nanocrystalline 3C-SiC are synthesized by standard laser technique and the purity of samples was determined by the k0-based Instrumental Neutron Activation Analysis (k0-INAA) method. Trace elements concentration in the 3C-SiC nanoparticles were determined by the radionuclides of appropriate elements. The trace element isotopes concentration have been calculated in percentage according to k0-INAA method.

SAFEGUARDS REPORT FOR THE NORTHROP PULSE RADIATION FACILITY

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

Feinauer, E.; Thomas, R.D.

1961-03-22

Ae description is given of the Northrop pulse Radiation Facility, (NPRF), which consists of a TRlGA Mark-F reactor and associated supporting equipment. The NPRF was designed to operate in the following modes: Mode 1-100 kw steady-state operation; Mode II--Pulsed operation up to a maximum transient giving a maximum measured fuel element temperature of 470 deg C, which corresponds to an energy release of about 18 Mw-sec (approximately 1.9% sigma K/ K). The movable reactor will be operated in three general areas in the pool: adjacent to the exposure room; adjacent to the beam ponts; or at intermediate positions. Based onmore » the analyses presented and operating experience with the prototype TRIGA Mark F and other TRlGA reactors, it is concluded that operation of the NPRF does not present any undue hazard to the health and safety of the operating personnel or the public. (auth)« less

Preparation and quantification of radioactive particles for tracking hydrodynamic behavior in multiphase reactors.

PubMed

Yunos, Mohd Amirul Syafiq Mohd; Hussain, Siti Aslina; Yusoff, Hamdan Mohamed; Abdullah, Jaafar

2014-09-01

Radioactive particle tracking (RPT) has emerged as a promising and versatile technique that can provide rich information about a variety of multiphase flow systems. However, RPT is not an off-the-shelf technique, and thus, users must customize RPT for their applications. This paper presents a simple procedure for preparing radioactive tracer particles created via irradiation with neutrons from the TRIGA Mark II research reactor. The present study focuses on the performance evaluation of encapsulated gold and scandium particles for applications as individual radioactive tracer particles using qualitative and quantitative neutron activation analysis (NAA) and an X-ray microcomputed tomography (X-ray Micro-CT) scanner installed at the Malaysian Nuclear Agency. Copyright © 2014 Elsevier Ltd. All rights reserved.

An experimental investigation of 235 sub UF sub 6 fission produced plasmas. [gas handling system for use with nuclear pumped laser experiments

NASA Technical Reports Server (NTRS)

Miley, G. H.

1981-01-01

A gas handling system capable of use with uranium fluoride was designed and constructed for use with nuclear pumped laser experiments using the TRIGA research reactor. By employing careful design and temperature controls, the UF6 can be first transported into the irradiation chamber, and then, at the conclusion of the experiment, returned to gas cylinders. The design of the system is described. Operating procedures for the UF6 and gas handling systems are included.

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

Metz, Lori A.; Friese, Judah I.; Finn, Erin C.

Critical assemblies provide one method of achieving a fast neutron spectrum that is close to a 235U fission-energy neutron spectrum for nuclear data measurements. Previous work has demonstrated the use of a natural boron carbide capsule for spectral-tailoring in a mixed spectrum reactor as an alternate and complementary method for performing fission-energy neutron experiments. Previous fission products measurements showed that the neutron spectrum achievable with natural boron carbide was not as hard as what can be achieved with critical assemblies. New measurements performed with the Washington State University TRIGA reactor using a boron carbide capsule 96% enriched in 10B formore » irradiations resulted in a neutron spectrum very similar to a critical assembly and a pure 235U fission spectrum. The current work describes an experiment involving a highly-enriched uranium target irradiated under the new 10B4C capsule. Fission product yields were measured following radiochemical separations and are presented here. Reactor dosimetry measurements for characterizing neutron spectra and fluence for the enriched boron carbide capsule and critical assemblies are also discussed.« less

Sister Lab Program Prospective Partner Nuclear Profile: Indonesia

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

Bissani, M; Tyson, S

2006-12-14

Indonesia has participated in cooperative technical programs with the IAEA since 1957, and has cooperated with regional partners in all of the traditional areas where nuclear science is employed: in medicine, public health (such as insect control and eradication programs), agriculture (e.g. development of improved varieties of rice), and the gas and oil industries. Recently, Indonesia has contributed significantly to the Reduced Enrichment Research and Training Reactor (RERTR) Program by conducting experiments to confirm the feasibility of Mo-99 production using high-density low enriched uranium (LEU) fuel, a primary goal of the RERTR Program. Indonesia's first research reactor, the TRIGA Markmore » II at Bandung, began operation in 1964 at 250 kW and was subsequently upgraded in 1971 to 1 MW and further upgraded in 2000 to 2 MW. This reactor was joined by another TRIGA Mark II, the 100-kW Kartini-PPNY at Yogyakarta, in 1979, and by the 30-MW G.A. Siwabessy multipurpose reactor in Serpong, which achieved criticality in July 1983. A 10-MW radioisotope production reactor, to be called the RPI-10, also was proposed for construction at Serpong in the late 1990s, but the project apparently was not carried out. In the five decades since its nuclear research program began, Indonesia has trained a cadre of scientific and technical staff who not only operate and conduct research with the current facilities, but also represent the nucleus of a skilled labor pool to support development of a nuclear power program. Although Indonesia's previous on-again, off-again consideration of nuclear power has not gotten very far in the past, it now appears that Indonesia again is giving serious consideration to beginning a national nuclear energy program. In June 2006, Research and Technology Minister Kusmayanto Kadiman said that his ministry was currently putting the necessary procedures in place to speed up the project to acquire a nuclear power plant, indicating that, ''We will need around five years to complete the project. If we can start the study, go to tender, and sign the contract for the project this year, the power plant could be on stream by 2011''. While this ambitious schedule may be a bit unrealistic, it suggests new momentum to move forward on the project. The favored site for the proposed plant is the Muria Peninsula, located on Java's north central coast.« 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

Radiation Damage Study in Natural Zircon Using Neutrons Irradiation

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

Lwin, Maung Tin Moe; Amin, Yusoff Mohd.; Kassim, Hasan Abu

2011-03-30

Changes of atomic displacements in crystalline structure of natural zircon (ZrSiO{sub 4}) can be studied by using neutron irradiation on the surface of zircon and compared the data from XRD measurements before and after irradiation. The results of neutron irradiation on natural zircon using Pneumatic Transfer System (PTS) at PUSPATI TRIGA Research Reactor in the Malaysian Nuclear Agency are discussed in this work. The reactor produces maximum thermal power output of 1 MWatt and the neutron flux of up to 1x10{sup 13} ncm{sup -2}s{sup -1}. From serial decay processes of uranium and thorium radionuclides in zircon crystalline structure, the emissionmore » of alpha particles can produce damage in terms of atomic displacements in zircon. Hence, zircon has been extensively studied as a possible candidate for immobilization of fission products and actinides.« less

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

NASA Astrophysics Data System (ADS)

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

2002-12-01

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

TRIGA Mark II nuclear reactor facility. Final report, 1 July 1980--30 June 1995

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

Ryan, B.C.

1997-05-01

This report is a final culmination of activities funded through the Department of Energy`s (DOE) University Reactor Sharing Program, Grant DE-FG02-80ER10273, during the period 1 July 1980 through 30 June 1995. Progress reports have been periodically issued to the DOE, namely the Reactor Facility Annual Reports C00-2082/2219-7 through C00-2082/10723-21, which are contained as an appendix to this report. Due to the extent of time covered by this grant, summary tables are presented. Table 1 lists the fiscal year financial obligations of the grant. As listed in the original grant proposals, the DOE grant financed 70% of project costs, namely themore » total amount spent of these projects minus materials costs and technical support. Thus the bulk of funds was spent directly on reactor operations. With the exception of a few years, spending was in excess of the grant amount. As shown in Tables 2 and 3, the Reactor Sharing grant funded a immense number of research projects in nuclear engineering, geology, animal science, chemistry, anthropology, veterinary medicine, and many other fields. A list of these users is provided. Out of the average 3000 visitors per year, some groups participated in classes involving the reactor such as Boy Scout Merit Badge classes, teacher`s workshops, and summer internships. A large number of these projects met the requirements for the Reactor Sharing grant, but were funded by the University instead.« less

Multi-University Southeast INIE Consortium

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

Ayman Hawari; Nolan Hertel; Mohamed Al-Sheikhly

2 Project Summary: The Multi-University Southeast INIE Consortium (MUSIC) was established in response to the US Department of Energy’s (DOE) Innovations in Nuclear Infrastructure and Education (INIE) program. MUSIC was established as a consortium composed of academic members and national laboratory partners. The members of MUSIC are the nuclear engineering programs and research reactors of Georgia Institute of Technology (GIT), North Carolina State University (NCSU), University of Maryland (UMD), University of South Carolina (USC), and University of Tennessee (UTK). The University of Florida (UF), and South Carolina State University (SCSU) were added to the MUSIC membership in the second year.more » In addition, to ensure proper coordination between the academic community and the nation’s premier research and development centers in the fields of nuclear science and engineering, MUSIC created strategic partnerships with Oak Ridge National Laboratory (ORNL) including the Spallation Neutron Source (SNS) project and the Joint Institute for Neutron Scattering (JINS), and the National Institute of Standards and Technology (NIST). A partnership was also created with the Armed Forces Radiobiology Research Institute (AFRRI) with the aim of utilizing their reactor in research if funding becomes available. Consequently, there are three university research reactors (URRs) within MUSIC, which are located at NCSU (1-MW PULSTAR), UMD (0.25-MW TRIGA) and UF (0.10-MW Argonaut), and the AFRRI reactor (1-MW TRIGA MARK F). The overall objectives of MUSIC are: a) Demonstrate that University Research Reactors (URR) can be used as modern and innovative instruments of research in the basic and applied sciences, which include applications in fundamental physics, materials science and engineering, nondestructive examination, elemental analysis, and contributions to research in the health and medical sciences, b) Establish a strong technical collaboration between the nuclear engineering faculty and the MUSIC URRs. This will be achieved by involving the faculty in the development of state-of-the-art research facilities at the URRs and subsequently, in the utilization of these facilities, c) Facilitate the use of the URRs by the science and engineering faculty within the individual institutions and by the general community of science and engineering, d) Develop a far-reaching educational component that is capable of addressing the needs of the nuclear science and engineering community. Specifically, the aim of this component will be to perform public outreach activities, contribute to the active recruitment of the next generation of nuclear professionals, strengthen the education of nuclear engineering students, and promote nuclear engineering education for minority students.« less

Validation of the MCNP computational model for neutron flux distribution with the neutron activation analysis measurement

NASA Astrophysics Data System (ADS)

Tiyapun, K.; Chimtin, M.; Munsorn, S.; Somchit, S.

2015-05-01

The objective of this work is to demonstrate the method for validating the predication of the calculation methods for neutron flux distribution in the irradiation tubes of TRIGA research reactor (TRR-1/M1) using the MCNP computer code model. The reaction rate using in the experiment includes 27Al(n, α)24Na and 197Au(n, γ)198Au reactions. Aluminium (99.9 wt%) and gold (0.1 wt%) foils and the gold foils covered with cadmium were irradiated in 9 locations in the core referred to as CT, C8, C12, F3, F12, F22, F29, G5, and G33. The experimental results were compared to the calculations performed using MCNP which consisted of the detailed geometrical model of the reactor core. The results from the experimental and calculated normalized reaction rates in the reactor core are in good agreement for both reactions showing that the material and geometrical properties of the reactor core are modelled very well. The results indicated that the difference between the experimental measurements and the calculation of the reactor core using the MCNP geometrical model was below 10%. In conclusion the MCNP computational model which was used to calculate the neutron flux and reaction rate distribution in the reactor core can be used for others reactor core parameters including neutron spectra calculation, dose rate calculation, power peaking factors calculation and optimization of research reactor utilization in the future with the confidence in the accuracy and reliability of the calculation.

77 FR 4807 - Revised Fee Policy for Acceptance of Foreign Research Reactor Spent Nuclear Fuel From High-Income...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-01-31

...This notice announces a change in the fee policy by the Department of Energy (DOE) for receipt and management of spent nuclear fuel (SNF) from foreign research reactors (FRR) containing uranium enriched in the U.S. in countries with high-income economies, as identified in the World Bank Development Report. The fee will increase in three phases (See Table 1) for all future SNF shipments (including Training, Research, Isotopes, General Atomics (TRIGA) from high-income economy countries. The first phase will take effect immediately and the fee will increase from no higher than $3,750 per kg total mass (not heavy metal mass) to $5,625 per kg total mass for SNF containing low enriched uranium (LEU). The second phase will be implemented automatically on January 1, 2014, and the fees will increase from $5,625 per kg total mass to $7,500 per kg total mass for shipments of SNF containing LEU and from no higher than $4,500 per kg total mass to $6,750 per kg total mass for SNF containing highly enriched uranium (HEU). The third phase will be implemented automatically on January 1, 2016, and the fee will increase from $6,750 per kg total mass to $9,000 per kg total mass for shipments of SNF containing HEU. DOE is also implementing a new minimum fee of $200,000 per shipment of any type and amount of eligible SNF to reflect a minimum cost of providing acceptance services. This minimum fee will take effect immediately. In the case where a reactor operator already has a signed and executed contract with DOE, DOE intends to negotiate an equitable adjustment to the fee in accordance with this revised fee policy. Under this revised fee policy, the fee for return of TRIGA fuel will be the same as that of aluminum based fuel. All other aspects of the fee policy are unaffected by this Notice. This is the first fee increase since the fee policy was established in 1996, and will help DOE offset a portion of the increase in operation costs of managing SNF. DOE will continue to pay the costs for shipping, receipt and management of SNF from other than high-income economy countries. All other conditions and policies as previously established for acceptance of FRR SNF will continue to apply. DOE reserves the right to revise the fee policy at any time to respond to changed circumstances. DOE also reserves the right to adjust the fee set in an acceptance contract if there are unique and compelling circumstances that make it in DOE's best interest to do so.

Epithermal neutron beam for BNCT research at Washington State University

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

Venhuizen, J.R.; Nigg, D.W.; Wheeler, F.J.

1999-09-01

Veterinary radiation oncology researchers at the Washington State University (WSU) School of Veterinary Medicine have made major contributions to the understanding of the in-vivo radiobiology of boron neutron capture therapy (BNCT) over the years. Recent attention has been focused on the development of a more convenient and cost-effective local epithermal-neutron beam facility for BNCT research and boronated pharmaceutical screening in large-animal models at WSU. The design of such a facility, to be installed in the thermal column region of the TRIGA research reactor at WSU, was performed in a collaborative effort of SWU and the Idaho National Engineering and Environmentalmore » Laboratory. Construction is now underway.« less

McClellan Nuclear Radiation Center (MNRC) TRIGA reactor: The national organization of test research and training reactors

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

Kiger, Kevin M.

1994-07-01

This year's TRTR conference is being hosted by the McClellan Nuclear Radiation Center. The conference will be held at the Red Lion Hotel in Sacramento, CA. The conference dates are scheduled for October 11-14, 1994. Deadlines for sponsorship commitment and papers have not been set, but are forthcoming. The newly remodeled Red Lion Hotel provides up-to-date conference facilities and one of the most desirable locations for dining, shopping and entertainment in the Sacramento area. While attendees are busy with the conference activities, a spouses program will be available. Although the agenda has not been set, the Sacramento area offers outingsmore » to San Francisco, Pier 39, Ghirardelli Square (famous for their chocolate), and a chance to discover 'El Dorado' in the gold country. Not to forget our own bit of history with visits to 'Old Sacramento and Old Folsom', where antiquities abound, to the world renown train museum and incredible eating establishments. (author)« less

Determining Pu-239 content by resonance transmission analysis using a filtered reactor beam.

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

Klann, R. T.

A novel technique has been developed at Argonne National Laboratory to determine the {sup 239}Pu content in EBR-II blanket elements using resonance transmission analysis (RTA) with a filtered reactor beam. The technique uses cadmium and gadolinium filters along with a {sup 239}Pu fission chamber to isolate the 0.3 eV resonance in {sup 239}Pu. In the energy range from 0.1 to 0.5 eV, the total microscopic cross-section of {sup 239}Pu is significantly larger than the cross-sections of {sup 238}U and {sup 235}U. This large difference in cross-section allows small amounts of {sup 239}Pu to be detected in uranium samples. Tests usingmore » a direct beam from a 250 kW TRIGA reactor have been performed with stacks of depleted uranium and {sup 239}Pu foils. Preliminary measurement results are in good agreement with the predicted results up to about two weight percent of {sup 239}Pu in the sample. In addition, measured {sup 239}Pu masses were in agreement with actual sample masses with uncertainties less than 3.8 percent.« less

Digitized neutron imaging with high spatial resolution at a low power research reactor: I. Analysis of detector performance

NASA Astrophysics Data System (ADS)

Zawisky, M.; Hameed, F.; Dyrnjaja, E.; Springer, J.

2008-03-01

Imaging techniques provide an indispensable tool for investigation of materials. Neutrons, due to their specific properties, offer a unique probe for many aspects of condensed matter. Neutron imaging techniques present a challenging experimental task, especially at a low power research reactor. The Atomic Institute with a 250 kW TRIGA MARK II reactor looks back at a long tradition in neutron imaging. Here we report on the advantages gained in a recent upgrade of the imaging instrument including the acquisition of a thin-plate scintillation detector, a single counting micro-channel plate detector, and an imaging plate detector in combination with a high resolution scanner. We analyze the strengths and limitations of each detector in the field of neutron radiography and tomography, and demonstrate that high resolution digitized imaging down to the 50 μm scale can be accomplished with weak beam intensities of 1.3×10 5 n/cm 2 s, if appropriate measures are taken for the inevitable extension of measurement times. In a separate paper we will present some promising first results from the fields of engineering and geology.

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

A high performance neutron powder diffractometer at 3 MW Triga Mark-II research reactor in Bangladesh

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

Kamal, I., E-mail: imtiaz-kamal26@yahoo.com; Yunus, S. M., E-mail: yunussm11@yahoo.com; Datta, T. K., E-mail: tk-datta4@yahoo.com

2016-07-12

A high performance neutron diffractometer called Savar Neutron Diffractometer (SAND) was built and installed at radial beam port-2 of TRIGA Mark II research reactor at AERE, Savar, Dhaka, Bangladesh. Structural studies of materials are being done by this technique to characterize materials crystallograpohically and magnetically. The micro-structural information obtainable by neutron scattering method is very essential for determining its technological applications. This technique is unique for understanding the magnetic behavior in magnetic materials. Ceramic, steel, electronic and electric industries can be benefited from this facility for improving their products and fabrication process. This instrument consists of a Popovicimonochromator with amore » large linear position sensitive detector array. The monochromator consists of nine blades of perfect single crystal of silicon with 6 mm thickness each. The monochromator design was optimized to provide maximum flux on 3 mm diameter cylindrical sample with a relatively flat angular dependence of resolution. Five different wave lengths can be selected by orienting the crystal at various angles. A sapphire filter was used before the primary collimator to minimize the first neutron. The detector assembly is composed of 15 linear position sensitive proportional counters placed at either 1.1 m or 1.6 m from the sample position and enclosed in a air pad supported high density polythene shield. Position sensing is obtained by charge division using 1-wide NIM position encoding modules (PEM). The PEMs communicate with the host computer via USB. The detector when placed at 1.1 m, subtends 30° (2θ) at each step and covers 120° in 4 steps. When the detector is placed at 1.6 m it subtends 20° at each step and covers 120° in 6 steps. The instrument supports both low and high temperature sample environment. The instrument supports both low and high temperature sample environment. The diffractometer is a state-of-the art technology for diffraction study in our country.« less

A high performance neutron powder diffractometer at 3 MW Triga Mark-II research reactor in Bangladesh

NASA Astrophysics Data System (ADS)

Kamal, I.; Yunus, S. M.; Datta, T. K.; Zakaria, A. K. M.; Das, A. K.; Aktar, S.; Hossain, S.; Berliner, R.; Yelon, W. B.

2016-07-01

A high performance neutron diffractometer called Savar Neutron Diffractometer (SAND) was built and installed at radial beam port-2 of TRIGA Mark II research reactor at AERE, Savar, Dhaka, Bangladesh. Structural studies of materials are being done by this technique to characterize materials crystallograpohically and magnetically. The micro-structural information obtainable by neutron scattering method is very essential for determining its technological applications. This technique is unique for understanding the magnetic behavior in magnetic materials. Ceramic, steel, electronic and electric industries can be benefited from this facility for improving their products and fabrication process. This instrument consists of a Popovicimonochromator with a large linear position sensitive detector array. The monochromator consists of nine blades of perfect single crystal of silicon with 6mm thickness each. The monochromator design was optimized to provide maximum flux on 3mm diameter cylindrical sample with a relatively flat angular dependence of resolution. Five different wave lengths can be selected by orienting the crystal at various angles. A sapphire filter was used before the primary collimator to minimize the first neutron. The detector assembly is composed of 15 linear position sensitive proportional counters placed at either 1.1 m or 1.6 m from the sample position and enclosed in a air pad supported high density polythene shield. Position sensing is obtained by charge division using 1-wide NIM position encoding modules (PEM). The PEMs communicate with the host computer via USB. The detector when placed at 1.1 m, subtends 30˚ (2θ) at each step and covers 120˚ in 4 steps. When the detector is placed at 1.6 m it subtends 20˚ at each step and covers 120˚ in 6 steps. The instrument supports both low and high temperature sample environment. The instrument supports both low and high temperature sample environment. The diffractometer is a state-of-the art technology for diffraction study in our country.

Test of a prototype neutron spectrometer based on diamond detectors in a fast reactor

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

Osipenko, M.; Ripani, M.; Ricco, G.

2015-07-01

A prototype of neutron spectrometer based on diamond detectors has been developed. This prototype consists of a {sup 6}Li neutron converter sandwiched between two CVD diamond crystals. The radiation hardness of the diamond crystals makes it suitable for applications in low power research reactors, while a low sensitivity to gamma rays and low leakage current of the detector permit to reach good energy resolution. A fast coincidence between two crystals is used to reject background. The detector was read out using two different electronic chains connected to it by a few meters of cable. The first chain was based onmore » conventional charge-sensitive amplifiers, the other used a custom fast charge amplifier developed for this purpose. The prototype has been tested at various neutron sources and showed its practicability. In particular, the detector was calibrated in a TRIGA thermal reactor (LENA laboratory, University of Pavia) with neutron fluxes of 10{sup 8} n/cm{sup 2}s and at the 3 MeV D-D monochromatic neutron source named FNG (ENEA, Rome) with neutron fluxes of 10{sup 6} n/cm{sup 2}s. The neutron spectrum measurement was performed at the TAPIRO fast research reactor (ENEA, Casaccia) with fluxes of 10{sup 9} n/cm{sup 2}s. The obtained spectra were compared to Monte Carlo simulations, modeling detector response with MCNP and Geant4. (authors)« less

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

NASA Astrophysics Data System (ADS)

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

2018-01-01

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

The alanine detector in BNCT dosimetry: Dose response in thermal and epithermal neutron fields

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

Schmitz, T., E-mail: schmito@uni-mainz.de; Bassler, N.; Blaickner, M.

Purpose: The response of alanine solid state dosimeters to ionizing radiation strongly depends on particle type and energy. Due to nuclear interactions, neutron fields usually also consist of secondary particles such as photons and protons of diverse energies. Various experiments have been carried out in three different neutron beams to explore the alanine dose response behavior and to validate model predictions. Additionally, application in medical neutron fields for boron neutron capture therapy is discussed. Methods: Alanine detectors have been irradiated in the thermal neutron field of the research reactor TRIGA Mainz, Germany, in five experimental conditions, generating different secondary particlemore » spectra. Further irradiations have been made in the epithermal neutron beams at the research reactors FiR 1 in Helsinki, Finland, and Tsing Hua open pool reactor in HsinChu, Taiwan ROC. Readout has been performed with electron spin resonance spectrometry with reference to an absorbed dose standard in a {sup 60}Co gamma ray beam. Absorbed doses and dose components have been calculated using the Monte Carlo codes FLUKA and MCNP. The relative effectiveness (RE), linking absorbed dose and detector response, has been calculated using the Hansen and Olsen alanine response model. Results: The measured dose response of the alanine detector in the different experiments has been evaluated and compared to model predictions. Therefore, a relative effectiveness has been calculated for each dose component, accounting for its dependence on particle type and energy. Agreement within 5% between model and measurement has been achieved for most irradiated detectors. Significant differences have been observed in response behavior between thermal and epithermal neutron fields, especially regarding dose composition and depth dose curves. The calculated dose components could be verified with the experimental results in the different primary and secondary particle fields. Conclusions: The alanine detector can be used without difficulty in neutron fields. The response has been understood with the model used which includes the relative effectiveness. Results and the corresponding discussion lead to the conclusion that application in neutron fields for medical purpose is limited by its sensitivity but that it is a useful tool as supplement to other detectors and verification of neutron source descriptions.« less

Trace element analysis of soil type collected from the Manjung and central Perak

NASA Astrophysics Data System (ADS)

Azman, Muhammad Azfar; Hamzah, Suhaimi; Rahman, Shamsiah Abdul; Elias, Md Suhaimi; Abdullah, Nazaratul Ashifa; Hashim, Azian; Shukor, Shakirah Abd; Kamaruddin, Ahmad Hasnulhadi Che

2015-04-01

Trace elements in soils primarily originated from their parent materials. Parents' material is the underlying geological material that has been undergone different types of chemical weathering and leaching processes. Soil trace elements concentrations may be increases as a result of continuous input from various human activities, including power generation, agriculture, mining and manufacturing. This paper describes the Neutron Activation Analysis (NAA) method used for the determination of trace elements concentrations in part per million (ppm) present in the terrestrial environment soil in Perak. The data may indicate any contamination of trace elements contributed from human activities in the area. The enrichment factors were used to check if there any contamination due to the human activities (power plants, agricultural, mining, etc.) otherwise the values would serve as a baseline data for future study. The samples were collected from 27 locations of different soil series in the area at two different depths: the top soil (0-15cm) and the sub soil (15-30cm). The collected soil samples were air dried at 60°C and passed through 2 µm sieve. Instrumental Neutron Activation Analysis (NAA) has been used for the determination of trace elements. Samples were activated in the Nuclear Malaysia TRIGA Mark II reactor followed by gamma spectrometric analysis. By activating the stable elements in the samples, the elements can be determined from the intensities of gamma energies emitted by the respected radionuclides.

(US-Japan workshop on fusion neutronics at Osaka University, Osaka, Japan, May 25--26, and visits to Savar, Bangladesh, April 20--May 19, Beijing, China, May 21--24, and Tokai, Japan, May 29--30): Foreign trip report

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

White, J.E.

1989-06-15

This trip was initiated by a request from IAEA for USA expert assistance in Bangladesh. The Bangladesh Atomic Energy Commission (BAEC) had acquired a 3MW TRIGA MARK-II research reactor and their specialist needed help in the development of computer codes and data for the effective utilization and analysis of their reactor. Nuclear data recognized as an international standard was installed on a BAEC computer at Savar. The traveler was invited to China (PRC) to discuss multigroup cross section processing methods used at ORNL. Also, the traveler participated in a US-Japan workshop on fusion neutronics at Osaka University where he discussedmore » the activities of RSIC. An orientation visit to JAERI resulted in the collection of information of potential use in US DOE programs, a better understanding of their research activities, and an opportunity to develop personal relationships with JAERI staff.« less

ORIGEN2 calculations supporting TRIGA irradiated fuel data package

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

Schmittroth, F.A.

ORIGEN2 calculations were performed for TRIGA spent fuel elements from the Hanford Neutron Radiography Facility. The calculations support storage and disposal and results include mass, activity,and decay heat. Comparisons with underwater dose-rate measurements were used to confirm and adjust the calculations.

Decontamination and decommissioning of the Mayaguez (Puerto Rico) facility

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

Jackson, P.K.; Freemerman, R.L.

1989-11-01

On February 6, 1987 the US Department of Energy (DOE) awarded the final phase of the decontamination and decommissioning of the nuclear and reactor facilities at the Center for Energy and Environmental Research (CEER), in Mayaguez, Puerto Rico. Bechtel National, Inc., was made the decontamination and decommissioning (D and D) contractor. The goal of the project was to enable DOE to proceed with release of the CEER facility for use by the University of Puerto Rico, who was the operator. This presentation describes that project and lesson learned during its progress. The CEER facility was established in 1957 as themore » Puerto Rico Nuclear Center, a part of the Atoms for Peace Program. It was a nuclear training and research institution with emphasis on the needs of Latin America. It originally consisted of a 1-megawatt Materials Testing Reactor (MTR), support facilities and research laboratories. After eleven years of operation the MTR was shutdown and defueled. A 2-megawatt TRIGA reactor was installed in 1972 and operated until 1976, when it woo was shutdown. Other radioactive facilities at the center included a 10-watt homogeneous L-77 training reactor, a natural uranium graphite-moderated subcritical assembly, a 200KV particle accelerator, and a 15,000 Ci Co-60 irradiation facility. Support facilities included radiochemistry laboratories, counting rooms and two hot cells. As the emphasis shifted to non-nuclear energy technology a name change resulted in the CEER designation, and plans were started for the decontamination and decommissioning effort.« less

Northrop TRIGA facility decommissioning plan versus actual results

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

Gardner, F.W.

1986-01-01

This paper compares the TRIGA facility decontamination and decommissioning (D and D) plan to the actual results and discusses key areas where operational activities were impacted by the final US Nuclear Regulatory Commission approved D and D plan. A discussion of fuel transport, release criteria, and release survey plans is included.

Multi-step Monte Carlo calculations applied to nuclear reactor instrumentation - source definition and renormalization to physical values

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

Radulovic, Vladimir; Barbot, Loic; Fourmentel, Damien

Significant efforts have been made over the last few years in the French Alternative Energies and Atomic Energy Commission (CEA) to adopt multi-step Monte Carlo calculation schemes in the investigation and interpretation of the response of nuclear reactor instrumentation detectors (e.g. miniature ionization chambers - MICs and self-powered neutron or gamma detectors - SPNDs and SPGDs). The first step consists of the calculation of the primary data, i.e. evaluation of the neutron and gamma flux levels and spectra in the environment where the detector is located, using a computational model of the complete nuclear reactor core and its surroundings. Thesemore » data are subsequently used to define sources for the following calculation steps, in which only a model of the detector under investigation is used. This approach enables calculations with satisfactory statistical uncertainties (of the order of a few %) within regions which are very small in size (the typical volume of which is of the order of 1 mm{sup 3}). The main drawback of a calculation scheme as described above is that perturbation effects on the radiation conditions caused by the detectors themselves are not taken into account. Depending on the detector, the nuclear reactor and the irradiation position, the perturbation in the neutron flux as primary data may reach 10 to 20%. A further issue is whether the model used in the second step calculations yields physically representative results. This is generally not the case, as significant deviations may arise, depending on the source definition. In particular, as presented in the paper, the injudicious use of special options aimed at increasing the computation efficiency (e.g. reflective boundary conditions) may introduce unphysical bias in the calculated flux levels and distortions in the spectral shapes. This paper presents examples of the issues described above related to a case study on the interpretation of the signal from different types of SPNDs, which were recently irradiated in the Jozef Stefan Institute TRIGA Mark II reactor in Ljubljana, Slovenia, and provides recommendations on how they can be overcome. The paper concludes with a discussion on the renormalization of the results from the second step calculations, to obtain accurate physical values. (authors)« less

Concentration of trace elements on branded cigarette in Malaysia

NASA Astrophysics Data System (ADS)

Azman, Muhammad Azfar; Yasir, Muhamad Samudi; Rahman, Irman Abdul; Hamzah, Suhaimi; Rahman, Shamsiah Abdul; Elias, Md Suhaimi; Abdullah, Nazaratul Ashifa; Hashim, Azian; Shukor, Shakirah Abd

2016-01-01

Tobacco is a plant that is used as a recreational drug since the beginning of its use by the Native Americans. Now with the development of the tobacco industry, smoking has become a norm for the public in Malaysia. Trace elements in plants are mostly due to the uptake processes from the soils into the roots of the plants. The concentration of the elements may also be influenced by the elements contained in the water and also fertilizers. This paper aim to analyze the concentration of the trace elements contained in the branded cigarettes sold in Malaysia by utilizing the neutron activation analysis. The tobaccos were taken out from the cigarettes. The collected samples were air dried and passed through 2 mm sieve. Instrumental Neutron Activation Analysis (NAA) has been used for the determination of trace elements. Samples were activated in the Nuclear Malaysia Triga Mark II reactor with a neutron flux of 2.0 x 1012 n cm-2 s-1. The samples then were analyzed using ORTEC Gamma Spectrometer a co-axial n-type HPGe detector with resolution of 2.0 keV at 1332 keV and relative efficiency of 20%. The data obtained could help in assessing the concentration of the trace elements that complying with the standard limitation dose proposed by World Health Organization (WHO).

Trace element analysis of soil type collected from the Manjung and central Perak

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

Azman, Muhammad Azfar, E-mail: m-azfar@nuclearmalaysia.gov.my; Hamzah, Suhaimi; Rahman, Shamsiah Abdul

2015-04-29

Trace elements in soils primarily originated from their parent materials. Parents’ material is the underlying geological material that has been undergone different types of chemical weathering and leaching processes. Soil trace elements concentrations may be increases as a result of continuous input from various human activities, including power generation, agriculture, mining and manufacturing. This paper describes the Neutron Activation Analysis (NAA) method used for the determination of trace elements concentrations in part per million (ppm) present in the terrestrial environment soil in Perak. The data may indicate any contamination of trace elements contributed from human activities in the area. Themore » enrichment factors were used to check if there any contamination due to the human activities (power plants, agricultural, mining, etc.) otherwise the values would serve as a baseline data for future study. The samples were collected from 27 locations of different soil series in the area at two different depths: the top soil (0-15cm) and the sub soil (15-30cm). The collected soil samples were air dried at 60°C and passed through 2 µm sieve. Instrumental Neutron Activation Analysis (NAA) has been used for the determination of trace elements. Samples were activated in the Nuclear Malaysia TRIGA Mark II reactor followed by gamma spectrometric analysis. By activating the stable elements in the samples, the elements can be determined from the intensities of gamma energies emitted by the respected radionuclides.« less

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

Ledney, G.D.; Madonna, G.S.; Elliott, T.B.

When host antimicrobial defenses are severely compromised by radiation or trauma in conjunction with radiation, death from sepsis results. To evaluate therapies for sepsis in radiation casualties, the authors developed models of acquired and induced bacterial infections in irradiated and irradiated-wounded mice. Animals were exposed to either a mixed radiation field of equal proportions of neutrons and gamma rays (n/gamma = 1) from a TRIGA reactor or pure gamma rays from 60 (Co sources). Skin wounds (15% of total body surface area) were inflicted under methoxyflurane anesthesia 1 h after irradiation. In all mice, wounding after irradiation decreased resistance tomore » infection. Treatments with the immunomodulator synthetic trehalose dicorynomycolate (S-TDCM) before or after mixed neutron-gamma irradiation or gamma irradiation increased survival. Therapy with S-TDCM for mice irradiated with either a mixed field or gamma rays increased resistance to Klebsiella pneumoniae-induced infections.« less

Boron absorption imaging in rat lung colon adenocarcinoma metastases

NASA Astrophysics Data System (ADS)

Altieri, S.; Bortolussi, S.; Bruschi, P.; Fossati, F.; Vittor, K.; Nano, R.; Facoetti, A.; Chiari, P.; Bakeine, J.; Clerici, A.; Ferrari, C.; Salvucci, O.

2006-05-01

Given the encouraging results from our previous work on the clinical application of BNCT on non-resectable, chemotherapy resistant liver metastases, we explore the possibility to extend our technique to lung metastases. A fundamental requirement for BNCT is achieving higher 10B concentrations in the metastases compared to those in healthy tissue. For this reason we developed a rat model with lung metastases in order to study the temporal distribution of 10B concentration in tissues and tumoral cells. Rats with induced lung metastases from colon adenocarcinoma were sacrificed two hours after intraperitoneal Boronphenylalanine infusion. The lungs were harvested, frozen in liquid nitrogen and subsequently histological sections underwent neutron autoradiography in the nuclear reactor Triga Mark II, University of Pavia. Our findings demonstrate higher Boron uptake in tumoral nodules compared to healthy lung parenchyma 2 hours after Boronphenylalanine infusion.

The Self-Powered Detector Simulation `MATiSSe' Toolbox applied to SPNDs for severe accident monitoring in PWRs

NASA Astrophysics Data System (ADS)

Barbot, Loïc; Villard, Jean-François; Fourrez, Stéphane; Pichon, Laurent; Makil, Hamid

2018-01-01

In the framework of the French National Research Agency program on nuclear safety and radioprotection, the `DIstributed Sensing for COrium Monitoring and Safety' project aims at developing innovative instrumentation for corium monitoring in case of severe accident in a Pressurized Water nuclear Reactor. Among others, a new under-vessel instrumentation based on Self-Powered Neutron Detectors is developed using a numerical simulation toolbox, named `MATiSSe'. The CEA Instrumentation Sensors and Dosimetry Lab developed MATiSSe since 2010 for Self-Powered Neutron Detectors material selection and geometry design, as well as for their respective partial neutron and gamma sensitivity calculations. MATiSSe is based on a comprehensive model of neutron and gamma interactions which take place in Selfpowered neutron detector components using the MCNP6 Monte Carlo code. As member of the project consortium, the THERMOCOAX SAS Company is currently manufacturing some instrumented pole prototypes to be tested in 2017. The full severe accident monitoring equipment, including the standalone low current acquisition system, will be tested during a joined CEA-THERMOCOAX experimental campaign in some realistic irradiation conditions, in the Slovenian TRIGA Mark II research reactor.

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

Sarta, Jose A.; Castiblanco, Luis A

With cooperation of the International Atomic Energy Agency (IAEA) and the Department of Energy (DOE) of the United States, several calculations and tasks related to the waste disposal of spent MTR fuel enriched nominally to 93% were carried out for the conversion of the IAN-R1 Research Reactor from MTR-HEU fuel to TRIGA-LEU fuel. In order to remove the spent MTR-HEU fuel of the core and store it safely a program was established at the Instituto de Ciencias Nucleares y Energias Alternativas (INEA). This program included training, acquisition of hardware and software, design and construction of a decay pool, transfer ofmore » the spent HEU fuel elements into the decay pool and his final transport to Savannah River in United States. In this paper are presented data of activities calculated for each relevant radionuclide present in spent MTR-HEU fuel elements of the IAN-R1 Research Reactor and the total activity. The total activity calculated takes in consideration contributions of fission, activation and actinides products. The data obtained were the base for shielding calculations for the decay pool concerning the storage of spent MTR-HEU fuel elements and the respective dosimetric evaluations in the transferring operations of fuel elements into the decay pool.« less

TRIGA: Telecommunications Protocol Processing Subsystem Using Reconfigurable Interoperable Gate Arrays

NASA Technical Reports Server (NTRS)

Pang, Jackson; Pingree, Paula J.; Torgerson, J. Leigh

2006-01-01

We present the Telecommunications protocol processing subsystem using Reconfigurable Interoperable Gate Arrays (TRIGA), a novel approach that unifies fault tolerance, error correction coding and interplanetary communication protocol off-loading to implement CCSDS File Delivery Protocol and Datalink layers. The new reconfigurable architecture offers more than one order of magnitude throughput increase while reducing footprint requirements in memory, command and data handling processor utilization, communication system interconnects and power consumption.

Identifying and quantifying short-lived fission products from thermal fission of HEU using portable HPGe detectors

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

Pierson, Bruce D.; Finn, Erin C.; Friese, Judah I.

2013-03-01

Due to the emerging potential for trafficking of special nuclear material, research programs are investigating current capabilities of commercially available portable gamma ray detection systems. Presented in this paper are the results of three different portable high-purity germanium (HPGe) detectors used to identify short-lived fission products generated from thermal neutron interrogation of small samples of highly enriched uranium. Samples were irradiated at the Washington State University (WSU) Nuclear Radiation Center’s 1MW TRIGA reactor. The three portable, HPGe detectors used were the ORTEC MicroDetective, the ORTEC Detective, and the Canberra Falcon. Canberra’s GENIE-2000 software was used to analyze the spectral datamore » collected from each detector. Ultimately, these three portable detectors were able to identify a large range of fission products showing potential for material discrimination.« less

Study of pipe thickness loss using a neutron radiography method

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

Mohamed, Abdul Aziz; Wahab, Aliff Amiru Bin; Yazid, Hafizal B.

2014-02-12

The purpose of this preliminary work is to study for thickness changes in objects using neutron radiography. In doing the project, the technique for the radiography was studied. The experiment was done at NUR-2 facility at TRIGA research reactor in Malaysian Nuclear Agency, Malaysia. Test samples of varying materials were used in this project. The samples were radiographed using direct technique. Radiographic images were recorded using Nitrocellulose film. The films obtained were digitized to processed and analyzed. Digital processing is done on the images using software Isee!. The images were processed to produce better image for analysis. The thickness changesmore » in the image were measured to be compared with real thickness of the objects. From the data collected, percentages difference between measured and real thickness are below than 2%. This is considerably very low variation from original values. Therefore, verifying the neutron radiography technique used in this project.« less

Use of boron nitride for neutron spectrum characterization and cross-section validation in the epithermal range through integral activation measurements

NASA Astrophysics Data System (ADS)

Radulović, Vladimir; Trkov, Andrej; Jaćimović, Radojko; Gregoire, Gilles; Destouches, Christophe

2016-12-01

A recent experimental irradiation and measurement campaign using containers made from boron nitride (BN) at the Jožef Stefan Institute (JSI) TRIGA Mark II reactor in Ljubljana, Slovenia, has shown the applicability of BN for neutron spectrum characterization and cross-section validation in the epithermal range through integral activation measurements. The first part of the paper focuses on the determination of the transmission function of a BN container through Monte Carlo calculations and experimental measurements. The second part presents the process of tayloring the sensitivity of integral activation measurements to specific needs and a selection of suitable radiative capture reactions for neutron spectrum characterization in the epithermal range. A BN container used in our experiments and its qualitative effect on the neutron spectrum in the irradiation position employed is displayed in the Graphical abstract.

Fault detection and analysis in nuclear research facility using artificial intelligence methods

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

Ghazali, Abu Bakar, E-mail: Abakar@uniten.edu.my; Ibrahim, Maslina Mohd

In this article, an online detection of transducer and actuator condition is discussed. A case study is on the reading of area radiation monitor (ARM) installed at the chimney of PUSPATI TRIGA nuclear reactor building, located at Bangi, Malaysia. There are at least five categories of abnormal ARM reading that could happen during the transducer failure, namely either the reading becomes very high, or very low/ zero, or with high fluctuation and noise. Moreover, the reading may be significantly higher or significantly lower as compared to the normal reading. An artificial neural network (ANN) and adaptive neuro-fuzzy inference system (ANFIS)more » are good methods for modeling this plant dynamics. The failure of equipment is based on ARM reading so it is then to compare with the estimated ARM data from ANN/ ANFIS function. The failure categories in either ‘yes’ or ‘no’ state are obtained from a comparison between the actual online data and the estimated output from ANN/ ANFIS function. It is found that this system design can correctly report the condition of ARM equipment in a simulated environment and later be implemented for online monitoring. This approach can also be extended to other transducers, such as the temperature profile of reactor core and also to include other critical actuator conditions such as the valves and pumps in the reactor facility provided that the failure symptoms are clearly defined.« less

Use of higher order signal moments and high speed digital sampling technique for neutron flux measurement

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

Baers, L.B.; Gutierrez, T.R.; Mendoza, R.A.

1993-08-01

The second (conventional variance or Campbell signal) , , , the third , and the modified fourth order [minus] 3*[sup 2] etc. central signal moments associated with the amplified (K) and filtered currents [i[sub 1], i[sub 2], x = K * (i[sub 2]-),] from two electrodes of an ex-core neutron sensitive fission detector have been measured versus the reactor power of the 1 MW TRIGA reactor in Mexico City. Two channels of a high speed (400 kHz) multiplexing data sampler and A/D converter with 12 bit resolution and one megawords buffer memory were used. The data were further retrieved intomore » a PC and estimates for auto- and cross-correlation moments up to the fifth order, coherence (/[radical]), skewness (/([radical]/)[sup 3]), excess (/[sup 2] - 3) etc. quantities were calculated off-line. A five mode operation of the detector was achieved including the conventional counting rates and currents in agreement with the theory and the authors previous results with analogue techniques. The signals were proportional to the neutron flux and reactor power in some flux ranges. The suppression of background noise is improved and the lower limit of the measurement range is extended as the order of moment is increased, in agreement with the theory. On the other hand the statistical uncertainty is increased. At increasing flux levels it was statistically more difficult to obtain flux estimates based on the higher order ([>=]3) moments.« less

Concentration of trace elements on branded cigarette in Malaysia

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

Azman, Muhammad Azfar, E-mail: m-azfar@nuclearmalaysia.gov.my; Hamzah, Suhaimi; Rahman, Shamsiah Abdul

Tobacco is a plant that is used as a recreational drug since the beginning of its use by the Native Americans. Now with the development of the tobacco industry, smoking has become a norm for the public in Malaysia. Trace elements in plants are mostly due to the uptake processes from the soils into the roots of the plants. The concentration of the elements may also be influenced by the elements contained in the water and also fertilizers. This paper aim to analyze the concentration of the trace elements contained in the branded cigarettes sold in Malaysia by utilizing themore » neutron activation analysis. The tobaccos were taken out from the cigarettes. The collected samples were air dried and passed through 2 mm sieve. Instrumental Neutron Activation Analysis (NAA) has been used for the determination of trace elements. Samples were activated in the Nuclear Malaysia Triga Mark II reactor with a neutron flux of 2.0 x 10{sup 12} n cm{sup -2} s{sup -1}. The samples then were analyzed using ORTEC Gamma Spectrometer a co-axial n-type HPGe detector with resolution of 2.0 keV at 1332 keV and relative efficiency of 20%. The data obtained could help in assessing the concentration of the trace elements that complying with the standard limitation dose proposed by World Health Organization (WHO)« less

Gamma spectrometry and chemical characterization of ceramic seeds with samarium-153 and holmium-166 for brachytherapy proposal.

PubMed

Valente, Eduardo S; Campos, Tarcísio P R

2010-12-01

Ceramic seeds were synthesized by the sol-gel technique with Si:Sm:Ca and Si:Ho:Ca. One set of seeds was irradiated in the TRIGA type nuclear reactor IPR-R1 and submitted to instrumental neutron activation analysis (INAA), K(0) method, to determine mass percentage concentration of natural samarium and holmium in the seed as well as to determine all existing radionuclides and their activities. Attention was paid to discrimination of Si-31, Ca-40, Ca-45, Ca-47, Ca-49, Sm-145, Sm-155, Sm-153 and Ho-166. A second sample was submitted to atomic emission spectrometry (ICP-AES) also to determine samarium and holmium concentrations in weight. A third sample was submitted to X-ray fluorescence spectrometry to qualitatively determine chemical composition. The measured activity was due to Sm-153 and Ho-166 with a well-characterized gamma spectrum. The X-ray fluorescence spectrum demonstrated that there is no discrepancy in seed composition. The maximum ranges in the water of beta particles from Sm-153 and Ho-166 decay were evaluated, as well as the dose rate and total dose delivered within the volume delimited by the range of the beta particles. The results are relevant for investigation of the viability of producing Sm-153 and Ho-166 radioactive seeds for use in brachytherapy. Copyright 2010 Elsevier Ltd. All rights reserved.

Thermal neutron flux measurement using self-powered neutron detector (SPND) at out-core locations of TRIGA PUSPATI Reactor (RTP)

NASA Astrophysics Data System (ADS)

Ali, Nur Syazwani Mohd; Hamzah, Khaidzir; Mohamad Idris, Faridah; Hairie Rabir, Mohamad

2018-01-01

The thermal neutron flux measurement has been conducted at the out-core location using self-powered neutron detectors (SPNDs). This work represents the first attempt to study SPNDs as neutron flux sensor for developing the fault detection system (FDS) focusing on neutron flux parameters. The study was conducted to test the reliability of the SPND’s signal by measuring the neutron flux through the interaction between neutrons and emitter materials of the SPNDs. Three SPNDs were used to measure the flux at four different radial locations which located at the fission chamber cylinder, 10cm above graphite reflector, between graphite reflector and tank liner and fuel rack. The measurements were conducted at 750 kW reactor power. The outputs from SPNDs were collected through data acquisition system and were corrected to obtain the actual neutron flux due to delayed responses from SPNDs. The measurements showed that thermal neutron flux between fission chamber location near to the tank liner and fuel rack were between 5.18 × 1011 nv to 8.45 × 109 nv. The average thermal neutron flux showed a good agreement with those from previous studies that has been made using simulation at the same core configuration at the nearest irradiation facilities with detector locations.

Automating High-Precision X-Ray and Neutron Imaging Applications with Robotics

DOE PAGES

Hashem, Joseph Anthony; Pryor, Mitch; Landsberger, Sheldon; ...

2017-03-28

Los Alamos National Laboratory and the University of Texas at Austin recently implemented a robotically controlled nondestructive testing (NDT) system for X-ray and neutron imaging. This system is intended to address the need for accurate measurements for a variety of parts and, be able to track measurement geometry at every imaging location, and is designed for high-throughput applications. This system was deployed in a beam port at a nuclear research reactor and in an operational inspection X-ray bay. The nuclear research reactor system consisted of a precision industrial seven-axis robot, 1.1-MW TRIGA research reactor, and a scintillator-mirror-camera-based imaging system. Themore » X-ray bay system incorporated the same robot, a 225-keV microfocus X-ray source, and a custom flat panel digital detector. The robotic positioning arm is programmable and allows imaging in multiple configurations, including planar, cylindrical, as well as other user defined geometries that provide enhanced engineering evaluation capability. The imaging acquisition device is coupled with the robot for automated image acquisition. The robot can achieve target positional repeatability within 17 μm in the 3-D space. Flexible automation with nondestructive imaging saves costs, reduces dosage, adds imaging techniques, and achieves better quality results in less time. Specifics regarding the robotic system and imaging acquisition and evaluation processes are presented. In conclusion, this paper reviews the comprehensive testing and system evaluation to affirm the feasibility of robotic NDT, presents the system configuration, and reviews results for both X-ray and neutron radiography imaging applications.« less

Automating High-Precision X-Ray and Neutron Imaging Applications with Robotics

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

Hashem, Joseph Anthony; Pryor, Mitch; Landsberger, Sheldon

Los Alamos National Laboratory and the University of Texas at Austin recently implemented a robotically controlled nondestructive testing (NDT) system for X-ray and neutron imaging. This system is intended to address the need for accurate measurements for a variety of parts and, be able to track measurement geometry at every imaging location, and is designed for high-throughput applications. This system was deployed in a beam port at a nuclear research reactor and in an operational inspection X-ray bay. The nuclear research reactor system consisted of a precision industrial seven-axis robot, 1.1-MW TRIGA research reactor, and a scintillator-mirror-camera-based imaging system. Themore » X-ray bay system incorporated the same robot, a 225-keV microfocus X-ray source, and a custom flat panel digital detector. The robotic positioning arm is programmable and allows imaging in multiple configurations, including planar, cylindrical, as well as other user defined geometries that provide enhanced engineering evaluation capability. The imaging acquisition device is coupled with the robot for automated image acquisition. The robot can achieve target positional repeatability within 17 μm in the 3-D space. Flexible automation with nondestructive imaging saves costs, reduces dosage, adds imaging techniques, and achieves better quality results in less time. Specifics regarding the robotic system and imaging acquisition and evaluation processes are presented. In conclusion, this paper reviews the comprehensive testing and system evaluation to affirm the feasibility of robotic NDT, presents the system configuration, and reviews results for both X-ray and neutron radiography imaging applications.« less

How we shipped our flip and standard too

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

Deigl, H.J.; Feltz, D.E.

1984-07-01

This paper highlights the planning and handling activities for the shipment of irradiated TRIGA fuel from Texas A and M University to the Argonne National Lab/West (ANL/West) reactor facility at Idaho Falls, Idaho. Attention is focused on the enormous time spent on the planning and preparations prior to the shipment. The actual handling time at the NSCR for three shipping packages containing a total 51 elements was only 4 days, but, the time spent in planning and preparation exceeded 16 months. The fuel was transferred for shipment without incident - and from a health physics standpoint the exercise went verymore » well. Whole body exposures and hand doses were minimal for such a large undertaking. ANL/West health physicists reported contamination of the lifting devices for the HFIR when they received the cask. These pieces were wipe tested and contamination was found to be less than 200 dpm. If they were contaminated we were extremely fortunate during handling not to contaminate our facility or personnel.« less

Theoretical neutron damage calculations in industrial robotic manipulators used for non-destructive imaging applications

DOE PAGES

Hashem, Joseph; Schneider, Erich; Pryor, Mitch; ...

2017-01-01

Our paper describes how to use MCNP to evaluate the rate of material damage in a robot incurred by exposure to a neutron flux. The example used in this work is that of a robotic manipulator installed in a high intensity, fast, and collimated neutron radiography beam port at the University of Texas at Austin's TRIGA Mark II research reactor. Our effort includes taking robotic technologies and using them to automate non-destructive imaging tasks in nuclear facilities where the robotic manipulator acts as the motion control system for neutron imaging tasks. Simulated radiation tests are used to analyze the radiationmore » damage to the robot. Once the neutron damage is calculated using MCNP, several possible shielding materials are analyzed to determine the most effective way of minimizing the neutron damage. Furthermore, neutron damage predictions provide users the means to simulate geometrical and material changes, thus saving time, money, and energy in determining the optimal setup for a robotic system installed in a radiation environment.« less

Acute Lethality after Fast-Neutron and X-Irradiation of Tribolium confusum

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

Glenn, Norman D.; Ducoff, Howard S.

1976-01-01

The acute lethal effects of fast neutrons and of X-rays on adults and larvae of T. confusum are compared. The time course of mortality of adults of the Oklahoma strain was the same after midlethal doses of neutrons and X-rays, although the neutrons were about twice as effective as X-rays in producing lethality, based on LD 50(35). The neutron RBE for adults of the Ebony mutant strain was also about 2, but that for Oklahoma larvae was about 3.85. Larvae surviving midlethal doses of neutrons showed a tendency toward wing abnormalities and delayed pupation. Dose-fractionation recovery with neutron doses inmore » the midlethal range was not detectable in the adults or in the larvae. A considerable sparing effect of dose fractionation was found in X-irradiated adults. Finally, also presented are techniques for using a beam port of a Triga research reactor for fast-neutron irradiation and a method of neutron and gamma dosimetry.« less

Theoretical neutron damage calculations in industrial robotic manipulators used for non-destructive imaging applications

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

Hashem, Joseph; Schneider, Erich; Pryor, Mitch

Our paper describes how to use MCNP to evaluate the rate of material damage in a robot incurred by exposure to a neutron flux. The example used in this work is that of a robotic manipulator installed in a high intensity, fast, and collimated neutron radiography beam port at the University of Texas at Austin's TRIGA Mark II research reactor. Our effort includes taking robotic technologies and using them to automate non-destructive imaging tasks in nuclear facilities where the robotic manipulator acts as the motion control system for neutron imaging tasks. Simulated radiation tests are used to analyze the radiationmore » damage to the robot. Once the neutron damage is calculated using MCNP, several possible shielding materials are analyzed to determine the most effective way of minimizing the neutron damage. Furthermore, neutron damage predictions provide users the means to simulate geometrical and material changes, thus saving time, money, and energy in determining the optimal setup for a robotic system installed in a radiation environment.« less

Pulsed activation measurement of the Doppler effect of uranium-238 over the temperature range 300 to 3115 K

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

Bhattacharyya, S.K.; Russell, G.J.; Foell, W.K.

The Doppler effect for /sup 235/U-enriched UO/sub 2/ fuel pellets has been measured by the Pulsed Activation Doppler (PAD) technique in a TRIGA reactor. A combination of static electrical preheating and pulsed fission heating during irradiation was used to perform the measurements at temperatures extending from 300 K to the melting point of UO/sub 2/ (3115 K). The /sup 235/U enrichment in the experimental samples investigated ranged from 0.22 to 12 percent by weight. Measurements were made at under partially molten conditions of UO/sub 2/. Two sizes of pellets were used, with nominal surface-to-mass ratio values of 0.63 and 1.08more » cm/sup 2//g, respectively. The experimentally determined values of the Doppler ratio were in good agreement with resonance integral ratios determined from GAROL calculations and extrapolations of the low-temperature Hellstrand correlation.« less

Reliability assessment of MVP-BURN and JENDL-4.0 related to nuclear transmutation of light platinum group elements

NASA Astrophysics Data System (ADS)

Terashima, Atsunori; Nilsson, Mikael; Ozawa, Masaki; Chiba, Satoshi

2017-09-01

The Aprés ORIENT research program, as a concept of advanced nuclear fuel cycle, was initiated in FY2011 aiming at creating stable, highly-valuable elements by nuclear transmutation from ↓ssion products. In order to simulate creation of such elements by (n, γ) reaction succeeded by β- decay in reactors, a continuous-energy Monte Carlo burnup calculation code MVP-BURN was employed. Then, it is one of the most important tasks to con↓rm the reliability of MVP-BURN code and evaluated neutron cross section library. In this study, both an experiment of neutron activation analysis in TRIGA Mark I reactor at University of California, Irvine and the corresponding burnup calculation using MVP-BURN code were performed for validation of the simulation on transmutation of light platinum group elements. Especially, some neutron capture reactions such as 102Ru(n, γ)103Ru, 104Ru(n, γ)105Ru, and 108Pd(n, γ)109Pd were dealt with in this study. From a comparison between the calculation (C) and the experiment (E) about 102Ru(n, γ)103Ru, the deviation (C/E-1) was signi↓cantly large. Then, it is strongly suspected that not MVP-BURN code but the neutron capture cross section of 102Ru belonging to JENDL-4.0 used in this simulation have made the big di↑erence as (C/E-1) >20%.

The Concentration Of Tritium In Urine And Internal Radiation Dose Estimation Of PTNBR Radiation Workers

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

Tjahaja, Poppy Intan; Sukmabuana, Putu; Aisyah, Neneng Nur

2010-12-23

The operation of Triga 2000 reactor in Nuclear Technology Center for Materials and Radiometry (PTNBR BATAN) normally produce tritium radionuclide which is the activation product of deuterium atom in reactor primary cooling water. According to previous monitoring, tritium was detected with the concentration of 8.236{+-}0.677 kBq/L and 1.704{+-}0.046 Bq/L in the primary cooling water and in reactor hall air, respectively. The tritium in reactor hall air chronically can be inhaled by the workers. In this research, tritium content in radiation workers' urine was determined to estimate the internal radiation doses received by the workers. About 50-100 mL of urine samplesmore » were collected from 48 PTNBR workers that is classified as 24 radiation workers and 24 administration staffs as a control. Urine samples of 25 mL were then prepared by active charcoal and KMnO{sub 4} addition and followed with complete distillation. The 2 mL of distillate was added with 13 mL scintillator, shaked vigorously and remained in cool and dark condition for about 24 hours. The tritium in the samples was then measured using liquid scintillation counter (LSC) for 1 hour. From the measurement results it was obtained that the tritium concentration in the urine of radiation workers were in the range of not detected and 5.191 Bq/mL, whereas in the administration staffs the concentration were between not detected and 4.607 Bq/mL. Internally radiation doses were calculated using the tritium concentration data, and it was found the averages about 0.602 {mu}Sv/year and 0.532 {mu}Sv/year for radiation workers and administration staffs, respectively. The doses received by the workers were lower than that of the permissible doses from tritium, i.e. 40 {mu}Sv/year.« less

Tracing footprints of environmental events in tree ring chemistry using neutron activation analysis

NASA Astrophysics Data System (ADS)

Sahin, Dagistan

The aim of this study is to identify environmental effects on tree-ring chemistry. It is known that industrial pollution, volcanic eruptions, dust storms, acid rain and similar events can cause substantial changes in soil chemistry. Establishing whether a particular group of trees is sensitive to these changes in soil environment and registers them in the elemental chemistry of contemporary growth rings is the over-riding goal of any Dendrochemistry research. In this study, elemental concentrations were measured in tree-ring samples of absolutely dated eleven modern forest trees, grown in the Mediterranean region, Turkey, collected and dated by the Malcolm and Carolyn Wiener Laboratory for Aegean and Near Eastern Dendrochronology laboratory at Cornell University. Correlations between measured elemental concentrations in the tree-ring samples were analyzed using statistical tests to answer two questions. Does the current concentration of a particular element depend on any other element within the tree? And, are there any elements showing correlated abnormal concentration changes across the majority of the trees? Based on the detailed analysis results, the low mobility of sodium and bromine, positive correlations between calcium, zinc and manganese, positive correlations between trace elements lanthanum, samarium, antimony, and gold within tree-rings were recognized. Moreover, zinc, lanthanum, samarium and bromine showed strong, positive correlations among the trees and were identified as possible environmental signature elements. New Dendrochemistry information found in this study would be also useful in explaining tree physiology and elemental chemistry in Pinus nigra species grown in Turkey. Elemental concentrations in tree-ring samples were measured using Neutron Activation Analysis (NAA) at the Pennsylvania State University Radiation Science and Engineering Center (RSEC). Through this study, advanced methodologies for methodological, computational and experimental NAA were developed to ensure an acceptable accuracy and certainty in the elemental concentration measurements in tree-ring samples. Two independent analysis methods of NAA were used; the well known k-zero method and a novel method developed in this study, called the Multi-isotope Iterative Westcott (MIW) method. The MIW method uses reaction rate probabilities for a group of isotopes, which can be calculated by a neutronic simulation or measured by experimentation, and determines the representative values for the neutron flux and neutron flux characterization parameters based on Westcott convention. Elemental concentration calculations for standard reference material and tree-ring samples were then performed using the MIW and k-zero analysis methods of the NAA and the results were cross verified. In the computational part of this study, a detailed burnup coupled neutronic simulation was developed to analyze real-time neutronic changes in a TRIGA Mark III reactor core, in this study, the Penn State Breazeale Reactor (PSBR) core. To the best of the author`s knowledge, this is the first burnup coupled neutronic simulation with realistic time steps and full fuel temperature profile for a TRIGA reactor using Monte Carlo Utility for Reactor Evolutions (MURE) code and Monte Carlo Neutral-Particle Code (MCNP) coupling. High fidelity and flexibility in the simulation was aimed to replicate the real core operation through the day. This approach resulted in an enhanced accuracy in neutronic representation of the PSBR core with respect to previous neutronic simulation models for the PSBR core. An important contribution was made in the NAA experimentation practices employed in Dendrochemistry studies at the RSEC. Automated laboratory control and analysis software for NAA measurements in the RSEC Radionuclide Applications Laboratory was developed. Detailed laboratory procedures were written in this study comprising preparation, handling and measurements of tree-ring samples in the Radionuclide Applications Laboratory.

Evaluation of the Telecommunications Protocol Processing Subsystem Using Reconfigurable Interoperable Gate Array

NASA Technical Reports Server (NTRS)

Pang, Jackson; Liddicoat, Albert; Ralston, Jesse; Pingree, Paula

2006-01-01

The current implementation of the Telecommunications Protocol Processing Subsystem Using Reconfigurable Interoperable Gate Arrays (TRIGA) is equipped with CFDP protocol and CCSDS Telemetry and Telecommand framing schemes to replace the CPU intensive software counterpart implementation for reliable deep space communication. We present the hardware/software co-design methodology used to accomplish high data rate throughput. The hardware CFDP protocol stack implementation is then compared against the two recent flight implementations. The results from our experiments show that TRIGA offers more than 3 orders of magnitude throughput improvement with less than one-tenth of the power consumption.

Siting process for disposal site of low level radiactive waste in Thailand

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

Yamkate, P.; Sriyotha, P.; Thiengtrongjit, S.

The radioactive waste in Thailand is composed of low level waste from the application of radioisotopes in medical treatment and industry, the operation of the 2 MW TRIGA Mark III Research Reactor and the production of radioisotopes at OAEP. In addition, the high activity of sealed radiation sources i.e. Cs-137 Co-60 and Ra-226 are also accumulated. Since the volume of treated waste has been gradually increased, the general needs for a repository become apparent. The near surface disposal method has been chosen for this aspect. The feasibility study on the underground disposal site has been done since 1982. The sitemore » selection criteria have been established, consisting of the rejection criteria, the technical performance criteria and the economic criteria. About 50 locations have been picked for consideration and 5 candidate sites have been selected and subsequent investigated. After thoroughly investigation, a definite location in Ratchburi Province, about 180 kilometers southwest of Bangkok, has been selected as the most suitable place for the near surface disposal of radioactive waste in Thailand.« less

Upgrade of the compact neutron spectrometer for high flux environments

NASA Astrophysics Data System (ADS)

Osipenko, M.; Bellucci, A.; Ceriale, V.; Corsini, D.; Gariano, G.; Gatti, F.; Girolami, M.; Minutoli, S.; Panza, F.; Pillon, M.; Ripani, M.; Trucchi, D. M.

2018-03-01

In this paper new version of the 6Li-based neutron spectrometer for high flux environments is described. The new spectrometer was built with commercial single crystal Chemical Vapour Deposition diamonds of electronic grade. These crystals feature better charge collection as well as higher radiation hardness. New metal contacts approaching ohmic conditions were deposited on the diamonds suppressing build-up of space charge observed in the previous prototypes. New passive preamplification of the signal at detector side was implemented to improve its resolution. This preamplification is based on the RF transformer not sensitive to high neutron flux. The compact mechanical design allowed to reduce detector size to a tube of 1 cm diameter and 13 cm long. The spectrometer was tested in the thermal column of TRIGA reactor and at the DD neutron generator. The test results indicate an energy resolution of 300 keV (FWHM), reduced to 72 keV (RMS) excluding energy loss, and coincidence timing resolution of 160 ps (FWHM). The measured data are in agreement with Geant4 simulations except for larger energy loss tail presumably related to imperfections of metal contacts and glue expansion.

On the development of radiation tolerant surveillance camera from consumer-grade components

NASA Astrophysics Data System (ADS)

Klemen, Ambrožič; Luka, Snoj; Lars, Öhlin; Jan, Gunnarsson; Niklas, Barringer

2017-09-01

In this paper an overview on the process of designing a radiation tolerant surveillance camera from consumer grade components and commercially available particle shielding materials is given. This involves utilization of Monte-Carlo particle transport code MCNP6 and ENDF/B-VII.0 nuclear data libraries, as well as testing the physical electrical systems against γ radiation, utilizing JSI TRIGA mk. II fuel elements as a γ-ray sources. A new, aluminum, 20 cm × 20 cm × 30 cm irradiation facility with electrical power and signal wire guide-tube to the reactor platform, was designed and constructed and used for irradiation of large electronic and optical components assemblies with activated fuel elements. Electronic components to be used in the camera were tested against γ-radiation in an independent manner, to determine their radiation tolerance. Several camera designs were proposed and simulated using MCNP, to determine incident particle and dose attenuation factors. Data obtained from the measurements and MCNP simulations will be used to finalize the design of 3 surveillance camera models, with different radiation tolerances.

Static and Dynamic Performance of Newly Developed ITER Relevant Insulation Systems after Neutron Irradiation

NASA Astrophysics Data System (ADS)

Prokopec, R.; Humer, K.; Fillunger, H.; Maix, R. K.; Weber, H. W.

2006-03-01

Fiber reinforced plastics will be used as insulation systems for the superconducting magnet coils of ITER. The fast neutron and gamma radiation environment present at the magnet location will lead to serious material degradation, particularly of the insulation. For this reason, advanced radiation-hard resin systems are of special interest. In this study various R-glass fiber / Kapton reinforced DGEBA epoxy and cyanate ester composites fabricated by the vacuum pressure impregnation method were investigated. All systems were irradiated at ambient temperature (340 K) in the TRIGA reactor (Vienna) to a fast neutron fluence of 1×1022 m-2 (E>0.1 MeV). Short-beam shear and static tensile tests were carried out at 77 K prior to and after irradiation. In addition, tension-tension fatigue measurements were used in order to assess the mechanical performance of the insulation systems under the pulsed operation conditions of ITER. For the cyanate ester based system the influence of interleaving Kapton layers on the static and dynamic material behavior was investigated as well.

Testing and linearity calibration of films of phenol compounds exposed to thermal neutron field for EPR dosimetry.

PubMed

Gallo, S; Panzeca, S; Longo, A; Altieri, S; Bentivoglio, A; Dondi, D; Marconi, R P; Protti, N; Zeffiro, A; Marrale, M

2015-12-01

This paper reports the preliminary results obtained by Electron Paramagnetic Resonance (EPR) measurements on films of IRGANOX® 1076 phenols with and without low content (5% by weight) of gadolinium oxide (Gd2O3) exposed in the thermal column of the Triga Mark II reactor of LENA (Laboratorio Energia Nucleare Applicata) of Pavia (Italy). Thanks to their size, the phenolic films here presented are good devices for the dosimetry of beams with high dose gradient and which require accurate knowledge of the precise dose delivered. The dependence of EPR signal as function of neutron dose was investigated in the fluence range between 10(11) cm(-2) and 10(14) cm(-2). Linearity of EPR response was found and the signal was compared with that of commercial alanine films. Our analysis showed that gadolinium oxide (5% by weight) can enhance the thermal neutron sensitivity more than 18 times. Irradiated dosimetric films of phenolic compound exhibited EPR signal fading of about 4% after 10 days from irradiation. Copyright © 2015 Elsevier Ltd. All rights reserved.

Status of the TRIGA-LASER experiment

NASA Astrophysics Data System (ADS)

Gorges, C.; Kaufmann, S.; Geppert, Ch.; Krämer, J.; Sánchez, R.; Nörtershäuser, W.

2017-11-01

We report on the newly developed control system called TRITON and the new data acquisition called TILDA as well as on improved isotope shift measurements of the isotopes 40,42,44,48Ca in the 4 s 2S1/2 → 4 p 2P3/2 (D2) transition at the TRIGA-LASER experiment in Mainz using collinear laser spectroscopy. Well known isotope shift measurements in the 4 s 2S1/2 → 4 p 2P1/2 (D1) transition act as calibration points to reduce the uncertainties in the D2-line to provide reference values for the determination of nuclear charge radii and quadrupole moments of neutron rich calcium isotopes at COLLAPS.

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.

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

NASA Astrophysics Data System (ADS)

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

2015-11-01

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

Final Report for the “WSU Neutron Capture Therapy Facility Support”

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

Gerald E. Tripard; Keith G. Fox

2006-08-24

The objective for the cooperative research program for which this report has been written was to provide separate NCT facility user support for the students, faculty and scientists who would be doing the U.S. Department of Energy Office (DOE) of Science supported advanced radiotargeted research at the WSU 1 megawatt TRIGA reactor. The participants were the Idaho National laboratory (INL, P.I., Dave Nigg), the Veterinary Medical Research Center of Washington State University (WSU, Janean Fidel and Patrick Gavin), and the Washington State University Nuclear Radiation Center (WSU, P.I., Gerald Tripard). A significant number of DOE supported modifications were made tomore » the WSU reactor in order to create an epithermal neutron beam while at the same time maintaining the other activities of the 1 MW reactor. These modifications were: (1) Removal of the old thermal column. (2) Construction and insertion of a new epithermal filter, collimator and shield. (3) Construction of a shielded room that could accommodate the very high radiation field created by an intense neutron beam. (4) Removal of the previous reactor core fuel cluster arrangement. (5) Design and loading of the new reactor core fuel cluster arrangement in order to optimize the neutron flux entering the epithermal neutron filter. (6) The integration of the shielded rooms interlocks and radiological controls into the SCRAM chain and operating electronics of the reactor. (7) Construction of a motorized mechanism for moving and remotely controlling the position of the entire reactor bridge. (8) The integration of the reactor bridge control electronics into the SCRAM chain and operating electronics of the reactor. (9) The design, construction and attachment to the support structure of the reactor of an irradiation box that could be inserted into position next to the face of the reactor. (Necessitated by the previously mentioned core rearrangement). All of the above modifications were successfully completed and tested. The resulting epithermal beam of 1 x 10{sup 9} n/sec-cm{sup 2} was measured by Idaho National Laboratory with assistance from WSU's Neutron Activation Analysis Group. The beam is as good as our initial proposals for the project had predicted. In addition to all of the design, construction and insertion of the hardware, shielding, electronics and radiation monitoring systems there was considerable manpower and effort put into changes in the Technical Specifications of the reactor and implementing procedures for use of the new facility. This staff involvement is one of the reasons we requested special facility support from the DOE. Once the facility was competed and all of the recalibrations and measurements made to characterize the differences between this reactor core and the previous core we began to assist INL in making their beam measurements with foils and phantoms. Although we proposed support for only one additional staff position to support this new NCT facility the staff support provided by the WSU Nuclear Radiation Center was greater than had been anticipated by our initial proposal. INL was also assisted in the testing of a heavy water (deuterated water) bladder that can be inserted into the collimator in order to produce an intense, external thermal neutron beam. The external epithermal and/or thermal neutron beam capability remains available for use, if funding becomes available for future research projects.« less

Micro-focused Small Angle Neutron Scattering and Imaging for Science and Engineering Using RTP--A Preliminary Study

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

Mohamed, Abdul Aziz; Al Rashid Megat Ahmad, Megat Harun; Md Idris, Faridah

2010-01-05

Malaysian Nuclear Agency's (Nuclear Malaysia) Small Angle Neutron Scattering (SANS) facility--(MYSANS)--is utilizing low flux of thermal neutron at the agency's 1 MW TRIGA reactor. As the design nature of the 8 m SANS facility can allow object resolution in the range between 5 and 80 nm to be obtained. It can be used to study alloys, ceramics and polymers in certain area of problems that relate to samples containing strong scatterers or contrast. The current SANS system at Malaysian Nuclear Agency is only capable to measure Q in limited range with a PSD (128x128) fixed at 4 m from themore » sample. The existing reactor hall that incorporate this MYSANS facility has a layout that prohibits the rebuilding of MYSANS therefore the position between the wavelength selector (HOPG) and sample and the PSD cannot be increased for wider Q range. The flux of the neutron at current sample holder is very low which around 10{sup 3} n/cm{sup 2}/sec. Thus it is important to rebuild the MYSANS to maximize the utilization of neutron. Over the years, the facility has undergone maintenance and some changes have been made. Modification on secondary shutter and control has been carried out to improve the safety level of the instrument. A compact micro-focus SANS method can suit this objective together with an improve cryostat system. This paper will explain some design concept and approaches in achieving higher flux and the modification needs to establish the micro-focused SANS.« less

High yields of hydrogen production from methanol steam reforming with a cross-U type reactor

PubMed Central

Zhang, Shubin; Chen, Junyu; Zhang, Xuelin; Liu, Xiaowei

2017-01-01

This paper presents a numerical and experimental study on the performance of a methanol steam reformer integrated with a hydrogen/air combustion reactor for hydrogen production. A CFD-based 3D model with mass and momentum transport and temperature characteristics is established. The simulation results show that better performance is achieved in the cross-U type reactor compared to either a tubular reactor or a parallel-U type reactor because of more effective heat transfer characteristics. Furthermore, Cu-based micro reformers of both cross-U and parallel-U type reactors are designed, fabricated and tested for experimental validation. Under the same condition for reforming and combustion, the results demonstrate that higher methanol conversion is achievable in cross-U type reactor. However, it is also found in cross-U type reactor that methanol reforming selectivity is the lowest due to the decreased water gas shift reaction under high temperature, thereby carbon monoxide concentration is increased. Furthermore, the reformed gas generated from the reactors is fed into a high temperature proton exchange membrane fuel cell (PEMFC). In the test of discharging for 4 h, the fuel cell fed by cross-U type reactor exhibits the most stable performance. PMID:29121067

High yields of hydrogen production from methanol steam reforming with a cross-U type reactor.

PubMed

Zhang, Shubin; Zhang, Yufeng; Chen, Junyu; Zhang, Xuelin; Liu, Xiaowei

2017-01-01

This paper presents a numerical and experimental study on the performance of a methanol steam reformer integrated with a hydrogen/air combustion reactor for hydrogen production. A CFD-based 3D model with mass and momentum transport and temperature characteristics is established. The simulation results show that better performance is achieved in the cross-U type reactor compared to either a tubular reactor or a parallel-U type reactor because of more effective heat transfer characteristics. Furthermore, Cu-based micro reformers of both cross-U and parallel-U type reactors are designed, fabricated and tested for experimental validation. Under the same condition for reforming and combustion, the results demonstrate that higher methanol conversion is achievable in cross-U type reactor. However, it is also found in cross-U type reactor that methanol reforming selectivity is the lowest due to the decreased water gas shift reaction under high temperature, thereby carbon monoxide concentration is increased. Furthermore, the reformed gas generated from the reactors is fed into a high temperature proton exchange membrane fuel cell (PEMFC). In the test of discharging for 4 h, the fuel cell fed by cross-U type reactor exhibits the most stable performance.

Measurement and simulation of thermal neutron flux distribution in the RTP core

NASA Astrophysics Data System (ADS)

Rabir, Mohamad Hairie B.; Jalal Bayar, Abi Muttaqin B.; Hamzah, Na'im Syauqi B.; Mustafa, Muhammad Khairul Ariff B.; Karim, Julia Bt. Abdul; Zin, Muhammad Rawi B. Mohamed; Ismail, Yahya B.; Hussain, Mohd Huzair B.; Mat Husin, Mat Zin B.; Dan, Roslan B. Md; Ismail, Ahmad Razali B.; Husain, Nurfazila Bt.; Jalil Khan, Zareen Khan B. Abdul; Yakin, Shaiful Rizaide B. Mohd; Saad, Mohamad Fauzi B.; Masood, Zarina Bt.

2018-01-01

The in-core thermal neutron flux distribution was determined using measurement and simulation methods for the Malaysian’s PUSPATI TRIGA Reactor (RTP). In this work, online thermal neutron flux measurement using Self Powered Neutron Detector (SPND) has been performed to verify and validate the computational methods for neutron flux calculation in RTP calculations. The experimental results were used as a validation to the calculations performed with Monte Carlo code MCNP. The detail in-core neutron flux distributions were estimated using MCNP mesh tally method. The neutron flux mapping obtained revealed the heterogeneous configuration of the core. Based on the measurement and simulation, the thermal flux profile peaked at the centre of the core and gradually decreased towards the outer side of the core. The results show a good agreement (relatively) between calculation and measurement where both show the same radial thermal flux profile inside the core: MCNP model over estimation with maximum discrepancy around 20% higher compared to SPND measurement. As our model also predicts well the neutron flux distribution in the core it can be used for the characterization of the full core, that is neutron flux and spectra calculation, dose rate calculations, reaction rate calculations, etc.

Transmutation of actinides in power reactors.

PubMed

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

2005-01-01

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

SPERT Destructive Test - I on Aluminum, Highly Enriched Plate Type Core

ScienceCinema

None

2018-01-16

SPERT - Special Power Excursion Reactor Tests Destructive Test number 1 On Aluminum, Highly Enriched Plate Type Core. A test studying the behavior of the reactor under destructive conditions on a light water moderated pool-type reactor with a plate-type core.

The Prompt Gamma Neutron Activation Analysis Facility at ICN—Pitesti

NASA Astrophysics Data System (ADS)

Bǎrbos, D.; Pǎunoiu, C.; Mladin, M.; Cosma, C.

2008-08-01

PGNAA is a very widely applicable technique for determining the presence and amount of many elements simultaneously in samples ranging in size from micrograms to many grams. PGNAA is characterized by its capability for nondestructive multi-elemental analysis and its ability to analyse elements that cannot be determined by INAA. By means of this PGNAA method we are able to increase the performace of INAA method. A facility has been developed at Institute for Nuclear Research—Piteşti so that the unique features of prompt gamma-ray neutron activation analysis can be used to measure trace and major elements in samples. The facility is linked at the radial neutron beam tube at ACPR-TRIGA reactor. During the PGNAA—facility is in use the ACPR reactor will be operated in steady-state mode at 250 KW maximum power. The facility consists of a radial beam-port, external sample position with shielding, and induced prompt gamma-ray counting system. Thermal neutron flux with energy lower than cadmium cut-off at the sample position was measured using thin gold foil is: φscd = 1.106 n/cm2/s with a cadmium ratio of:80. The gamma-ray detection system consist of an HpGe detector of 16% efficiency (detector model GC1518) with 1.85 keV resolution capability. The HpGe is mounted with its axis at 90° with respect to the incident neutron beam at distance about 200mm from the sample position. To establish the performance capabilities of the facility, irradiation of pure element or sample compound standards were performed to identify the gama-ray energies from each element and their count rates.

Early Years of Neutron Scattering and Its Manpower Development in Indonesia

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

Marsongkohadi

In this paper I shall give a short history of the development of neutron scattering at the Research Centre for Nuclear Techniques (PPTN), in Bandung, and the early development of a more advanced facilities at the Neutron Scattering Laboratory (NSL BATAN), Centre of Technology for Nuclear Industrial Materials, in Serpong. The first research reactor in Indonesia was the TRIGA MARK II in Bandung, which became operational in 1965, with a power of 250 KW, upgraded to 1 MW in 1971, and to 2 MW in 2000. The neutron scattering activities was started in 1967, with the design and construction ofmore » the first powder diffractometer, and put in operation in 1970. It was followed by the second instrument, the filter detector spectrometer built in 1975 in collaboration with the Bhabha Atomic Research Centre (BARC), India. A powder diffractometer for magnetic studies was built in 1980, and finally, a modification of the filter detector spectrometer to measure textures was made in 1986. A brief description of the design and construction of the instruments, and a highlight of some research topics will be presented. Early developments of neutron scattering activities at the 30 MW, RSG-GAS reactor in Serpong in choosing suitable research program, which will be mainly centred around materials testing/characterization, and materials/condensed matter researches has been agreed. Instrument planning and layout which were appropriate to carry out the program had been decided. Manpower development for the neutron scattering laboratory is a severe problem. The efforts to overcome this problem has been solved. International Cooperation through workshops and on the job trainings also support the supply of qualified manpower.« less

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

NASA Astrophysics Data System (ADS)

Permana, Sidik

2017-07-01

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

Dismantling of Loop-Type Channel Equipment of MR Reactor in NRC 'Kurchatov Institute' - 13040

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

Volkov, Victor; Danilovich, Alexey; Zverkov, Yuri

2013-07-01

In 2009 the project of decommissioning of MR and RTF reactors was developed and approved by the Expert Authority of the Russian Federation (Gosexpertiza). The main objective of the decommissioning works identified in this project: - complete dismantling of reactor equipment and systems; - decontamination of reactor premises and site in accordance with the established sanitary and hygienic standards. At the preparatory stage (2008-2010) of the project the following works were executed: loop-type channels' dismantling in the storage pool; experimental fuel assemblies' removal from spent fuel repositories in the central hall; spent fuel assembly removal from the liquid-metal-cooled loop-type channelmore » of the reactor core and its placement into the SNF repository; and reconstruction of engineering support systems to the extent necessary for reactor decommissioning. The project assumes three main phases of dismantling and decontamination: - dismantling of equipment/pipelines of cooling circuits and loop-type channels, and auxiliary reactor equipment (2011-2012); - dismantling of equipment in underground reactor premises and of both MR and RTF in-vessel devices (2013-2014); - decontamination of reactor premises; rehabilitation of the reactor site; final radiation survey of reactor premises, loop-type channels and site; and issuance of the regulatory authorities' de-registration statement (2015). In 2011 the decommissioning license for the two reactors was received and direct MR decommissioning activities started. MR primary pipelines and loop-type facilities situated in the underground reactor hall were dismantled. Works were also launched to dismantle the loop-type channels' equipment in underground reactor premises; reactor buildings were reconstructed to allow removal of dismantled equipment; and the MR/RTF decommissioning sequence was identified. In autumn 2011 - spring 2012 results of dismantling activities performed are: - equipment from underground rooms (No. 66, 66A, 66B, 72, 64, 63) - as well as from water and gas loop corridors - was dismantled, with the total radwaste weight of 53 tons and the total removed activity of 5,0 x 10{sup 10} Bq; - loop-type channel equipment from underground reactor hall premises was dismantled; - 93 loop-type channels were characterized, chopped and removed, with radwaste of 2.6 x 10{sup 13} Bq ({sup 60}Co) and 1.5 x 10{sup 13} Bq ({sup 137}Cs) total activity removed from the reactor pool, fragmented and packaged. Some of this waste was placed into the high-level waste (HLW) repository of the Center. Dismantling works were executed with application of remotely operated mechanisms, which promoted decrease of radiation impact on the personnel. The average individual dose for the personnel was 1.9 mSv/year in 2011, and the collective dose is estimated as 0.0605 man x Sv/year. (authors)« less

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

NASA Technical Reports Server (NTRS)

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

1973-01-01

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

Operating characteristic analysis of a 400 mH class HTS DC reactor in connection with a laboratory scale LCC type HVDC system

NASA Astrophysics Data System (ADS)

Kim, Sung-Kyu; Kim, Kwangmin; Park, Minwon; Yu, In-Keun; Lee, Sangjin

2015-11-01

High temperature superconducting (HTS) devices are being developed due to their advantages. Most line commutated converter based high voltage direct current (HVDC) transmission systems for long-distance transmission require large inductance of DC reactor; however, generally, copper-based reactors cause a lot of electrical losses during the system operation. This is driving researchers to develop a new type of DC reactor using HTS wire. The authors have developed a 400 mH class HTS DC reactor and a laboratory scale test-bed for line-commutated converter type HVDC system and applied the HTS DC reactor to the HVDC system to investigate their operating characteristics. The 400 mH class HTS DC reactor is designed using a toroid type magnet. The HVDC system is designed in the form of a mono-pole system with thyristor-based 12-pulse power converters. In this paper, the investigation results of the HTS DC reactor in connection with the HVDC system are described. The operating characteristics of the HTS DC reactor are analyzed under various operating conditions of the system. Through the results, applicability of an HTS DC reactor in an HVDC system is discussed in detail.

The development of a thermal hydraulic feedback mechanism with a quasi-fixed point iteration scheme for control rod position modeling for the TRIGSIMS-TH application

NASA Astrophysics Data System (ADS)

Karriem, Veronica V.

Nuclear reactor design incorporates the study and application of nuclear physics, nuclear thermal hydraulic and nuclear safety. Theoretical models and numerical methods implemented in computer programs are utilized to analyze and design nuclear reactors. The focus of this PhD study's is the development of an advanced high-fidelity multi-physics code system to perform reactor core analysis for design and safety evaluations of research TRIGA-type reactors. The fuel management and design code system TRIGSIMS was further developed to fulfill the function of a reactor design and analysis code system for the Pennsylvania State Breazeale Reactor (PSBR). TRIGSIMS, which is currently in use at the PSBR, is a fuel management tool, which incorporates the depletion code ORIGEN-S (part of SCALE system) and the Monte Carlo neutronics solver MCNP. The diffusion theory code ADMARC-H is used within TRIGSIMS to accelerate the MCNP calculations. It manages the data and fuel isotopic content and stores it for future burnup calculations. The contribution of this work is the development of an improved version of TRIGSIMS, named TRIGSIMS-TH. TRIGSIMS-TH incorporates a thermal hydraulic module based on the advanced sub-channel code COBRA-TF (CTF). CTF provides the temperature feedback needed in the multi-physics calculations as well as the thermal hydraulics modeling capability of the reactor core. The temperature feedback model is using the CTF-provided local moderator and fuel temperatures for the cross-section modeling for ADMARC-H and MCNP calculations. To perform efficient critical control rod calculations, a methodology for applying a control rod position was implemented in TRIGSIMS-TH, making this code system a modeling and design tool for future core loadings. The new TRIGSIMS-TH is a computer program that interlinks various other functional reactor analysis tools. It consists of the MCNP5, ADMARC-H, ORIGEN-S, and CTF. CTF was coupled with both MCNP and ADMARC-H to provide the heterogeneous temperature distribution throughout the core. Each of these codes is written in its own computer language performing its function and outputs a set of data. TRIGSIMS-TH provides an effective use and data manipulation and transfer between different codes. With the implementation of feedback and control- rod-position modeling methodologies, the TRIGSIMS-TH calculations are more accurate and in a better agreement with measured data. The PSBR is unique in many ways and there are no "off-the-shelf" codes, which can model this design in its entirety. In particular, PSBR has an open core design, which is cooled by natural convection. Combining several codes into a unique system brings many challenges. It also requires substantial knowledge of both operation and core design of the PSBR. This reactor is in operation decades and there is a fair amount of studies and developments in both PSBR thermal hydraulics and neutronics. Measured data is also available for various core loadings and can be used for validation activities. The previous studies and developments in PSBR modeling also aids as a guide to assess the findings of the work herein. In order to incorporate new methods and codes into exiting TRIGSIMS, a re-evaluation of various components of the code was performed to assure the accuracy and efficiency of the existing CTF/MCNP5/ADMARC-H multi-physics coupling. A new set of ADMARC-H diffusion coefficients and cross sections was generated using the SERPENT code. This was needed as the previous data was not generated with thermal hydraulic feedback and the ARO position was used as the critical rod position. The B4C was re-evaluated for this update. The data exchange between ADMARC-H and MCNP5 was modified. The basic core model is given a flexibility to allow for various changes within the core model, and this feature was implemented in TRIGSIMS-TH. The PSBR core in the new code model can be expanded and changed. This allows the new code to be used as a modeling tool for design and analyses of future code loadings.

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.

Rocky Flats CAAS System Recalibrated, Retested, and Analyzed to Install in the Criticality Experiments Facility at the Nevada Test Site

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

Kim, S; Heinrichs, D; Biswas, D

2009-05-27

Neutron detectors and control panels transferred from the Rocky Flats Plant (RFP) were recalibrated and retested for redeployment to the CEF. Testing and calibration were successful with no failure to any equipment. Detector sensitivity was tested at a TRIGA reactor, and the response to thermal neutron flux was satisfactory. MCNP calculated minimum fission yield ({approx} 2 x 10{sup 15} fissions) was applied to determine the thermal flux at selected detector positions at the CEF. Thermal flux levels were greater than 6.39 x 10{sup 6} (n/cm{sup 2}-sec), which was about four orders of magnitude greater than the minimum alarm flux. Calculationsmore » of detector survivable distances indicate that, to be out of lethal area, a detector needs to be placed greater than 15 ft away from a maximum credible source. MCNP calculated flux/dose results were independently verified by COG. CAAS calibration and the testing confirmed that the RFP CAAS system is performing its functions as expected. New criteria for the CAAS detector placement and 12-rad zone boundaries at the CEF are established. All of the CAAS related documents and hardware have been transferred from LLNL to NSTec for installation at the CEF high bay areas.« less

Neutron activation determination of iridium, gold, platinum, and silver in geologic samples

USGS Publications Warehouse

Millard, H.T.

1987-01-01

Low-level methods for the determination of iridium and other noble metals have become increasingly important in recent years due to interest in locating abundance anomalies associated with the Cretaceous and Tertiary (K-T) boundary. Typical iridium anomalies are in the range of 1 to 100 ??g/kg (ppb). Thus methods with detection limits near 0.1 ??g/kg should be adequate to detect K-T boundary anomalies. Radiochemical neutron activation analysis methods continue to be required although instrumental neutron activation analysis techniques employing elaborate gamma-counters are under development. In the procedure developed in this study samples irradiated in the epithermal neutron facility of the U. S. Geological Survey TRIGA Reactor (Denver, Colorado) are treated with a mini-fire assay technique. The iridium, gold, and silver are collected in a 1-gram metallic lead button. Primary contaminants at this stage are arsenic and antimony. These can be removed by heating the button with a mixture of sodium perioxide and sodium hydroxide. The resulting 0.2-gram lead bead is counted in a Compton suppression spectrometer. Carrier yields are determined by reirradiation of the lead beads. This procedure has been applied to the U.S.G.S. Standard Rock PCC-1 and samples from K-T boundary sites in the Western Interior of North America. ?? 1987 Akade??miai Kiado??.

PRex: An Experiment to Investigate Detection of Near-field Particulate Deposition from a Simulated Underground Nuclear Weapons Test Vent

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

Keillor, Martin E.; Arrigo, Leah M.; Baciak, James E.

2016-01-01

An experiment to release radioactive particles representative of small-scale venting from an underground nuclear test was conducted to gather data in support of treaty capability development activities. For this experiment, a CO2-driven “air cannon” was used to release lanthanum-140 at ambient temperatures. The radioisotope 140La was chosen as a representative fission fragment with a short half-life and prominent gamma-ray emissions; the choice was also influenced by the successful production and use of 140La with low levels of radioactive contaminants in the Defence Research and Development Canada Field Trials. The source was created through activation of high-purity lanthanum oxide at themore » TRIGA research reactor of Washington State University, Pullman, Washington. Witness plates and air samplers were laid out in an irregular grid covering the area for which the plume was anticipated to deposit based on climatological wind records. A vehicle-mounted spectrometer, and handheld and backpack instruments ranging from polyvinyl toluene to high purity germanium, were used to survey the plume. Additionally, three soil sampling techniques were investigated. The relative sensitivity and utility of sampling and survey methods are discussed in the context of On-Site Inspection.« less

Rare earth elements in core marine sediments of coastal East Malaysia by instrumental neutron activation analysis.

PubMed

Ashraf, Ahmadreza; Saion, Elias; Gharibshahi, Elham; Mohamed Kamari, Halimah; Chee Kong, Yap; Suhaimi Hamzah, Mohd; Suhaimi Elias, Md

2016-01-01

A study was carried out on the concentration of REEs (Dy, Sm, Eu,Yb, Lu, La and Ce) that are present in the core marine sediments of East Malaysia from three locations at South China Sea and one location each at Sulu Sea and Sulawesi Sea. The sediment samples were collected at a depth of between 49 and 109 m, dried, and crushed to powdery form. The entire core sediments prepared for Instrumental Neutron Activation Analysis (INAA) were weighted approximately 0.0500 g to 0.1000 g for short irradiation and 0.1500 g to 0.2000 g for long irradiation. The samples were irradiated with a thermal neutron flux of 4.0×10(12) cm(-2) s(-1) in a TRIGA Mark II research reactor operated at 750 kW. Blank samples and standard reference materials SL-1 were also irradiated for calibration and quality control purposes. It was found that the concentration of REEs varies in the range from 0.11 to 36.84 mg/kg. The chondrite-normalized REEs for different stations suggest that all the REEs are from similar origins. There was no significant REEs contamination as the enrichment factors normalized for Fe fall in the range of 0.42-2.82. Copyright © 2015 Elsevier Ltd. All rights reserved.

135Xe measurements with a two-element CZT-based radioxenon detector for nuclear explosion monitoring.

PubMed

Ranjbar, Lily; Farsoni, Abi T; Becker, Eric M

2017-04-01

Measurement of elevated concentrations of xenon radioisotopes ( 131m Xe, 133m Xe, 133 Xe and 135 Xe) in the atmosphere has been shown to be a very powerful method for verifying whether or not a detected explosion is nuclear in nature. These isotopes are among the few with enough mobility and with half-lives long enough to make their detection at long distances realistic. Existing radioxenon detection systems used by the Comprehensive Nuclear-Test-Ban Treaty Organization (CTBTO) suffer from problems such as complexity, need for high maintenance and memory effect. To study the response of CdZnTe (CZT) detectors to xenon radioisotopes and investigate whether it is capable of mitigating the aforementioned issues with the current radioxenon detection systems, a prototype detector utilizing two coplanar CZT detectors was built and tested at Oregon State University. The detection system measures xenon radioisotopes through beta-gamma coincidence technique by detecting coincidence events between the two detectors. In this paper, we introduce the detector design and report our measurement results with radioactive lab sources and 135 Xe produced in the OSU TRIGA reactor. Minimum Detectable Concentration (MDC) for 135 Xe was calculated to be 1.47 ± 0.05 mBq/m 3 . Copyright © 2017 Elsevier Ltd. All rights reserved.

Study on relationship of performance shaping factor in human error probability with prevalent stress of PUSPATI TRIGA reactor operators

NASA Astrophysics Data System (ADS)

Rahim, Ahmad Nabil Bin Ab; Mohamed, Faizal; Farid, Mohd Fairus Abdul; Fazli Zakaria, Mohd; Sangau Ligam, Alfred; Ramli, Nurhayati Binti

2018-01-01

Human factor can be affected by prevalence stress measured using Depression, Anxiety and Stress Scale (DASS). From the respondents feedback can be summarized that the main factor causes the highest prevalence stress is due to the working conditions that require operators to handle critical situation and make a prompt critical decisions. The relationship between the prevalence stress and performance shaping factors found that PSFFitness and PSFWork Process showed positive Pearson’s Correlation with the score of .763 and .826 while the level of significance, p = .028 and p = .012. These positive correlations with good significant values between prevalence stress and human performance shaping factor (PSF) related to fitness, work processes and procedures. The higher the stress level of the respondents, the higher the score of selected for the PSFs. This is due to the higher levels of stress lead to deteriorating physical health and cognitive also worsened. In addition, the lack of understanding in the work procedures can also be a factor that causes a growing stress. The higher these values will lead to the higher the probabilities of human error occur. Thus, monitoring the level of stress among operators RTP is important to ensure the safety of RTP.

Evolution of the collective radiation dose of nuclear reactors from the 2nd through to the 3rd generation and 4th generation sodium-cooled fast reactors

NASA Astrophysics Data System (ADS)

Guidez, Joel; Saturnin, Anne

2017-11-01

During the operation of a nuclear reactor, the external individual doses received by the personnel are measured and recorded, in conformity with the regulations in force. The sum of these measurements enables an evaluation of the annual collective dose expressed in man·Sv/year. This information is a useful tool when comparing the different design types and reactors. This article discusses the evolution of the collective dose for several types of reactors, mainly based on publications from the NEA and the IAEA. The spread of good practices (optimization of working conditions and of the organization, sharing of lessons learned, etc.) and ongoing improvements in reactor design have meant that over time, the doses of various origins received by the personnel have decreased. In the case of sodium-cooled fast reactors (SFRs), the compilation and summarizing of various documentary resources has enabled them to be situated and compared to other types of reactors of the second and third generations (respectively pressurized water reactors in operation and EPR under construction). From these results, it can be seen that the doses received during the operation of SFR are significantly lower for this type of reactor.

77 FR 16868 - Quality Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-03-22

... Fuel Elements for Use in Research and Test Reactors AGENCY: Nuclear Regulatory Commission. ACTION... Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test Reactors.'' This guide describes... plate-type uranium-aluminum fuel elements used in research and test reactors (RTRs). DATES: Submit...

78 FR 25481 - Investigations Regarding Eligibility To Apply for Worker Adjustment Assistance

Federal Register 2010, 2011, 2012, 2013, 2014

2013-05-01

............ 03/25/13 03/12/13 (Company). 82591 CIBA Vision (State/One-Stop) Des Plaines, IL........ 03/25/13 03... Processing (State/One-Stop). 82593 Matheson Tri-Gas (Workers).. Newark, CA 03/26/13 03/14/13 82594 BioTec...

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

Biofilm reactors for industrial bioconversion processes: employing potential of enhanced reaction rates

PubMed Central

Qureshi, Nasib; Annous, Bassam A; Ezeji, Thaddeus C; Karcher, Patrick; Maddox, Ian S

2005-01-01

This article describes the use of biofilm reactors for the production of various chemicals by fermentation and wastewater treatment. Biofilm formation is a natural process where microbial cells attach to the support (adsorbent) or form flocs/aggregates (also called granules) without use of chemicals and form thick layers of cells known as "biofilms." As a result of biofilm formation, cell densities in the reactor increase and cell concentrations as high as 74 gL-1 can be achieved. The reactor configurations can be as simple as a batch reactor, continuous stirred tank reactor (CSTR), packed bed reactor (PBR), fluidized bed reactor (FBR), airlift reactor (ALR), upflow anaerobic sludge blanket (UASB) reactor, or any other suitable configuration. In UASB granular biofilm particles are used. This article demonstrates that reactor productivities in these reactors have been superior to any other reactor types. This article describes production of ethanol, butanol, lactic acid, acetic acid/vinegar, succinic acid, and fumaric acid in addition to wastewater treatment in the biofilm reactors. As the title suggests, biofilm reactors have high potential to be employed in biotechnology/bioconversion industry for viable economic reasons. In this article, various reactor types have been compared for the above bioconversion processes. PMID:16122390

75 FR 36717 - Washington State University; Notice of Acceptance for Docketing and Opportunity for Hearing on...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-06-28

... University (the licensee, WSU) to operate the Washington State University Modified TRIGA Nuclear Radiation... NUCLEAR REGULATORY COMMISSION [Docket No. 50-27; Facility Operating License No. R-76; NRC-2010... Safeguards Information and Sensitive Unclassified Non-Safeguards Information AGENCY: Nuclear Regulatory...

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

Zdarek, J.; Pecinka, L.

Leak-before-break (LBB) analysis of WWER type reactors in the Czech and Sloval Republics is summarized in this paper. Legislative bases, required procedures, and validation and verification of procedures are discussed. A list of significant issues identified during the application of LBB analysis is presented. The results of statistical evaluation of crack length characteristics are presented and compared for the WWER 440 Type 230 and 213 reactors and for the WWER 1000 Type 302, 320 and 338 reactors.

High Purity and Yield of Boron Nitride Nanotubes Using Amorphous Boron and a Nozzle-Type Reactor

PubMed Central

Kim, Jaewoo; Seo, Duckbong; Yoo, Jeseung; Jeong, Wanseop; Seo, Young-Soo; Kim, Jaeyong

2014-01-01

Enhancement of the production yield of boron nitride nanotubes (BNNTs) with high purity was achieved using an amorphous boron-based precursor and a nozzle-type reactor. Use of a mixture of amorphous boron and Fe decreases the milling time for the preparation of the precursor for BNNTs synthesis, as well as the Fe impurity contained in the B/Fe interdiffused precursor nanoparticles by using a simple purification process. We also explored a nozzle-type reactor that increased the production yield of BNNTs compared to a conventional flow-through reactor. By using a nozzle-type reactor with amorphous boron-based precursor, the weight of the BNNTs sample after annealing was increased as much as 2.5-times with much less impurities compared to the case for the flow-through reactor with the crystalline boron-based precursor. Under the same experimental conditions, the yield and quantity of BNNTs were estimated as much as ~70% and ~1.15 g/batch for the former, while they are ~54% and 0.78 g/batch for the latter. PMID:28788161

High Purity and Yield of Boron Nitride Nanotubes Using Amorphous Boron and a Nozzle-Type Reactor.

PubMed

Kim, Jaewoo; Seo, Duckbong; Yoo, Jeseung; Jeong, Wanseop; Seo, Young-Soo; Kim, Jaeyong

2014-08-11

Enhancement of the production yield of boron nitride nanotubes (BNNTs) with high purity was achieved using an amorphous boron-based precursor and a nozzle-type reactor. Use of a mixture of amorphous boron and Fe decreases the milling time for the preparation of the precursor for BNNTs synthesis, as well as the Fe impurity contained in the B/Fe interdiffused precursor nanoparticles by using a simple purification process. We also explored a nozzle-type reactor that increased the production yield of BNNTs compared to a conventional flow-through reactor. By using a nozzle-type reactor with amorphous boron-based precursor, the weight of the BNNTs sample after annealing was increased as much as 2.5-times with much less impurities compared to the case for the flow-through reactor with the crystalline boron-based precursor. Under the same experimental conditions, the yield and quantity of BNNTs were estimated as much as ~70% and ~1.15 g/batch for the former, while they are ~54% and 0.78 g/batch for the latter.

Multi-Megawatt Space Nuclear Power Generation

DTIC Science & Technology

1993-06-28

electric generation, both for open- and closed-cycle opera- tion. These reactors use the particulate fuel of the type developed for HTGR reactors. What...commercial HTGR power reactors, the particles are held in place and directly cooled. Figure 2.7 shows the two types of fuel particles developed for...of MW(e), for pulsed energy devices. The FBR would use HTGR -type particle fuel , contained in a annular bed be- tween two porous frits. Helium would

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

Pre-Licensing Evaluation of Legacy SFR Metallic Fuel Data

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

Yacout, A. M.; Billone, M. C.

2016-09-16

The US sodium cooled fast reactor (SFR) metallic fuel performance data that are of interest to advanced fast reactors applications, can be attributed mostly to the Integral Fast Reactor (IFR) program between 1984 and 1994. Metallic fuel data collected prior to the IFR program were associated with types of fuel that are not of interest to future advanced reactors deployment (e.g., previous U-Fissium alloy fuel). The IFR fuels data were collected from irradiation of U-Zr based fuel alloy, with and without Pu additions, and clad in different types of steels, including HT9, D9, and 316 stainless-steel. Different types of datamore » were generated during the program, and were based on the requirements associated with the DOE Advanced Liquid Metal Cooled Reactor (ALMR) program.« less

Elemental analysis of granite by instrumental neutron activation analysis (INAA) and X-ray fluorescence analysis (XRF).

PubMed

El-Taher, A

2012-01-01

The instrumental neutron activation analysis technique (INAA) was used for qualitative and quantitative analysis of granite samples collected from four locations in the Aswan area in South Egypt. The samples were prepared together with their standards and simultaneously irradiated in a neutron flux of 7×10(11)n/cm(2)s in the TRIGA Mainz research reactor. Gamma-ray spectra from an hyper-pure germanium detector were analyzed. The present study provides the basic data of elemental concentrations of granite rocks. The following elements have been determined Na, Mg, K, Fe, Mn, Sc, Cr, Ti, Co, Zn, Ga, Rb, Zr, Nb, Sn, Ba, Cs, La, Ce, Nd, Sm, Eu, Yb, Lu, Hf, Ta, Th and U. The X-ray fluorescence (XRF) was used for comparison and to detect elements, which can be detected only by XRF such as F, S, Cl, Co, Cu, Mo, Ni, Pb, Se and V. The data presented here are our contribution to understanding the elemental composition of the granite rocks. Because there are no existing databases for the elemental analysis of granite, our results are a start to establishing a database for the Egyptian granite. It is hoped that the data presented here will be useful to those dealing with geochemistry, granite chemistry and related fields. Copyright © 2011 Elsevier Ltd. All rights reserved.

Investigations of Au-198 as radiotracer in laboratory porous media using gamma camera: a preliminary study

NASA Astrophysics Data System (ADS)

Othman, N.; Kamal, W. H. B. Wan; Yusof, N. H.; Engku Chik, E. M. F.; Yunos, M. A. S.; Adnan, M. A. K.; Shari, M. R.

2018-01-01

Preliminary experiment has been carried out using irradiated Au-198 as radiotracer inside the laboratory porous media. The objectives are to check the compatibility of Au-198 as the radiotracer inside the porous media as well as to provide insights of fluid hydrodynamics inside the media using gamma camera.198Au is gamma emitter isotope with half-life of 2.7 days and energy of 0.41 MeV (99%). The porous media consists of fine sandstone with grain size 850μm, lubricant as the mimic of original oil in plant (OOIP) or trapped oil and a layer of cement on top of the rig as the bed rock. Gamma camera is arranged next to the porous media in order to capture the movement of radiotracer which has been set to 1minute per frame. Initially, the gold wire which has isotope of 197Au was irradiated inside the rotary rack of Reactor Triga PUSPATI (RTP) to produce 198Au. RTP is located in Nuclear Malaysia, Bangi has energy of 750kW and neutron flux of 5 × 102 n/cm2/s. 198Au, which is in liquid form, is injected inside the porous media and monitored and recorded by gamma camera. The gamma camera gives a quantitative determination of local fluid saturations over the area of observation.

66.7-keV γ -line intensity of 171Tm determined via neutron activation

NASA Astrophysics Data System (ADS)

Weigand, M.; Heftrich, T.; Düllmann, Ch. E.; Eberhardt, K.; Fiebiger, S.; Glorius, J.; Göbel, K.; Haas, R.; Langer, C.; Lohse, S.; Reifarth, R.; Renisch, D.; Wolf, C.

2018-03-01

Background: About 50% of the heavy elements are produced in stars during the slow neutron capture process. The analysis of branching points allows to set constraints on the temperature and the neutron density in the interior of stars. The temperature dependence of the branch point 171Tm is weak. Hence, the 171Tm neutron capture cross section can be used to constrain the neutron density during the main component of the s process in thermally pulsing asymptotic giant branch stars. Purpose: In order to perform neutron capture experiments on 171Tm, sample material has to be produced and characterized. The characterization is done by γ spectroscopy, relying on the intensities of the involved γ lines. Only the 66.7-keV γ line can be observed whose intensity was uncertain so far. Method: An enriched 170Er sample was activated with thermal neutrons at the TRIGA (Training, Research, Isotopes, General Atomics) research reactor at the Johannes Gutenberg-Universität Mainz. The activation resulted in an easily quantifiable number of 171Er nuclei that subsequently decayed to 171Tm. Result: The intensity of the 66.7-keV γ line of the 171Tm decay was measured to Iγ=(0.144 ±0.010 )% . Conclusions: Our result is in good agreement with the value found in the literature.

Root-cause analysis of the better performance of the coarse-mesh finite-difference method for CANDU-type reactors

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

Shen, W.

2012-07-01

Recent assessment results indicate that the coarse-mesh finite-difference method (FDM) gives consistently smaller percent differences in channel powers than the fine-mesh FDM when compared to the reference MCNP solution for CANDU-type reactors. However, there is an impression that the fine-mesh FDM should always give more accurate results than the coarse-mesh FDM in theory. To answer the question if the better performance of the coarse-mesh FDM for CANDU-type reactors was just a coincidence (cancellation of errors) or caused by the use of heavy water or the use of lattice-homogenized cross sections for the cluster fuel geometry in the diffusion calculation, threemore » benchmark problems were set up with three different fuel lattices: CANDU, HWR and PWR. These benchmark problems were then used to analyze the root cause of the better performance of the coarse-mesh FDM for CANDU-type reactors. The analyses confirm that the better performance of the coarse-mesh FDM for CANDU-type reactors is mainly caused by the use of lattice-homogenized cross sections for the sub-meshes of the cluster fuel geometry in the diffusion calculation. Based on the analyses, it is recommended to use 2 x 2 coarse-mesh FDM to analyze CANDU-type reactors when lattice-homogenized cross sections are used in the core analysis. (authors)« less

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

ERIC Educational Resources Information Center

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

2017-01-01

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

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

PubMed

Seyfi, Somayye; Gharib, Morteza

2015-10-01

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

Development of a neutronics calculation method for designing commercial type Japanese sodium-cooled fast reactor

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

Takeda, T.; Shimazu, Y.; Hibi, K.

2012-07-01

Under the R and D project to improve the modeling accuracy for the design of fast breeder reactors the authors are developing a neutronics calculation method for designing a large commercial type sodium- cooled fast reactor. The calculation method is established by taking into account the special features of the reactor such as the use of annular fuel pellet, inner duct tube in large fuel assemblies, large core. The Verification and Validation, and Uncertainty Qualification (V and V and UQ) of the calculation method is being performed by using measured data from the prototype FBR Monju. The results of thismore » project will be used in the design and analysis of the commercial type demonstration FBR, known as the Japanese Sodium fast Reactor (JSFR). (authors)« less

A preliminary user-friendly, digital console for the control room parameters supervision in old-generation Nuclear Plants

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

Memmi, F.; Falconi, L.; Cappelli, M.

2012-07-01

Improvements in the awareness of a system status is an essential requirement to achieve safety in every kind of plant. In particular, in the case of Nuclear Power Plants (NPPs), a progress is crucial to enhance the Human Machine Interface (HMI) in order to optimize monitoring and analyzing processes of NPP operational states. Firstly, as old-fashioned plants are concerned, an upgrading of the whole console instrumentation is desirable in order to replace an analog visualization with a full-digital system. In this work, we present a novel instrument able to interface the control console of a nuclear reactor, developed by usingmore » CompactRio, a National Instruments embedded architecture and its dedicated programming language. This real-time industrial controller composed by a real-time processor and FPGA modules has been programmed to visualize the parameters coming from the reactor, and to storage and reproduce significant conditions anytime. This choice has been made on the basis of the FPGA properties: high reliability, determinism, true parallelism and re-configurability, achieved by a simple programming method, based on LabVIEW real-time environment. The system architecture exploits the FPGA capabilities of implementing custom timing and triggering, hardware-based analysis and co-processing, and highest performance control algorithms. Data stored during the supervisory phase can be reproduced by loading data from a measurement file, re-enacting worthwhile operations or conditions. The system has been thought to be used in three different modes, namely Log File Mode, Supervisory Mode and Simulation Mode. The proposed system can be considered as a first step to develop a more complete Decision Support System (DSS): indeed this work is part of a wider project that includes the elaboration of intelligent agents and meta-theory approaches. A synoptic has been created to monitor every kind of action on the plant through an intuitive sight. Furthermore, another important aim of this work is the possibility to have a front panel available on a web interface: CompactRio acts as a remote server and it is accessible on a dedicated LAN. This supervisory system has been tested and validated on the basis of the real control console for the 1-MW TRIGA reactor RC-1 at the ENEA, Casaccia Research Center. In this paper we show some results obtained by recording each variable as the reactor reaches its maximum level of power. The choice of a research reactor for testing the developed system relies on its training and didactic importance for the education of plant operators: in this context a digital instrument can offer a better user-friendly tool for learning and training. It is worthwhile to remark that such a system does not interfere with the console instrumentation, the latter continuing to preserve the total control. (authors)« less

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

NASA Technical Reports Server (NTRS)

Lantz, E.; Mayo, W.

1972-01-01

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

Optimally moderated nuclear fission reactor and fuel source therefor

DOEpatents

Ougouag, Abderrafi M [Idaho Falls, ID; Terry, William K [Shelley, ID; Gougar, Hans D [Idaho Falls, ID

2008-07-22

An improved nuclear fission reactor of the continuous fueling type involves determining an asymptotic equilibrium state for the nuclear fission reactor and providing the reactor with a moderator-to-fuel ratio that is optimally moderated for the asymptotic equilibrium state of the nuclear fission reactor; the fuel-to-moderator ratio allowing the nuclear fission reactor to be substantially continuously operated in an optimally moderated state.

In-reactor performance of LWR-type tritium target rods

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

Lanning, D.D.; Paxton, M.M.; Crumbaugh, L.

Pacific Northwest Laboratory has conducted several 1-yr irradiation tests of light water reactor-type tritium target rods. These tests have been sponsored by the U.S. Department of Energy's Office of New Production Reactors. The first test, designated water capsule-1 (WC-1), was conducted in the Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory from November 1989 to December 1990. The test vehicle contained a single 4-ft target rod within a pressurized water capsule. The capsule maintained the rod at pressurized water reactor (PWR)-type water temperature and pressure conditions. Posttest nondestructive examinations of the WC-1 rod involved visual examinations, dimensional checks,more » gamma scanning, and neutron radiography. The results indicate that the rod maintained the integrity of its pressure seal and was otherwise unaltered both mechanically and dimensionally by its irradiation and posttest handling.« less

[Research on change process of nitrosation granular sludge in continuous stirred-tank reactor].

PubMed

Yin, Fang-Fang; Liu, Wen-Ru; Wang, Jian-Fang; Wu, Peng; Shen, Yao-Liang

2014-11-01

In order to investigate the effect of different types of reactors on the nitrosation granular sludge, a continuous stirred-tank reactor (CSTR) was studied, using mature nitrosation granular sludge cultivated in sequencing batch reactor (SBR) as seed sludge. Results indicated that the change of reactor type and influent mode could induce part of granules to lose stability with gradual decrease in sludge settling ability during the initial period of operation. However, the flocs in CSTR achieved fast granulation in the following reactor operation. In spite of the changes of particle size distribution, e. g. the decreasing number of granules with diameter larger than 2.5 mm and the increasing number of granules with diameter smaller than 0.3 mm, granular sludge held the absolute predominance of sludge morphology in CSTR during the entire experimental period. Moreover, results showed that the change of reactor type and influent mode didn't affect the nitrite accumulation rate which was still kept at about 85% in effluent. Additionally, the average activity of the sludge in CSTR was stronger than that of the seed sludge, because the newly generated small particles in CSTR had higher specific reactive activity than the larger granules.

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.

Monte Carlo Analysis of the Battery-Type High Temperature Gas Cooled Reactor

NASA Astrophysics Data System (ADS)

Grodzki, Marcin; Darnowski, Piotr; Niewiński, Grzegorz

2017-12-01

The paper presents a neutronic analysis of the battery-type 20 MWth high-temperature gas cooled reactor. The developed reactor model is based on the publicly available data being an `early design' variant of the U-battery. The investigated core is a battery type small modular reactor, graphite moderated, uranium fueled, prismatic, helium cooled high-temperature gas cooled reactor with graphite reflector. The two core alternative designs were investigated. The first has a central reflector and 30×4 prismatic fuel blocks and the second has no central reflector and 37×4 blocks. The SERPENT Monte Carlo reactor physics computer code, with ENDF and JEFF nuclear data libraries, was applied. Several nuclear design static criticality calculations were performed and compared with available reference results. The analysis covered the single assembly models and full core simulations for two geometry models: homogenous and heterogenous (explicit). A sensitivity analysis of the reflector graphite density was performed. An acceptable agreement between calculations and reference design was obtained. All calculations were performed for the fresh core state.

NUCLEAR REACTOR CONTROL SYSTEM

DOEpatents

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

1959-11-01

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

10 CFR 52.1 - Definitions.

Code of Federal Regulations, 2014 CFR

2014-01-01

... authorization means the authorization provided by the Director of New Reactors or the Director of Nuclear... identical nuclear reactors (modules) and each module is a separate nuclear reactor capable of being operated... nuclear power reactor of the type described in 10 CFR 50.22. The approval may be for either the final...

10 CFR 52.1 - Definitions.

Code of Federal Regulations, 2011 CFR

2011-01-01

... authorization means the authorization provided by the Director of New Reactors or the Director of Nuclear... identical nuclear reactors (modules) and each module is a separate nuclear reactor capable of being operated... nuclear power reactor of the type described in 10 CFR 50.22. The approval may be for either the final...

10 CFR 52.1 - Definitions.

Code of Federal Regulations, 2013 CFR

2013-01-01

... authorization means the authorization provided by the Director of New Reactors or the Director of Nuclear... identical nuclear reactors (modules) and each module is a separate nuclear reactor capable of being operated... nuclear power reactor of the type described in 10 CFR 50.22. The approval may be for either the final...

10 CFR 52.1 - Definitions.

Code of Federal Regulations, 2012 CFR

2012-01-01

... authorization means the authorization provided by the Director of New Reactors or the Director of Nuclear... identical nuclear reactors (modules) and each module is a separate nuclear reactor capable of being operated... nuclear power reactor of the type described in 10 CFR 50.22. The approval may be for either the final...

NASA Reactor Facility Hazards Summary. Volume 1

NASA Technical Reports Server (NTRS)

1959-01-01

The Lewis Research Center of the National Aeronautics and Space Administration proposes to build a nuclear research reactor which will be located in the Plum Brook Ordnance Works near Sandusky, Ohio. The purpose of this report is to inform the Advisory Committee on Reactor Safeguards of the U. S. Atomic Energy Commission in regard to the design Lq of the reactor facility, the characteristics of the site, and the hazards of operation at this location. The purpose of this research reactor is to make pumped loop studies of aircraft reactor fuel elements and other reactor components, radiation effects studies on aircraft reactor materials and equipment, shielding studies, and nuclear and solid state physics experiments. The reactor is light water cooled and moderated of the MTR-type with a primary beryllium reflector and a secondary water reflector. The core initially will be a 3 by 9 array of MTR-type fuel elements and is designed for operation up to a power of 60 megawatts. The reactor facility is described in general terms. This is followed by a discussion of the nuclear characteristics and performance of the reactor. Then details of the reactor control system are discussed. A summary of the site characteristics is then presented followed by a discussion of the larger type of experiments which may eventually be operated in this facility. The considerations for normal operation are concluded with a proposed method of handling fuel elements and radioactive wastes. The potential hazards involved with failures or malfunctions of this facility are considered in some detail. These are examined first from the standpoint of preventing them or minimizing their effects and second from the standpoint of what effect they might have on the reactor facility staff and the surrounding population. The most essential feature of the design for location at the proposed site is containment of the maximum credible accident.

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

NASA Astrophysics Data System (ADS)

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

2001-02-01

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

Dynamic characteristics of a VK-50 reactor operating under conditions of the loss of a normal feedwater flow

NASA Astrophysics Data System (ADS)

Semidotskiy, I. I.; Kurskiy, A. S.

2013-12-01

The paper describes the conditions of the ATWS type with virtually complete cessation of the feed-water flow at the operating power level of a reactor of the VK-50 type. Under these conditions, the role of spatial kinetics in the system of feedback between thermohydraulic and nuclear processes with bulk boiling of the coolant in the reactor core is clearly seen. This feature determines the specific character of experimental data obtained and the suitability of their use for verification of the associated codes used for calculating water-water reactors.

Spherical torus fusion reactor

DOEpatents

Martin Peng, Y.K.M.

1985-10-03

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

NUCLEAR REACTOR FUEL-BREEDER FUEL ELEMENT

DOEpatents

Currier, E.L. Jr.; Nicklas, J.H.

1962-08-14

A fuel-breeder fuel element was developed for a nuclear reactor wherein discrete particles of fissionable material are dispersed in a matrix of fertile breeder material. The fuel element combines the advantages of a dispersion type and a breeder-type. (AEC)

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

NASA Astrophysics Data System (ADS)

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

2017-09-01

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

Design and fabrication of a fixed-bed batch type pyrolysis reactor for pilot scale pyrolytic oil production in Bangladesh

NASA Astrophysics Data System (ADS)

Aziz, Mohammad Abdul; Al-khulaidi, Rami Ali; Rashid, MM; Islam, M. R.; Rashid, MAN

2017-03-01

In this research, a development and performance test of a fixed-bed batch type pyrolysis reactor for pilot scale pyrolysis oil production was successfully completed. The characteristics of the pyrolysis oil were compared to other experimental results. A solid horizontal condenser, a burner for furnace heating and a reactor shield were designed. Due to the pilot scale pyrolytic oil production encountered numerous problems during the plant’s operation. This fixed-bed batch type pyrolysis reactor method will demonstrate the energy saving concept of solid waste tire by creating energy stability. From this experiment, product yields (wt. %) for liquid or pyrolytic oil were 49%, char 38.3 % and pyrolytic gas 12.7% with an operation running time of 185 minutes.

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

DTIC Science & Technology

2010-09-29

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

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.

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.

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...

Characterization of Used Nuclear Fuel with Multivariate Analysis for Process Monitoring

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

Dayman, Kenneth J.; Coble, Jamie B.; Orton, Christopher R.

2014-01-01

The Multi-Isotope Process (MIP) Monitor combines gamma spectroscopy and multivariate analysis to detect anomalies in various process streams in a nuclear fuel reprocessing system. Measured spectra are compared to models of nominal behavior at each measurement location to detect unexpected changes in system behavior. In order to improve the accuracy and specificity of process monitoring, fuel characterization may be used to more accurately train subsequent models in a full analysis scheme. This paper presents initial development of a reactor-type classifier that is used to select a reactor-specific partial least squares model to predict fuel burnup. Nuclide activities for prototypic usedmore » fuel samples were generated in ORIGEN-ARP and used to investigate techniques to characterize used nuclear fuel in terms of reactor type (pressurized or boiling water reactor) and burnup. A variety of reactor type classification algorithms, including k-nearest neighbors, linear and quadratic discriminant analyses, and support vector machines, were evaluated to differentiate used fuel from pressurized and boiling water reactors. Then, reactor type-specific partial least squares models were developed to predict the burnup of the fuel. Using these reactor type-specific models instead of a model trained for all light water reactors improved the accuracy of burnup predictions. The developed classification and prediction models were combined and applied to a large dataset that included eight fuel assembly designs, two of which were not used in training the models, and spanned the range of the initial 235U enrichment, cooling time, and burnup values expected of future commercial used fuel for reprocessing. Error rates were consistent across the range of considered enrichment, cooling time, and burnup values. Average absolute relative errors in burnup predictions for validation data both within and outside the training space were 0.0574% and 0.0597%, respectively. The errors seen in this work are artificially low, because the models were trained, optimized, and tested on simulated, noise-free data. However, these results indicate that the developed models may generalize well to new data and that the proposed approach constitutes a viable first step in developing a fuel characterization algorithm based on gamma spectra.« less

Computational Analysis Supporting the Design of a New Beamline for the Mines Neutron Radiography Facility

NASA Astrophysics Data System (ADS)

Wilson, C.; King, J.

The Colorado School of Mines installed a neutron radiography system at the United States Geological Survey TRIGA reactor in 2012. An upgraded beamline could dramatically improve the imaging capabilities of this system. This project performed computational analyses to support the design of a new beamline, with the major goals of minimizing beam divergence and maximizing beam intensity. The new beamline will consist of a square aluminum tube with an 11.43 cm (4.5 in) inner side length and 0.635 cm (0.25 in) thick walls. It is the same length as the original beam tube (8.53 m) and is composed of 1.22 m (4 ft) and 1.52 m (5 ft) flanged sections which bolt together. The bottom 1.22 m of the beamline is a cylindrical aluminum pre-collimator which is 0.635 cm (0.25 in) thick, with an inner diameter of 5.08 cm (2 in). Based on Monte Carlo model results, when a pre-collimator is present, the use of a neutron absorbing liner on the inside surface of the beam tube has almost no effect on the angular distribution of the neutron current at the collimator exit. The use of a pre-collimator may result in a non-uniform flux profile at the image plane; however, as long as the collimator is at least three times longer than the pre-collimator, the flux distortion is acceptably low.

An Assessment of the Potential Use of BNNTs for Boron Neutron Capture Therapy.

PubMed

Ferreira, Tiago H; Miranda, Marcelo C; Rocha, Zildete; Leal, Alexandre S; Gomes, Dawidson A; Sousa, Edesia M B

2017-04-12

Currently, nanostructured compounds have been standing out for their optical, mechanical, and chemical features and for the possibilities of manipulation and regulation of complex biological processes. One of these compounds is boron nitride nanotubes (BNNTs), which are a nanostructured material analog to carbon nanotubes, but formed of nitrogen and boron atoms. BNNTs present high thermal stability along with high chemical inertia. Among biological applications, its biocompatibility, cellular uptake, and functionalization potential can be highlighted, in addition to its eased utilization due to its nanometric size and tumor cell internalization. When it comes to new forms of therapy, we can draw attention to boron neutron capture therapy (BNCT), an experimental radiotherapy characterized by a boron-10 isotope carrier inside the target and a thermal neutron beam focused on it. The activation of the boron-10 atom by a neutron generates a lithium atom, a gamma ray, and an alpha particle, which can be used to destroy tumor tissues. The aim of this work was to use BNNTs as a boron-10 carrier for BNCT and to demonstrate its potential. The nanomaterial was characterized through XRD, FTIR, and SEM. The WST-8 assay was performed to confirm the cell viability of BNNTs. The cells treated with BNNTs were irradiated with the neutron beam of a Triga reactor, and the apoptosis caused by the activation of the BNNTs was measured with a calcein AM/propidium iodide test. The results demonstrate that this nanomaterial is a promising candidate for cancer therapy through BNCT.

Texas Intense Positron Source (TIPS)

NASA Astrophysics Data System (ADS)

O'Kelly, D.

2003-03-01

The Texas Intense Positron Source (TIPS) is a state of the art variable energy positron beam under construction at the Nuclear Engineering Teaching Laboratory (NETL). Projected intensities on the order of the order of 10^7 e+/second using ^64Cu as the positron source are expected. Owing to is short half-life (t1/2 12.8 hrs), plans are to produce the ^64Cu isotope on-site using beam port 1 of NETL TRIGA Mark II reactor. Following tungsten moderation, the positrons will be electrostatically focused and accelerated from few 10's of eV up to 30 keV. This intensity and energy range should allow routine performance of several analytical techniques of interest to surface scientists (PALS, PADB and perhaps PAES and LEPD.) The TIPS project is being developed in parallel phases. Phase I of the project entails construction of the vacuum system, source chamber, main beam line, electrostatic/magnetic focusing and transport system as well as moderator design. Initial construction, testing and characterization of moderator and beam transport elements are underway and will use a commercially available 10 mCi ^22Na radioisotope as a source of positrons. Phase II of the project is concerned primarily with the Cu source geometry and thermal properties as well as production and physical handling of the radioisotope. Additional instrument optimizing based upon experience gained during Phase I will be incorporated in the final design. Current progress of both phases will be presented along with motivations and future directions.

MATSIM -The Development and Validation of a Numerical Voxel Model based on the MATROSHKA Phantom

NASA Astrophysics Data System (ADS)

Beck, Peter; Rollet, Sofia; Berger, Thomas; Bergmann, Robert; Hajek, Michael; Latocha, Marcin; Vana, Norbert; Zechner, Andrea; Reitz, Guenther

The AIT Austrian Institute of Technology coordinates the project MATSIM (MATROSHKA Simulation) in collaboration with the Vienna University of Technology and the German Aerospace Center. The aim of the project is to develop a voxel-based model of the MATROSHKA anthro-pomorphic torso used at the International Space Station (ISS) as foundation to perform Monte Carlo high-energy particle transport simulations for different irradiation conditions. Funded by the Austrian Space Applications Programme (ASAP), MATSIM is a co-investigation with the European Space Agency (ESA) ELIPS project MATROSHKA, an international collaboration of more than 18 research institutes and space agencies from all over the world, under the science and project lead of the German Aerospace Center. The MATROSHKA facility is designed to determine the radiation exposure of an astronaut onboard ISS and especially during an ex-travehicular activity. The numerical model developed in the frame of MATSIM is validated by reference measurements. In this report we give on overview of the model development and compare photon and neutron irradiations of the detector-equipped phantom torso with Monte Carlo simulations using FLUKA. Exposure to Co-60 photons was realized in the standard ir-radiation laboratory at Seibersdorf, while investigations with neutrons were performed at the thermal column of the Vienna TRIGA Mark-II reactor. The phantom was loaded with passive thermoluminescence dosimeters. In addition, first results of the calculated dose distribution within the torso are presented for a simulated exposure in low-Earth orbit.

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

DTIC Science & Technology

2010-06-10

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

Application of a novel type impinging streams reactor in solid-liquid enzyme reactions and modeling of residence time distribution using GDB model.

PubMed

Fatourehchi, Niloufar; Sohrabi, Morteza; Dabir, Bahram; Royaee, Sayed Javid; Haji Malayeri, Adel

2014-02-05

Solid-liquid enzyme reactions constitute important processes in biochemical industries. The isomerization of d-glucose to d-fructose, using the immobilized glucose isomerase (Sweetzyme T), as a typical example of solid-liquid catalyzed reactions has been carried out in one stage and multi-stage novel type of impinging streams reactors. Response surface methodology was applied to determine the effects of certain pertinent parameters of the process namely axial velocity (A), feed concentration (B), nozzles' flow rates (C) and enzyme loading (D) on the performance of the apparatus. The results obtained from the conversion of glucose in this reactor were much higher than those expected in conventional reactors, while residence time was decreased dramatically. Residence time distribution (RTD) in a one-stage impinging streams reactor was investigated using colored solution as the tracer. The results showed that the flow pattern in the reactor was close to that in a continuous stirred tank reactor (CSTR). Based on the analysis of flow region in the reactor, gamma distribution model with bypass (GDB) was applied to study the RTD of the reactor. The results indicated that RTD in the impinging streams reactor could be described by the latter model. Copyright © 2013 Elsevier Inc. All rights reserved.

Performance of compact fast pyrolysis reactor with Auger-type modules for the continuous liquid biofuel production

NASA Astrophysics Data System (ADS)

Nishimura, Shun; Ebitani, Kohki

2018-01-01

Development of a compact fast pyrolysis reactor constructed using Auger-type technology to afford liquid biofuel with high yield has been an interesting concept in support of local production for local consumption. To establish a widely useable module package, details of the performance of the developing compact module reactor were investigated. This study surveyed the properties of as-produced pyrolysis oil as a function of operation time, and clarified the recent performance of the developing compact fast pyrolysis reactor. Results show that after condensation in the scrubber collector, e.g. approx. 10 h for a 25 kg/h feedstock rate, static performance of pyrolysis oil with approximately 20 MJ/kg (4.8 kcal/g) calorific values were constantly obtained after an additional 14 h. The feeding speed of cedar chips strongly influenced the time for oil condensation process: i.e. 1.6 times higher feeding speed decreased the condensation period by half (approx. 5 h in the case of 40 kg/h). Increasing the reactor throughput capacity is an important goal for the next stage in the development of a compact fast pyrolysis reactor with Auger-type modules.

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

A brief history of design studies on innovative nuclear reactors

NASA Astrophysics Data System (ADS)

Sekimoto, Hiroshi

2014-09-01

In a short period after the success of CP1, many types of nuclear reactors were proposed and investigated. However, soon only a small number of reactors were selected for practical use. Around 1970, only LWRs with small number of CANDUs were operated in the western world, and FBRs were under development. It was about the time when Apollo moon landing was accomplished. However, at the same time, the future of human being was widely considered pessimistic and Limits to Growth was published. In the end of 1970's the TMI accident occurred and many nuclear reactor contracts were cancelled in USA and any more contracts had not been concluded until recent years. From the reflection of this accident, many Inherent Safe Reactors (ISRs) were proposed, though none of them were constructed. A common idea of ISRs is smallness of their size. Tokyo Institute of Technology (TokyoTech) held a symposium on small reactors, SR/TIT, in 1991, where many types of small ISRs were presented. Recently small reactors attract interest again. The most ideas employed in these reactors were the same discussed in SR/TIT. In 1980's the radioactive wastes from fuel cycle became a severe problem around the world. In TokyoTech, this issue was discussed mainly from the viewpoint of nuclear transmutations. The neutron economy became inevitable for these innovative nuclear reactors especially small long-life reactors and transmutation reactors.

Period meter for reactors

DOEpatents

Rusch, Gordon K.

1976-01-06

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

Numerical Simulations of a 96-rod Polysilicon CVD Reactor

NASA Astrophysics Data System (ADS)

Guoqiang, Tang; Cong, Chen; Yifang, Cai; Bing, Zong; Yanguo, Cai; Tihu, Wang

2018-05-01

With the rapid development of the photovoltaic industry, pressurized Siemens belljar-type polysilicon CVD reactors have been enlarged from 24 rods to 96 rods in less than 10 years aimed at much greater single-reactor productivity. A CFD model of an industry-scale 96-rod CVD reactor was established to study the internal temperature distribution and the flow field of the reactor. Numerical simulations were carried out and compared with actual growth results from a real CVD reactor. Factors affecting polysilicon depositions such as inlet gas injections, flow field, and temperature distribution in the CVD reactor are studied.

Comparative evaluation of solar, fission, fusion, and fossil energy resources. Part 2: Power from nuclear fission

NASA Technical Reports Server (NTRS)

Clement, J. D.

1973-01-01

Different types of nuclear fission reactors and fissionable materials are compared. Special emphasis is placed upon the environmental impact of such reactors. Graphs and charts comparing reactor facilities in the U. S. are presented.

An atmospheric pressure flow reactor: Gas phase kinetics and mechanism in tropospheric conditions without wall effects

NASA Technical Reports Server (NTRS)

Koontz, Steven L.; Davis, Dennis D.; Hansen, Merrill

1988-01-01

A new type of gas phase flow reactor, designed to permit the study of gas phase reactions near 1 atm of pressure, is described. A general solution to the flow/diffusion/reaction equations describing reactor performance under pseudo-first-order kinetic conditions is presented along with a discussion of critical reactor parameters and reactor limitations. The results of numerical simulations of the reactions of ozone with monomethylhydrazine and hydrazine are discussed, and performance data from a prototype flow reactor are presented.

Treatment of screened dairy manure by upflow anaerobic fixed bed reactors packed with waste tyre rubber and a combination of waste tyre rubber and zeolite: effect of the hydraulic retention time.

PubMed

Umaña, Oscar; Nikolaeva, Svetlana; Sánchez, Enrique; Borja, Rafael; Raposo, Francisco

2008-10-01

Two laboratory-scale anaerobic fixed bed reactors were evaluated while treating dairy manure at upflow mode and semicontinuous feeding. One reactor was packed with a combination of waste tyre rubber and zeolite (R1) while the other had only waste tyre rubber as a microorganism immobilization support (R2). Effluent quality improved when the hydraulic retention time (HRT) increased from 1.0 to 5.5 days. Higher COD, BOD5, total and volatile solids removal efficiencies were always achieved in the reactor R1. No clogging was observed during the operation period. Methane yield was also a function of the HRT and of the type of support used, and was 12.5% and 40% higher in reactor R1 than in R2 for HRTs of 5.5 and 1.0 days, respectively. The results obtained demonstrated that this type of reactor is capable of operating with dairy manure at a HRT 5 times lower than that used in a conventional reactor.

Factors affecting cleanup of exhaust gases from a pressurized, fluidized-bed coal combustor

NASA Technical Reports Server (NTRS)

Rollbuhler, R. J.; Kobak, J. A.

1980-01-01

The cleanup of effluent gases from the fluidized-bed combustion of coal is examined. Testing conditions include the type and feed rate of the coal and the sulfur sorbent, the coal-sorbent ratio, the coal-combustion air ratio, the depth of the reactor fluidizing bed, and the technique used to physically remove fly ash from the reactor effluent gases. Tests reveal that the particulate loading matter in the effluent gases is a function not only of the reactor-bed surface gas velocity, but also of the type of coal being burnt and the time the bed is operating. At least 95 percent of the fly ash particules in the effluent gas are removed by using a gas-solids separator under controlled operating conditions. Gaseous pollutants in the effluent (nitrogen and sulfur oxides) are held within the proposed Federal limits by controlling the reactor operating conditions and the type and quantity of sorbent material.

Thermomechanical analysis of fast-burst reactors

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

Miller, J.D.

1994-08-01

Fast-burst reactors are designed to provide intense, short-duration pulses of neutrons. The fission reaction also produces extreme time-dependent heating of the nuclear fuel. An existing transient-dynamic finite element code was modified specifically to compute the time-dependent stresses and displacements due to thermal shock loads of reactors. Thermomechanical analysis was then applied to determine structural feasibility of various concepts for an EDNA-type reactor and to optimize the mechanical design of the new SPR III-M reactor.

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

Bissani, M; Tyson, S

Vietnam's nuclear program began in the 1960s with the installation at Dalat of a 250 kW TRIGA Mk-II research reactor under the U.S. Atoms for Peace Program. The reactor was shut down and its core removed only a few years later, and the nuclear research program was suspended until after the end of the civil war in the late 1970s. The Soviet Union assisted Vietnam in restoring the Dalat reactor to an operational status in 1984, trained a cadre of scientific and technical staff in its operation, and contributed to the development of nuclear science for the medical and agriculturalmore » sectors. In the agricultural area in particular, Vietnamese experts have been very successful in developing mutant strains of rice, and continue to work with the IAEA to yield strains that have a shorter growing period, increased resistance to disease, and other desirable characteristics. Rice has always been the main crop in Vietnam, but technical cooperation with the IAEA and other states has enabled the country to become one of the top rice producers in the world, exporting much of its annual crop to over two dozen countries annually. More recently, Vietnam's government has shown increasing interest in developing a civil nuclear program to supplement its fossil fuel and other energy resources. Projections from a variety of open sources, ranging from the IAEA, the U.S. Department of Energy's Energy Information Administration (EIA), the Vietnamese government, energy corporations, and think tanks all predict a massive increase in energy consumption--especially electricity--within Vietnam and the region as a whole. This growth in consumption will require a corresponding increase in energy production, which in Vietnam is currently satisfied mainly by fossil fuels (coal) and renewable energy (hydropower and biomass); Vietnam has a refining capacity of about 800 barrels/day. Most of its crude oil is exported to generate export income, and is not used to generate electricity. Although Vietnam is able to meet most of its needs through its own resources, it consumes more electricity than it produces (approximately six billion KWh/a). Open sources indicate that increasing exports of manufactured goods and a corresponding growth in the industrial sector (16% since 2003) will lead to a greater energy shortfall beyond the next decade--between 35 and 65 billion KWh/year after 2020. As a signatory to the Kyoto Treaty and other regional environmental accords, and seeking an alternative to its stretched hydroelectric resources, Vietnam is looking for a means to increase energy production while minimizing emissions of greenhouse gases.« less

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

Nygaard, E. T.; Williams, M. M. R.; Angelo, P. L.

Babcock and Wilcox Technical Services Group (B and W) has identified aqueous homogeneous reactors (AHRs) as a technology well suited to produce the medical isotope molybdenum 99 (Mo-99). AHRs have never been specifically designed or built for this specialized purpose. However, AHRs have a proven history of being safe research reactors. In fact, in 1958, AHRs had 'a longer history of operation than any other type of research reactor using enriched fuel' and had 'experimentally demonstrated to be among the safest of all various type of research reactor now in use [1].' A 'Level 1' model representing B and W'smore » proposed Medical Isotope Production System (MIPS) reactor has been developed. The Level 1 model couples a series of differential equations representing neutronics, temperature, and voiding. Neutronics are represented by point reactor kinetics while temperature and voiding terms are axially varying (one-dimensional). While this model was developed specifically for the MIPS reactor, its applicability to the Japanese TRACY reactor was assessed. The results from the Level 1 model were in good agreement with TRACY experimental data and found to be conservative over most of the time domains considered. The Level 1 model was used to study the MIPS reactor. An analysis showed the Level 1 model agreed well with a more complex computational model of the MIPS reactor (a FETCH model). Finally, a significant reactivity insertion was simulated with the Level 1 model to study the MIPS reactor's time-dependent response. (authors)« less

A brief history of design studies on innovative nuclear reactors

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

Sekimoto, Hiroshi, E-mail: hsekimot@gmail.com

2014-09-30

In a short period after the success of CP1, many types of nuclear reactors were proposed and investigated. However, soon only a small number of reactors were selected for practical use. Around 1970, only LWRs with small number of CANDUs were operated in the western world, and FBRs were under development. It was about the time when Apollo moon landing was accomplished. However, at the same time, the future of human being was widely considered pessimistic and Limits to Growth was published. In the end of 1970’s the TMI accident occurred and many nuclear reactor contracts were cancelled in USAmore » and any more contracts had not been concluded until recent years. From the reflection of this accident, many Inherent Safe Reactors (ISRs) were proposed, though none of them were constructed. A common idea of ISRs is smallness of their size. Tokyo Institute of Technology (TokyoTech) held a symposium on small reactors, SR/TIT, in 1991, where many types of small ISRs were presented. Recently small reactors attract interest again. The most ideas employed in these reactors were the same discussed in SR/TIT. In 1980’s the radioactive wastes from fuel cycle became a severe problem around the world. In TokyoTech, this issue was discussed mainly from the viewpoint of nuclear transmutations. The neutron economy became inevitable for these innovative nuclear reactors especially small long-life reactors and transmutation reactors.« 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.

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

Baines, B.D.

The development of the two types of Jason reactor is reported (10-kw Standard Jason, 100-kw Jason). Essential data are given on their construction and operation. The projects which were, or could be, carried out with these reactors are briefiy mentioned, with special emphasis on the adaptability of the reactor to various uses. (autb)

Current status of the development of high density LEU fuel for Russian research reactors

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

Vatulin, A.; Dobrikova, I.; Suprun, V.

2008-07-15

One of the main directions of the Russian RERTR program is to develop U-Mo fuel and fuel elements/FA with this fuel. The development is carried out both for existing reactors, and for new advanced designs of reactors. Many organizations in Russia, i.e. 'TVEL', RDIPE, RIAR, IRM, NPCC participate in the work. Two fuels are under development: dispersion and monolithic U-Mo fuel, as well two types of FA to use the dispersion U-Mo fuel: with tubular type fuel elements and with pin type fuel elements. The first stage of works was successfully completed. This stage included out-pile, in-pile and post irradiationmore » examinations of U-Mo dispersion fuel in experimental tubular and pin fuel elements under parameters similar to operation conditions of Russian design pool-type research reactors. The results received both in Russia and abroad enabled to go on to the next stage of development which includes irradiation tests both of full-scale IRT pin-type and tube-type fuel assemblies with U-Mo dispersion fuel and of mini-fuel elements with modified U-Mo dispersion fuel and monolithic fuel. The paper gives a generalized review of the results of U-Mo fuel development accomplished by now. (author)« less

CFD optimization of continuous stirred-tank (CSTR) reactor for biohydrogen production.

PubMed

Ding, Jie; Wang, Xu; Zhou, Xue-Fei; Ren, Nan-Qi; Guo, Wan-Qian

2010-09-01

There has been little work on the optimal configuration of biohydrogen production reactors. This paper describes three-dimensional computational fluid dynamics (CFD) simulations of gas-liquid flow in a laboratory-scale continuous stirred-tank reactor used for biohydrogen production. To evaluate the role of hydrodynamics in reactor design and optimize the reactor configuration, an optimized impeller design has been constructed and validated with CFD simulations of the normal and optimized impeller over a range of speeds and the numerical results were also validated by examination of residence time distribution. By integrating the CFD simulation with an ethanol-type fermentation process experiment, it was shown that impellers with different type and speed generated different flow patterns, and hence offered different efficiencies for biohydrogen production. The hydrodynamic behavior of the optimized impeller at speeds between 50 and 70 rev/min is most suited for economical biohydrogen production. Copyright 2010 Elsevier Ltd. All rights reserved.

Transesterification of rapeseed oil for biodiesel production in trickle-bed reactors packed with heterogeneous Ca/Al composite oxide-based alkaline catalyst.

PubMed

Meng, Yong-Lu; Tian, Song-Jiang; Li, Shu-Fen; Wang, Bo-Yang; Zhang, Min-Hua

2013-05-01

A conventional trickle bed reactor and its modified type both packed with Ca/Al composite oxide-based alkaline catalysts were studied for biodiesel production by transesterification of rapeseed oil and methanol. The effects of the methanol usage and oil flow rate on the FAME yield were investigated under the normal pressure and methanol boiling state. The oil flow rate had a significant effect on the FAME yield for the both reactors. The modified trickle bed reactor kept over 94.5% FAME yield under 0.6 mL/min oil flow rate and 91 mL catalyst bed volume, showing a much higher conversion and operational stability than the conventional type. With the modified trickle bed reactor, both transesterification and methanol separation could be performed simultaneously, and glycerin and methyl esters were separated additionally by gravity separation. Copyright © 2013 Elsevier Ltd. All rights reserved.

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

NASA Astrophysics Data System (ADS)

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

2013-09-01

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

Low drift type N thermocouples in out-of-pile advanced gas reactor mock-up test: metallurgical analysis

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

Scervini, M.; Palmer, J.; Haggard, D.C.

2015-07-01

Thermocouples are the most commonly used sensors for temperature measurement in nuclear reactors. They are crucial for the control of current nuclear reactors and for the development of GEN IV reactors. In nuclear applications thermocouples are strongly affected by intense neutron fluxes. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time dependent change in composition and, as a consequence, a time dependent drift of the thermocouple signal. Thermocouple drift can be very significant for in-pile temperature measurements and may render the temperature sensors unreliable after exposure to nuclear radiation formore » relatively short times compared to the life required for temperature sensors in nuclear applications. Previous experiences with type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly type N thermocouples are expected to be only slightly affected by neutron fluxes. Currently the use of Nickel based thermocouples is limited to temperatures lower than 1000 deg. C due to drift related to phenomena other than nuclear irradiation. As part of a collaboration between Idaho National Laboratory (INL) and the University of Cambridge a variety of Type N thermocouples have been exposed at INL in an Advanced Gas Reactor mock-up test at 1150 deg. C for 2000 h, 1200 deg. C for 2000 h, 125 deg. C for 200 h and 1300 deg. C for 200 h, and later analysed metallurgically at the University of Cambridge. The use of electron microscopy allows to identify the metallurgical changes occurring in the thermocouples during high temperature exposure and correlate the time dependent thermocouple drift with the microscopic changes experienced by the thermoelements of different thermocouple designs. In this paper conventional Inconel 600 sheathed type N thermocouples and a type N using a customized sheath developed at the University of Cambridge have been investigated. The rationale for the superior performance of the type N using a customized sheath developed at the University of Cambridge is explained in comparison with the behavior of conventional type N Inconel 600 sheathed thermocouples. (authors)« less

10 CFR 72.6 - License required; types of licenses.

Code of Federal Regulations, 2011 CFR

2011-01-01

... SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General... the receipt, handling, storage, and transfer of reactor-related GTCC are specific licenses. Any... hereby issued to receive title to and own spent fuel, high-level radioactive waste, or reactor-related...

10 CFR 72.6 - License required; types of licenses.

Code of Federal Regulations, 2010 CFR

2010-01-01

... SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General... the receipt, handling, storage, and transfer of reactor-related GTCC are specific licenses. Any... hereby issued to receive title to and own spent fuel, high-level radioactive waste, or reactor-related...

REVIEW OF POWER AND HEAT REACTOR DESIGNS. Domestic and Foreign

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

Appleby, E.R., comp

1963-10-01

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

Nuclear reactor shield including magnesium oxide

DOEpatents

Rouse, Carl A.; Simnad, Massoud T.

1981-01-01

An improvement in nuclear reactor shielding of a type used in reactor applications involving significant amounts of fast neutron flux, the reactor shielding including means providing structural support, neutron moderator material, neutron absorber material and other components as described below, wherein at least a portion of the neutron moderator material is magnesium in the form of magnesium oxide either alone or in combination with other moderator materials such as graphite and iron.

REUSABLE ADSORBENTS FOR DILUTE SOLUTIONS SEPARATION. 6. BATCH AND CONTINUOUS REACTORS FOR ADSORPTION AND DEGRADATION OF 1,2-DICHLOROBENZENE FROM DILUTE WASTEWATER STREAMS USING TITANIA AS A PHOTOCATALYST. (R828598C753)

EPA Science Inventory

Two types of external lamp reactors were investigated for the titania catalyzed photodegradation of 1,2-dichlorobenzene (DCB) from a dilute water stream. The first one was a batch mixed slurry reactor and the second one was a semi-batch reactor with continuous feed recycle wit...

The effect of transient loading on the performance of a mesophilic anaerobic contact reactor at constant feed strength.

PubMed

Sentürk, Elif; Ince, Mahir; Engin, Guleda Onkal

2012-12-15

Anaerobic contact reactor is a high rate anaerobic process consisting of an agitated reactor and a solids settling tank for recycling. It was proved earlier that this type of reactor design offers highly efficient performance in the conversion of organic matter to biogas. In this study, the effect of transient loading on reactor performance in terms of a number of key intermediates and parameters such as, COD removal, pH and alkalinity change, VFAs, effluent MLSS concentration and biogas efficiency over time was examined. For this purpose, a step increase of organic loading rate from 3.35kg COD/m(3)day to 15.61kg COD/m(3)day was employed. The hydraulic retention time decreased to a value of 8.42h by an increase in the influent flow-rate during the transient loading. It was observed that the mesophilic anaerobic contact reactor (MACR) was quite resistant to large transient shocks. The reactor recovered back to its baseline performance only in 15h after the shock loading was stopped. Hence, it can be concluded that this type of reactor design has a high potential in treating food processing wastewaters with varying flow characteristics. Copyright © 2012 Elsevier B.V. All rights reserved.

MTR WING, TRA604. PRECAST CONCRETE PANELS AND DIMENSIONS. TYPES A, ...

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

MTR WING, TRA-604. PRECAST CONCRETE PANELS AND DIMENSIONS. TYPES A, B, C, D, E, AND F; AND HOW THEY ARE CONNECTED. TYPES C AND D ARE ON WEST SIDE WHERE GLASS BLOCKS SURROUND ENTRY DOOR. BLAW-KNOX 3150-804-20, SHEET #1, 11/1950. INL INDEX NO. 531-0604-62-098-100644, REV. 0. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Accident analysis of heavy water cooled thorium breeder reactor

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

Yulianti, Yanti; Su’ud, Zaki; Takaki, Naoyuki

2015-04-16

Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k,more » and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The power reactor has a peak value before reactor has new balance condition. The analysis showed that temperatures of fuel and claddings during accident are still below limitations which are in secure condition.« less

Accident analysis of heavy water cooled thorium breeder reactor

NASA Astrophysics Data System (ADS)

Yulianti, Yanti; Su'ud, Zaki; Takaki, Naoyuki

2015-04-01

Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The power reactor has a peak value before reactor has new balance condition. The analysis showed that temperatures of fuel and claddings during accident are still below limitations which are in secure condition.

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...

78 FR 33132 - Quality Verification for Plate-Type Uranium-Aluminum Fuel Elements for Use in Research and Test...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-03

... Fuel Elements for Use in Research and Test Reactors AGENCY: Nuclear Regulatory Commission. ACTION... Research and Test Reactors.'' This guide describes a method that the staff of the NRC considers acceptable... assurance program for verifying the quality of plate-type uranium-aluminum fuel elements used in research...

Catalytic fast pyrolysis of white oak wood in-situ using a bubbling fluidized bed reactor

USDA-ARS?s Scientific Manuscript database

Catalytic fast pyrolysis was performed on white oak wood using two zeolite-type catalysts as bed material in a bubbling fluidized bed reactor. The two catalysts chosen, based on a previous screening study, were Ca2+ exchanged Y54 (Ca-Y54) and a proprietary ß-zeolite type catalyst (catalyst M) both ...

Energy production using fission fragment rockets

NASA Astrophysics Data System (ADS)

Chapline, G.; Matsuda, Y.

1991-08-01

Fission fragment rockets are nuclear reactors with a core consisting of thin fibers in a vacuum, and which use magnetic fields to extract the fission fragments from the reactor core. As an alternative to ordinary nuclear reactors, fission fragment rockets would have the following advantages: approximately twice the efficiency if the fission fragment energy can be directly converted into electricity; reduction of the buildup of a fission fragment inventory in the reactor could avoid a Chernobyl type disaster; and collection of the fission fragments outside the reactor could simplify the waste disposal problem.

Determining Reactor Fuel Type from Continuous Antineutrino Monitoring

NASA Astrophysics Data System (ADS)

Jaffke, Patrick; Huber, Patrick

2017-09-01

We investigate the ability of an antineutrino detector to determine the fuel type of a reactor. A hypothetical 5-ton antineutrino detector is placed 25 m from the core and measures the spectral shape and rate of antineutrinos emitted by fission fragments in the core for a number of 90-d periods. Our results indicate that four major fuel types can be differentiated from the variation of fission fractions over the irradiation time with a true positive probability of detection at approximately 95%. In addition, we demonstrate that antineutrinos can identify the burnup at which weapons-grade mixed-oxide (MOX) fuel would be reduced to reactor-grade MOX, on average, providing assurance that plutonium-disposition goals are met. We also investigate removal scenarios where plutonium is purposefully diverted from a mixture of MOX and low-enriched uranium fuel. Finally, we discuss how our analysis is impacted by a spectral distortion around 6 MeV observed in the antineutrino spectrum measured from commercial power reactors.

Final report, PT IP-535-C: Test of smaller VSR`s in DR reactor

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

Vaughn, A.D.

1963-04-17

Because of rod-sticking problems at DR Reactor, a knuckle rod of B Reactor design was installed in vertical safety channel number 28. The substitute VSR, which has a smaller diameter than the original DR rod, has been tested for its operational feasibility including both drop time and reactivity effect. The reactivity effect of the rod was estimated by comparison of the reactivity transient caused by insertion of the specific B-type rod after scramming into the pile, with similar transients caused by normal vertical safety rod in an adjacent channel. This document lists the indicated relative control strength of the rodmore » as an empirical basis for future safety calculations. Results indicate that the B-type knuckel rod in DR reactor is about 80% as strong as a normal DR vertical safety rod if used in equivalent flux distribution and location; this makes it reasonable to assume that the local control strength of the B-type knuckel rod is 98 {mu}b.« less

Early Program Development

NASA Image and Video Library

1963-01-01

This artist's concept from 1963 shows a proposed NERVA (Nuclear Engine for Rocket Vehicle Application) incorporating the NRX-A1, the first NERVA-type cold flow reactor. The NERVA engine, based on Kiwi nuclear reactor technology, was intended to power a RIFT (Reactor-In-Flight-Test) nuclear stage, for which Marshall Space Flight Center had development responsibility.

Gas-phase optical fiber photocatalytic reactors for indoor air application: a preliminary study on performance indicators

NASA Astrophysics Data System (ADS)

Palmiste, Ü.; Voll, H.

2017-10-01

The development of advanced air cleaning technologies aims to reduce building energy consumption by reduction of outdoor air flow rates while keeping the indoor air quality at an acceptable level by air cleaning. Photocatalytic oxidation is an emerging technology for gas-phase air cleaning that can be applied in a standalone unit or a subsystem of a building mechanical ventilation system. Quantitative information on photocatalytic reactor performance is required to evaluate the technical and economic viability of the advanced air cleaning by PCO technology as an energy conservation measure in a building air conditioning system. Photocatalytic reactors applying optical fibers as light guide or photocatalyst coating support have been reported as an approach to address the current light utilization problems and thus, improve the overall efficiency. The aim of the paper is to present a preliminary evaluation on continuous flow optical fiber photocatalytic reactors based on performance indicators commonly applied for air cleaners. Based on experimental data, monolith-type optical fiber reactor performance surpasses annular-type optical fiber reactors in single-pass removal efficiency, clean air delivery rate and operating cost efficiency.

Transport Corrections in Nodal Diffusion Codes for HTR Modeling

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

Abderrafi M. Ougouag; Frederick N. Gleicher

2010-08-01

The cores and reflectors of High Temperature Reactors (HTRs) of the Next Generation Nuclear Plant (NGNP) type are dominantly diffusive media from the point of view of behavior of the neutrons and their migration between the various structures of the reactor. This means that neutron diffusion theory is sufficient for modeling most features of such reactors and transport theory may not be needed for most applications. Of course, the above statement assumes the availability of homogenized diffusion theory data. The statement is true for most situations but not all. Two features of NGNP-type HTRs require that the diffusion theory-based solutionmore » be corrected for local transport effects. These two cases are the treatment of burnable poisons (BP) in the case of the prismatic block reactors and, for both pebble bed reactor (PBR) and prismatic block reactor (PMR) designs, that of control rods (CR) embedded in non-multiplying regions near the interface between fueled zones and said non-multiplying zones. The need for transport correction arises because diffusion theory-based solutions appear not to provide sufficient fidelity in these situations.« less

Energy from nuclear fission()

NASA Astrophysics Data System (ADS)

Ripani, M.

2015-08-01

The main features of nuclear fission as physical phenomenon will be revisited, emphasizing its peculiarities with respect to other nuclear reactions. Some basic concepts underlying the operation of nuclear reactors and the main types of reactors will be illustrated, including fast reactors, showing the most important differences among them. The nuclear cycle and radioactive-nuclear-waste production will be also discussed, along with the perspectives offered by next generation nuclear assemblies being proposed. The current situation of nuclear power in the world, its role in reducing carbon emission and the available resources will be briefly illustrated.

Precise Measurement of αK for the 39.76-keV E3 transition in 103Rh: Further Test of Internal Conversion Theory

NASA Astrophysics Data System (ADS)

Bryant, Justin; Park, Hyo In; Nica, Ninel; Iacob, Victor; Hardy, John

2017-09-01

We have extended our series of precision measurements of internal conversion coefficients (ICC) to include the 39.76-keV, E3 transition in 103Rh. Our goal has been to test the Dirac-Fock ICC calculations, specifically with respect to the role of the atomic vacancy created in the conversion process. We prepared a sample from pure (natural) ruthenium chloride by converting the sample to ruthenium oxide, electrochemically depositing it on an aluminum backing, and subsequently activating it with thermal neutrons at the Texas A&M TRIGA reactor for 20 hours. Decay spectra were then recorded for roughly 120 hours with a HPGe detector that has been precisely efficiency calibrated (+/-.15% relative precision). In the acquired spectra, all impurities were identified and corrected for accordingly. A program was written using the ROOT framework developed by CERN to extract the area of the 39.76-keV gamma-ray peak from 103Rh, which partially overlapped the Kα x-ray peaks from a 153Gd impurity. From the ratio of the 39.76-keV peak to the Ruthenium K x rays, we determined a preliminary value for the ICC: αk(39.76) =134.6(19). This result agrees well with the theoretical calculation including the atomic vacancy, 135.2, and disagrees with the calculation excluding the vacancy, 127.4. This is consistent with our previous measurements, indicating that the atomic vacancy must be taken into account. Thanks to the NSF, DOE and Welch Foundation.

Quick release latch for reactor scram

DOEpatents

Johnson, Melvin L.; Shawver, Bruce M.

1976-01-01

A simple, reliable, and fast-acting means for releasing a control element and allowing it to be inserted rapidly into the core region of a nuclear reactor for scram purposes. A latch mechanism grips a coupling head on a nuclear control element to connect the control element to the control drive assembly. The latch mechanism is closed by tensioning a cable or rod with an actuator. The control element is released by de-energizing the actuator, providing fail-safe, rapid release of the control element to effect reactor shutdown. A sensing rod provides indication that the control element is properly positioned in the latch. Two embodiments are illustrated, one involving a collet-type latch mechanism, the other a pliers-type latch mechanism with the actuator located inside the reactor vessel.

BISON and MARMOT Development for Modeling Fast Reactor Fuel Performance

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

Gamble, Kyle Allan Lawrence; Williamson, Richard L.; Schwen, Daniel

2015-09-01

BISON and MARMOT are two codes under development at the Idaho National Laboratory for engineering scale and lower length scale fuel performance modeling. It is desired to add capabilities for fast reactor applications to these codes. The fast reactor fuel types under consideration are metal (U-Pu-Zr) and oxide (MOX). The cladding types of interest include 316SS, D9, and HT9. The purpose of this report is to outline the proposed plans for code development and provide an overview of the models added to the BISON and MARMOT codes for fast reactor fuel behavior. A brief overview of preliminary discussions on themore » formation of a bilateral agreement between the Idaho National Laboratory and the National Nuclear Laboratory in the United Kingdom is presented.« less

Quick release latch for reactor scram

DOEpatents

Johnson, M.L.; Shawver, B.M.

1975-09-16

A simple, reliable, and fast-acting means for releasing a control element and allowing it to be inserted rapidly into the core region of a nuclear reactor for scram purposes is described. A latch mechanism grips a coupling head on a nuclear control element to connect the control element to the control drive assembly. The latch mechanism is closed by tensioning a cable or rod with an actuator. The control element is released by de-energizing the actuator, providing fail-safe, rapid release of the control element to effect reactor shutdown. A sensing rod provides indication that the control element is properly positioned in the latch. Two embodiments are illustrated, one involving a collet- type latch mechanism, the other a pliers-type latch mechanism with the actuator located inside the reactor vessel. (auth)

A novel approach of solid waste management via aromatization using multiphase catalytic pyrolysis of waste polyethylene.

PubMed

Gaurh, Pramendra; Pramanik, Hiralal

2018-01-01

A new and innovative approach was adopted to increase the yield of aromatics like, benzene, toluene and xylene (BTX) in the catalytic pyrolysis of waste polyethylene (PE). The BTX content was significantly increased due to effective interaction between catalystZSM-5 and target molecules i.e., lower paraffins within the reactor. The thermal and catalytic pyrolysis both were performed in a specially designed semi-batch reactor at the temperature range of 500 °C-800 °C. Catalytic pyrolysis were performed in three different phases within the reactor batch by batch systematically, keeping the catalyst in A type- vapor phase, B type- liquid phase and C type- vapor and liquid phase (multiphase), respectively. Total aromatics (BTX) of 6.54 wt% was obtained for thermal pyrolysis at a temperature of 700 °C. In contrary, for the catalytic pyrolysis A, B and C types reactor arrangement, the aromatic (BTX) contents were progressively increased, nearly 6 times from 6.54 wt% (thermal pyrolysis) to 35.06 wt% for C-type/multiphase (liquid and vapor phase). The pyrolysis oil were characterized using GC-FID, FT-IR, ASTM distillation and carbon residue test to evaluate its end use and aromatic content. Copyright © 2017 Elsevier Ltd. All rights reserved.

Fail-safe reactivity compensation method for a nuclear reactor

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

Nygaard, Erik T.; Angelo, Peter L.; Aase, Scott B.

The present invention relates generally to the field of compensation methods for nuclear reactors and, in particular to a method for fail-safe reactivity compensation in solution-type nuclear reactors. In one embodiment, the fail-safe reactivity compensation method of the present invention augments other control methods for a nuclear reactor. In still another embodiment, the fail-safe reactivity compensation method of the present invention permits one to control a nuclear reaction in a nuclear reactor through a method that does not rely on moving components into or out of a reactor core, nor does the method of the present invention rely on themore » constant repositioning of control rods within a nuclear reactor in order to maintain a critical state.« less

HORIZONTAL BOILING REACTOR SYSTEM

DOEpatents

Treshow, M.

1958-11-18

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

REACTOR PHYSICS MODELING OF SPENT RESEARCH REACTOR FUEL FOR TECHNICAL NUCLEAR FORENSICS

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

Nichols, T.; Beals, D.; Sternat, M.

2011-07-18

Technical nuclear forensics (TNF) refers to the collection, analysis and evaluation of pre- and post-detonation radiological or nuclear materials, devices, and/or debris. TNF is an integral component, complementing traditional forensics and investigative work, to help enable the attribution of discovered radiological or nuclear material. Research is needed to improve the capabilities of TNF. One research area of interest is determining the isotopic signatures of research reactors. Research reactors are a potential source of both radiological and nuclear material. Research reactors are often the least safeguarded type of reactor; they vary greatly in size, fuel type, enrichment, power, and burn-up. Manymore » research reactors are fueled with highly-enriched uranium (HEU), up to {approx}93% {sup 235}U, which could potentially be used as weapons material. All of them have significant amounts of radiological material with which a radioactive dispersal device (RDD) could be built. Therefore, the ability to attribute if material originated from or was produced in a specific research reactor is an important tool in providing for the security of the United States. Currently there are approximately 237 operating research reactors worldwide, another 12 are in temporary shutdown and 224 research reactors are reported as shut down. Little is currently known about the isotopic signatures of spent research reactor fuel. An effort is underway at Savannah River National Laboratory (SRNL) to analyze spent research reactor fuel to determine these signatures. Computer models, using reactor physics codes, are being compared to the measured analytes in the spent fuel. This allows for improving the reactor physics codes in modeling research reactors for the purpose of nuclear forensics. Currently the Oak Ridge Research reactor (ORR) is being modeled and fuel samples are being analyzed for comparison. Samples of an ORR spent fuel assembly were taken by SRNL for analytical and radiochemical analysis. The fuel assembly was modeled using MONTEBURNS(MCNP5/ ORIGEN2.2) and MCNPX/CINDER90. The results from the models have been compared to each other and to the measured data.« less

Radioactive waste from decommissioning of fast reactors (through the example of BN-800)

NASA Astrophysics Data System (ADS)

Rybin, A. A.; Momot, O. A.

2017-01-01

Estimation of volume of radioactive waste from operating and decommissioning of fast reactors is introduced. Preliminary estimation has shown that the volume of RW from decommissioning of BN-800 is amounted to 63,000 cu. m. Comparison of the amount of liquid radioactive waste derived from operation of different reactor types is performed. Approximate costs of all wastes disposal for complete decommissioning of BN-800 reactor are estimated amounting up to approx. 145 million.

Properties of bio-oil generated by a pyrolysis of forest cedar residuals with the movable Auger-type reactor

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

Nishimura, Shun; Ebitani, Kohki, E-mail: ebitani@jaist.ac.jp; Miyazato, Akio

Our research project has developed the new movable reactor for bio-oil production in 2013 on the basis of Auger-type system. This package would be a great impact due to the concept of local production for local consumption in the hilly and mountainous area in not only Japan but also in the world. Herein, we would like to report the properties of the bio-oil generated by the developing Auger-type movable reactor. The synthesized bio-oil possessed C: 46.2 wt%, H: 6.5 wt%, N: wt%, S: <0.1 wt%, O: 46.8 wt% and H{sub 2}O: 18.4 wt%, and served a good calorific value ofmore » 18.1 MJ/kg. The spectroscopic and mass analyses such as FT-IR, GC-MS, {sup 13}C-NMR and FT-ICR MS supported that the bio-oil was composed by the fine mixtures of methoxy phenols and variety of alcohol or carboxylic acid functional groups. Thus, it is suggested that the bio-oil generated by the new movable Auger-type reactor has a significant potential as well as the existing bio-oil reported previously.« less

Station Blackout Analysis of HTGR-Type Experimental Power Reactor

NASA Astrophysics Data System (ADS)

Syarip; Zuhdi, Aliq; Falah, Sabilul

2018-01-01

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

CONTROL SYSTEM

DOEpatents

Shannon, R.H.; Williamson, H.E.

1962-10-30

A boiling water type nuclear reactor power system having improved means of control is described. These means include provisions for either heating the coolant-moderator prior to entry into the reactor or shunting the coolantmoderator around the heating means in response to the demand from the heat engine. These provisions are in addition to means for withdrawing the control rods from the reactor. (AEC)

Synthesis and cure kinetics of liquefied wood/phenol/formaldehyde resins

Treesearch

Hui Pan; Todd F. Shupe; Chung-Yun Hse

2008-01-01

Wood liquefaction was conducted at a 2/1 phenol/wood ratio in two different reactors: (1) an atmospheric three-necked flask reactor and (2) a sealed Parr reactor. The liquefied wood mixture (liquefied wood, unreacted phenol, and wood residue) was further condensed with formaldehyde under acidic conditions to synthesize two novolac-type liquefied wood/phenol/...

Immobilized biocatalytic process development and potential application in membrane separation: a review.

PubMed

Chakraborty, Sudip; Rusli, Handajaya; Nath, Arijit; Sikder, Jaya; Bhattacharjee, Chiranjib; Curcio, Stefano; Drioli, Enrico

2016-01-01

Biocatalytic membrane reactors have been widely used in different industries including food, fine chemicals, biological, biomedical, pharmaceuticals, environmental treatment and so on. This article gives an overview of the different immobilized enzymatic processes and their advantages over the conventional chemical catalysts. The application of a membrane bioreactor (MBR) reduces the energy consumption, and system size, in line with process intensification. The performances of MBR are considerably influenced by substrate concentration, immobilized matrix material, types of immobilization and the type of reactor. Advantages of a membrane associated bioreactor over a free-enzyme biochemical reaction, and a packed bed reactor are, large surface area of immobilization matrix, reuse of enzymes, better product recovery along with heterogeneous reactions, and continuous operation of the reactor. The present research work highlights immobilization techniques, reactor setup, enzyme stability under immobilized conditions, the hydrodynamics of MBR, and its application, particularly, in the field of sugar, starch, drinks, milk, pharmaceutical industries and energy generation.

Successive and large-scale synthesis of InP/ZnS quantum dots in a hybrid reactor and their application to white LEDs

NASA Astrophysics Data System (ADS)

Kim, Kyungnam; Jeong, Sohee; Woo, Ju Yeon; Han, Chang-Soo

2012-02-01

We report successive and large-scale synthesis of InP/ZnS core/shell nanocrystal quantum dots (QDs) using a customized hybrid flow reactor, which is based on serial combination of a batch-type mixer and a flow-type furnace. InP cores and InP/ZnS core/shell QDs were successively synthesized in the hybrid reactor in a simple one-step process. In this reactor, the flow rate of the solutions was typically 1 ml min-1, 100 times larger than that of conventional microfluidic reactors. In order to synthesize high-quality InP/ZnS QDs, we controlled both the flow rate and the crystal growth temperature. Finally, we obtained high-quality InP/ZnS QDs in colors from bluish green to red, and we demonstrated that these core/shell QDs could be incorporated into white-light-emitting diode (LED) devices to improve color rendering performance.

Successive and large-scale synthesis of InP/ZnS quantum dots in a hybrid reactor and their application to white LEDs.

PubMed

Kim, Kyungnam; Jeong, Sohee; Woo, Ju Yeon; Han, Chang-Soo

2012-02-17

We report successive and large-scale synthesis of InP/ZnS core/shell nanocrystal quantum dots (QDs) using a customized hybrid flow reactor, which is based on serial combination of a batch-type mixer and a flow-type furnace. InP cores and InP/ZnS core/shell QDs were successively synthesized in the hybrid reactor in a simple one-step process. In this reactor, the flow rate of the solutions was typically 1 ml min(-1), 100 times larger than that of conventional microfluidic reactors. In order to synthesize high-quality InP/ZnS QDs, we controlled both the flow rate and the crystal growth temperature. Finally, we obtained high-quality InP/ZnS QDs in colors from bluish green to red, and we demonstrated that these core/shell QDs could be incorporated into white-light-emitting diode (LED) devices to improve color rendering performance.

The use of experimental data in an MTR-type nuclear reactor safety analysis

NASA Astrophysics Data System (ADS)

Day, Simon E.

Reactivity initiated accidents (RIAs) are a category of events required for research reactor safety analysis. A subset of this is unprotected RIAs in which mechanical systems or human intervention are not credited in the response of the system. Light-water cooled and moderated MTR-type ( i.e., aluminum-clad uranium plate fuel) reactors are self-limiting up to some reactivity insertion limit beyond which fuel damage occurs. This characteristic was studied in the Borax and Spert reactor tests of the 1950s and 1960s in the USA. This thesis considers the use of this experimental data in generic MTR-type reactor safety analysis. The approach presented herein is based on fundamental phenomenological understanding and uses correlations in the reactor test data with suitable account taken for differences in important system parameters. Specifically, a semi-empirical approach is used to quantify the relationship between the power, energy and temperature rise response of the system as well as parametric dependencies on void coefficient and the degree of subcooling. Secondary effects including the dependence on coolant flow are also examined. A rigorous curve fitting approach and error assessment is used to quantify the trends in the experimental data. In addition to the initial power burst stage of an unprotected transient, the longer term stability of the system is considered with a stylized treatment of characteristic power/temperature oscillations (chugging). A bridge from the HEU-based experimental data to the LEU fuel cycle is assessed and outlined based on existing simulation results presented in the literature. A cell-model based parametric study is included. The results are used to construct a practical safety analysis methodology for determining reactivity insertion safety limits for a light-water moderated and cooled MTR-type core.

Key Assets for a Sustainable Low Carbon Energy Future

NASA Astrophysics Data System (ADS)

Carre, Frank

2011-10-01

Since the beginning of the 21st century, concerns of energy security and climate change gave rise to energy policies focused on energy conservation and diversified low-carbon energy sources. Provided lessons of Fukushima accident are evidently accounted for, nuclear energy will probably be confirmed in most of today's nuclear countries as a low carbon energy source needed to limit imports of oil and gas and to meet fast growing energy needs. Future challenges of nuclear energy are then in three directions: i) enhancing safety performance so as to preclude any long term impact of severe accident outside the site of the plant, even in case of hypothetical external events, ii) full use of Uranium and minimization long lived radioactive waste burden for sustainability, and iii) extension to non-electricity energy products for maximizing the share of low carbon energy source in transportation fuels, industrial process heat and district heating. Advanced LWRs (Gen-III) are today's best available technologies and can somewhat advance nuclear energy in these three directions. However, breakthroughs in sustainability call for fast neutron reactors and closed fuel cycles, and non-electric applications prompt a revival of interest in high temperature reactors for exceeding cogeneration performances achievable with LWRs. Both types of Gen-IV nuclear systems by nature call for technology breakthroughs to surpass LWRs capabilities. Current resumption in France of research on sodium cooled fast neutron reactors (SFRs) definitely aims at significant progress in safety and economic competitiveness compared to earlier reactors of this type in order to progress towards a new generation of commercially viable sodium cooled fast reactor. Along with advancing a new generation of sodium cooled fast reactor, research and development on alternative fast reactor types such as gas or lead-alloy cooled systems (GFR & LFR) is strategic to overcome technical difficulties and/or political opposition specific to sodium. In conclusion, research and technology breakthroughs in nuclear power are needed for shaping a sustainable low carbon future. International cooperation is key for sharing costs of research and development of the required novel technologies and cost of first experimental reactors needed to demonstrate enabling technologies. At the same time technology breakthroughs are developed, pre-normative research is required to support codification work and harmonized regulations that will ultimately apply to safety and security features of resulting innovative reactor types and fuel cycles.

NEUTRONIC REACTORS

DOEpatents

Wigner, E.P.; Young, G.J.

1958-10-14

A method is presented for loading and unloading rod type fuel elements of a neutronic reactor of the heterogeneous, solld moderator, liquid cooled type. In the embodiment illustrated, the fuel rods are disposed in vertical coolant channels in the reactor core. The fuel rods are loaded and unloaded through the upper openings of the channels which are immersed in the coolant liquid, such as water. Unloading is accomplished by means of a coffer dam assembly having an outer sleeve which is placed in sealing relation around the upper opening. A radiation shield sleeve is disposed in and reciprocable through the coffer dam sleeve. A fuel rod engaging member operates through the axial bore in the radiation shield sleeve to withdraw the fuel rod from its position in the reactor coolant channel into the shield, the shield snd rod then being removed. Loading is accomplished in the reverse procedure.

Computational and Experimental Investigations of the Coolant Flow in the Cassette Fissile Core of a KLT-40S Reactor

NASA Astrophysics Data System (ADS)

Dmitriev, S. M.; Varentsov, A. V.; Dobrov, A. A.; Doronkov, D. V.; Pronin, A. N.; Sorokin, V. D.; Khrobostov, A. E.

2017-07-01

Results of experimental investigations of the local hydrodynamic and mass-exchange characteristics of a coolant flowing through the cells in the characteristic zones of a fuel assembly of a KLT-40S reactor plant downstream of a plate-type spacer grid by the method of diffusion of a gas tracer in the coolant flow with measurement of its velocity by a five-channel pneumometric probe are presented. An analysis of the concentration distribution of the tracer in the coolant flow downstream of a plate-type spacer grid in the fuel assembly of the KLT-40S reactor plant and its velocity field made it possible to obtain a detailed pattern of this flow and to determine its main mechanisms and features. Results of measurement of the hydraulic-resistance coefficient of a plate-type spacer grid depending on the Reynolds number are presented. On the basis of the experimental data obtained, recommendations for improvement of the method of calculating the flow rate of a coolant in the cells of the fissile core of a KLT-40S reactor were developed. The results of investigations of the local hydrodynamic and mass-exchange characteristics of the coolant flow in the fuel assembly of the KLT-40S reactor plant were accepted for estimating the thermal and technical reliability of the fissile cores of KLT-40S reactors and were included in the database for verification of computational hydrodynamics programs (CFD codes).

The procedure and results of calculations of the equilibrium isotopic composition of a demonstration subcritical molten salt reactor

NASA Astrophysics Data System (ADS)

Nevinitsa, V. A.; Dudnikov, A. A.; Blandinskiy, V. Yu.; Balanin, A. L.; Alekseev, P. N.; Titarenko, Yu. E.; Batyaev, V. F.; Pavlov, K. V.; Titarenko, A. Yu.

2015-12-01

A subcritical molten salt reactor with an external neutron source is studied computationally as a facility for incineration and transmutation of minor actinides from spent nuclear fuel of reactors of VVER-1000 type and for producing 233U from 232Th. The reactor configuration is chosen, the requirements to be imposed on the external neutron source are formulated, and the equilibrium isotopic composition of heavy nuclides and the key parameters of the fuel cycle are calculated.

MTR, TRA603. FIRST FLOOR PLAN. REACTOR AT CENTER. TWENTYMETER CHOPPER ...

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

MTR, TRA-603. FIRST FLOOR PLAN. REACTOR AT CENTER. TWENTY-METER CHOPPER HOUSE. COFFIN TURNING ROLLS. REMOVABLE PANEL OVER CANAL ON EAST SIDE. NEW PLUG STORAGE ACCESS. DOOR SCHEDULE INDICATES STEEL (FOR VAULT), WIRE MESH, AND HOLLOW METAL TYPES. STORAGE AND ISSUE ROOM. SAFETY SHOWERS. DOORWAY TO WING, TRA-604. BLAW-KNOX 3150-803-2, 7/1950. INL INDEX NO. 531-0603-00-098-100561, REV. 10. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Monte Carlo Shielding Comparative Analysis Applied to TRIGA HEU and LEU Spent Fuel Transport

NASA Astrophysics Data System (ADS)

Margeanu, C. A.; Margeanu, S.; Barbos, D.; Iorgulis, C.

2010-12-01

The paper is a comparative study of LEU and HEU fuel utilization effects for the shielding analysis during spent fuel transport. A comparison against the measured data for HEU spent fuel, available from the last stage of spent fuel repatriation fulfilled in the summer of 2008, is also presented. All geometrical and material data for the shipping cask were considered according to NAC-LWT Cask approved model. The shielding analysis estimates radiation doses to shipping cask wall surface, and in air at 1 m and 2 m, respectively, from the cask, by means of 3D Monte Carlo MORSE-SGC code. Before loading into the shipping cask, TRIGA spent fuel source terms and spent fuel parameters have been obtained by means of ORIGEN-S code. Both codes are included in ORNL's SCALE 5 programs package. The actinides contribution to total fuel radioactivity is very low in HEU spent fuel case, becoming 10 times greater in LEU spent fuel case. Dose rates for both HEU and LEU fuel contents are below regulatory limits, LEU spent fuel photon dose rates being greater than HEU ones. Comparison between HEU spent fuel theoretical and measured dose rates in selected measuring points shows a good agreement, calculated values being greater than the measured ones both to cask wall surface (about 34% relative difference) and in air at 1 m distance from cask surface (about 15% relative difference).

Dumbo: A pachydermal rocket motor

NASA Technical Reports Server (NTRS)

Kirk, Bill

1991-01-01

A brief historical account is given of the Dumbo nuclear reactor, a type of folded flow reactor that could be used for rocket propulsion. Much of the information is given in viewgraph form. Viewgraphs show details of the reactor system, fuel geometry, and key characteristics of the system (folded flow, use of fuel washers, large flow area, small fuel volume, hybrid modulator, and cermet fuel).

10 CFR 70.20b - General license for carriers of transient shipments of formula quantities of strategic special...

Code of Federal Regulations, 2010 CFR

2010-01-01

... significance, special nuclear material of low strategic significance, and irradiated reactor fuel. 70.20b..., special nuclear material of low strategic significance, and irradiated reactor fuel. (a) A general license... requirements of § 73.67 of this chapter. (3) Irradiated reactor fuel of the type and quantity subject to the...

10 CFR 70.20b - General license for carriers of transient shipments of formula quantities of strategic special...

Code of Federal Regulations, 2011 CFR

2011-01-01

... significance, special nuclear material of low strategic significance, and irradiated reactor fuel. 70.20b..., special nuclear material of low strategic significance, and irradiated reactor fuel. (a) A general license... requirements of § 73.67 of this chapter. (3) Irradiated reactor fuel of the type and quantity subject to the...

Thorium and Molten Salt Reactors: "Essential Questions for Classroom Discussions"

ERIC Educational Resources Information Center

DiLisi, Gregory A.; Hirsch, Allison; Murray, Meredith; Rarick, Richard

2018-01-01

A little-known type of nuclear reactor called the "molten salt reactor" (MSR), in which nuclear fuel is dissolved in a liquid carrier salt, was proposed in the 1940s and developed at the Oak Ridge National Laboratory in the 1960s. Recently, the MSR has generated renewed interest as a remedy for the drawbacks associated with conventional…

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

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

NASA Astrophysics Data System (ADS)

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

1997-09-01

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

High throughput semiconductor deposition system

DOEpatents

Young, David L.; Ptak, Aaron Joseph; Kuech, Thomas F.; Schulte, Kevin; Simon, John D.

2017-11-21

A reactor for growing or depositing semiconductor films or devices. The reactor may be designed for inline production of III-V materials grown by hydride vapor phase epitaxy (HVPE). The operating principles of the HVPE reactor can be used to provide a completely or partially inline reactor for many different materials. An exemplary design of the reactor is shown in the attached drawings. In some instances, all or many of the pieces of the reactor formed of quartz, such as welded quartz tubing, while other reactors are made from metal with appropriate corrosion resistant coatings such as quartz or other materials, e.g., corrosion resistant material, or stainless steel tubing or pipes may be used with a corrosion resistant material useful with HVPE-type reactants and gases. Using HVPE in the reactor allows use of lower-cost precursors at higher deposition rates such as in the range of 1 to 5 .mu.m/minute.

Reactivity control assembly for nuclear reactor. [LMFBR

DOEpatents

Bollinger, L.R.

1982-03-17

This invention, which resulted from a contact with the United States Department of Energy, relates to a control mechanism for a nuclear reactor and, more particularly, to an assembly for selectively shifting different numbers of reactivity modifying rods into and out of the core of a nuclear reactor. It has been proposed heretofore to control the reactivity of a breeder reactor by varying the depth of insertion of control rods (e.g., rods containing a fertile material such as ThO/sub 2/) in the core of the reactor, thereby varying the amount of neutron-thermalizing coolant and the amount of neutron-capturing material in the core. This invention relates to a mechanism which can advantageously be used in this type of reactor control system.

The effect of the composition of plutonium loaded on the reactivity change and the isotopic composition of fuel produced in a fast reactor

NASA Astrophysics Data System (ADS)

Blandinskiy, V. Yu.

2014-12-01

This paper presents the results of a numerical investigation into burnup and breeding of nuclides in metallic fuel consisting of a mixture of plutonium and depleted uranium in a fast reactor with sodium coolant. The feasibility of using plutonium contained in spent nuclear fuel from domestic thermal reactors and weapons-grade plutonium is discussed. It is shown that the largest production of secondary fuel and the least change in the reactivity over the reactor lifetime can be achieved when employing plutonium contained in spent nuclear fuel from a reactor of the RBMK-1000 type.

Radiotoxicity and decay heat power of spent nuclear fuel of VVER type reactors at long-term storage.

PubMed

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

2005-01-01

Radiotoxicity and decay heat power of the spent nuclear fuel of VVER-1000 type reactors are calculated during storage time up to 300,000 y. Decay heat power of radioactive waste (radwaste) determines parameters of the heat removal system for the safe storage of spent nuclear fuel. Radiotoxicity determines the radiological hazard of radwaste after its leakage and penetration into the environment.

Solid0Core Heat-Pipe Nuclear Batterly Type Reactor

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

Ehud Greenspan

This project was devoted to a preliminary assessment of the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor to have a solid core from which heat is removed by liquid-metal heat pipes (HP).

Thermal breeder fuel enrichment zoning

DOEpatents

Capossela, Harry J.; Dwyer, Joseph R.; Luce, Robert G.; McCoy, Daniel F.; Merriman, Floyd C.

1992-01-01

A method and apparatus for improving the performance of a thermal breeder reactor having regions of higher than average moderator concentration are disclosed. The fuel modules of the reactor core contain at least two different types of fuel elements, a high enrichment fuel element and a low enrichment fuel element. The two types of fuel elements are arranged in the fuel module with the low enrichment fuel elements located between the high moderator regions and the high enrichment fuel elements. Preferably, shim rods made of a fertile material are provided in selective regions for controlling the reactivity of the reactor by movement of the shim rods into and out of the reactor core. The moderation of neutrons adjacent the high enrichment fuel elements is preferably minimized as by reducing the spacing of the high enrichment fuel elements and/or using a moderator having a reduced moderating effect.

Next generation fuel irradiation capability in the High Flux Reactor Petten

NASA Astrophysics Data System (ADS)

Fütterer, Michael A.; D'Agata, Elio; Laurie, Mathias; Marmier, Alain; Scaffidi-Argentina, Francesco; Raison, Philippe; Bakker, Klaas; de Groot, Sander; Klaassen, Frodo

2009-07-01

This paper describes selected equipment and expertise on fuel irradiation testing at the High Flux Reactor (HFR) in Petten, The Netherlands. The reactor went critical in 1961 and holds an operating license up to at least 2015. While HFR has initially focused on Light Water Reactor fuel and materials, it also played a decisive role since the 1970s in the German High Temperature Reactor (HTR) development program. A variety of tests related to fast reactor development in Europe were carried out for next generation fuel and materials, in particular for Very High Temperature Reactor (V/HTR) fuel, fuel for closed fuel cycles (U-Pu and Th-U fuel cycle) and transmutation, as well as for other innovative fuel types. The HFR constitutes a significant European infrastructure tool for the development of next generation reactors. Experimental facilities addressed include V/HTR fuel tests, a coated particle irradiation rig, and tests on fast reactor, transmutation and thorium fuel. The rationales for these tests are given, results are provided and further work is outlined.

Using thermal balance model to determine optimal reactor volume and insulation material needed in a laboratory-scale composting reactor.

PubMed

Wang, Yongjiang; Pang, Li; Liu, Xinyu; Wang, Yuansheng; Zhou, Kexun; Luo, Fei

2016-04-01

A comprehensive model of thermal balance and degradation kinetics was developed to determine the optimal reactor volume and insulation material. Biological heat production and five channels of heat loss were considered in the thermal balance model for a representative reactor. Degradation kinetics was developed to make the model applicable to different types of substrates. Simulation of the model showed that the internal energy accumulation of compost was the significant heat loss channel, following by heat loss through reactor wall, and latent heat of water evaporation. Lower proportion of heat loss occurred through the reactor wall when the reactor volume was larger. Insulating materials with low densities and low conductive coefficients were more desirable for building small reactor systems. Model developed could be used to determine the optimal reactor volume and insulation material needed before the fabrication of a lab-scale composting system. Copyright © 2016 Elsevier Ltd. All rights reserved.

NEUTRONIC REACTOR SYSTEM

DOEpatents

Treshow, M.

1959-02-10

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

The IRIS Spool-Type Reactor Coolant Pump

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

Kujawski, J.M.; Kitch, D.M.; Conway, L.E.

2002-07-01

IRIS (International Reactor Innovative and Secure) is a light water cooled, 335 MWe power reactor which is being designed by an international consortium as part of the US DOE NERI Program. IRIS features an integral reactor vessel that contains all the major reactor coolant system components including the reactor core, the coolant pumps, the steam generators and the pressurizer. This integral design approach eliminates the large coolant loop piping, and thus eliminates large loss-of-coolant accidents (LOCAs) as well as the individual component pressure vessels and supports. In addition, IRIS is being designed with a long life core and enhanced safetymore » to address the requirements defined by the US DOE for Generation IV reactors. One of the innovative features of the IRIS design is the adoption of a reactor coolant pump (called 'spool' pump) which is completely contained inside the reactor vessel. Background, status and future developments of the IRIS spool pump are presented in this paper. (authors)« less

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

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

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

Today, ISIS research reactor is an essential tool for Education and Training programs organized by the National Institute for Nuclear Science and Technology (INSTN) from CEA. In the field of nuclear instrumentation, the INSTN offers both, theoretical courses and training courses on the use of neutron detection systems taking advantage of the ISIS research reactor for the supply of a wide range of neutron fluxes. This paper describes the content of the training carried out on the use of neutron detectors and detection systems, on-site or remote. The ISIS reactor is a 700 kW open core pool type reactor. Themore » facility is very flexible since neutron detectors can be inserted into the core or its vicinity, and be used at different levels of power according to the needs of the course. Neutron fluxes, typically ranging from 1 to 10{sup 12} n/cm{sup 2}.s, can be obtained for the characterisation of the neutron detectors and detection systems. For the monitoring of the neutron density at low level of power, the Instrumentation and Control (I and C) system of the reactor is equipped with two detection systems, named BN1 and BN2. Each way contains a fission chamber, type CFUL01, connected to an electronic system type SIREX.The system works in pulse mode and exhibits two outputs: the counting rate and the doubling time. For the high level of power, the I and C is equipped with two detection systems HN1 and HN2.Each way contain a boron ionization chamber (type CC52) connected to an electronics system type SIREX. The system works in current mode and has two outputs: the current and the doubling time. For each mode, the trainees can observe and measure the signal at the different stages of the electronic system, with an oscilloscope. They can understand the role of each component of the detection system: detector, cable and each electronic block. The limitation of the detection modes and their operating range can be established from the measured signal. The trainees can also modify the settings of the electronic system, such as the high voltage and the discrimination level in order to obtain all the characteristic curves of the detectors. These curves are used to define the right setting of the electronic system and to discuss the expected degradation of the detector signal resulting from the detector damage under the integrated neutron and gamma fluxes. Moreover, in addition to the study of the neutron detection systems itself, the integration of the measurements made by these detection systems in the logic of the safety system of the nuclear reactor is also addressed. Providing the trainees with an extensive overview of each part of the neutron monitoring instrumentation apply to a nuclear reactor, hands-on measurements on the ISIS reactor play a major role in ensuring a practical and comprehensive understanding of the neutron detection system and their integration in the safety system of nuclear reactors. It also gives a solid background for the follow up and the development of the neutron detection systems. In addition to on-reactor training, Internet Reactor Laboratory capability has been implemented on the ISIS reactor in 2014. For the Internet Reactor Laboratory an extensive video conference system has been implemented on ISIS reactor. The system includes 4 cameras and the transmission of the video signal given by the supervision system of the reactor which records and processes the data of the reactor. According to the pedagogic needs during the training courses, the lecturer on the ISIS reactor chooses to broadcast the relevant information at each stage of the course. For example, graph showing the histogram of the counting and current as a function of the time, or the electrical signal observed on the oscilloscope, can be broadcasted trough internet. By interacting through the video conference, the remote classroom is able to ask for changes in the reactor power or settings of the detection systems. They can also ask for the broadcast of some particular information. At the guest institution, the information is displayed in two parts or screens, as shown in the Figure 3. Concerning the interaction with - and the feedback from - the remote classroom, the camera of the video system in the remote classroom is used to ensure the contact between the trainees and the lecturer and reactor operators. Thus, the Internet Reactor Laboratory is complementary to the on reactor training courses. It allows distant learning, reducing the overall cost of the course when this is necessary. It can efficiently be used for the development of the human resources needed by the nuclear industry and the nuclear programs in countries without research reactors.« less

In-vessel composting of household wastes

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

Iyengar, Srinath R.; Bhave, Prashant P.

The process of composting has been studied using five different types of reactors, each simulating a different condition for the formation of compost; one of which was designed as a dynamic complete-mix type household compost reactor. A lab-scale study was conducted first using the compost accelerators culture (Trichoderma viridae, Trichoderma harzianum, Trichorus spirallis, Aspergillus sp., Paecilomyces fusisporus, Chaetomium globosum) grown on jowar (Sorghum vulgare) grains as the inoculum mixed with cow-dung slurry, and then by using the mulch/compost formed in the respective reactors as the inoculum. The reactors were loaded with raw as well as cooked vegetable waste for amore » period of 4 weeks and then the mulch formed was allowed to maturate. The mulch was analysed at various stages for the compost and other environmental parameters. The compost from the designed aerobic reactor provides good humus to build up a poor physical soil and some basic plant nutrients. This proves to be an efficient, eco-friendly, cost-effective, and nuisance-free solution for the management of household solid wastes.« less

The procedure and results of calculations of the equilibrium isotopic composition of a demonstration subcritical molten salt reactor

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

Nevinitsa, V. A., E-mail: Neviniza-VA@nrcki.ru; Dudnikov, A. A.; Blandinskiy, V. Yu.

2015-12-15

A subcritical molten salt reactor with an external neutron source is studied computationally as a facility for incineration and transmutation of minor actinides from spent nuclear fuel of reactors of VVER-1000 type and for producing {sup 233}U from {sup 232}Th. The reactor configuration is chosen, the requirements to be imposed on the external neutron source are formulated, and the equilibrium isotopic composition of heavy nuclides and the key parameters of the fuel cycle are calculated.

NEUTRONIC REACTOR FUEL ELEMENT

DOEpatents

Picklesimer, M.L.; Thurber, W.C.

1961-01-01

A chemically nonreactive fuel composition for incorporation in aluminum- clad, plate type fuel elements for neutronic reactors is described. The composition comprises a mixture of aluminum and uranium carbide particles, the uranium carbide particles containing at least 80 wt.% UC/sub 2/.

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

Comparative performance of fixed-film biological filters: Application of reactor theory

USGS Publications Warehouse

Watten, B.J.; Sibrell, P.L.

2006-01-01

Nitrification is classified as a two-step consecutive reaction where R1 represents the rate of formation of the intermediate product NO2-N and R2 represents the rate of formation of the final product NO3-N. The relative rates of R1 and R2 are influenced by reactor type characterized hydraulically as plug-flow, plug-flow with dispersion and mixed-flow. We develop substrate conversion models for fixed-film biofilters operating in the first-order kinetic regime based on application of chemical reactor theory. Reactor type, inlet conditions and the biofilm kinetic constants Ki (h-1) are used to predict changes in NH4-N, NO2-N, NO3-N and BOD5. The inhibiting effects of the latter on R1 and R2 were established based on the ?? relation, e.g.:{A formula is presented}where BOD5,max is the concentration that causes nitrification to cease and N is a variable relating Ki to increasing BOD5. Conversion models were incorporated in spreadsheet programs that provided steady-state concentrations of nitrogen and BOD5 at several points in a recirculating aquaculture system operating with input values for fish feed rate, reactor volume, microscreen performance, make-up and recirculating flow rates. When rate constants are standardized, spreadsheet use demonstrates plug-flow reactors provide higher rates of R1 and R2 than mixed-flow reactors thereby reducing volume requirements for target concentrations of NH4-N and NO2-N. The benefit provided by the plug-flow reactor varies with hydraulic residence time t as well as the effective vessel dispersion number, D/??L. Both reactor types are capable of providing net increases in NO2-N during treatment but the rate of decrease in the mixed-flow case falls well behind that predicted for plug-flow operation. We show the potential for a positive net change in NO2-N increases with decreases in the dimensionless ratios K2, (R2 )/K1,( R1 ) and [NO2-N]/[NH4-N] and when the product K1, (R1) t provides low to moderate NH4-N conversions. Maintaining high levels of the latter reduces the effective reactor utilization rate (%) defined here as (RNavg/RNmax)100 where RNavg is the mean reactive nitrogen concentration ([NH4-N] + [NO2-N]) within the reactor, and RNmax represents the feed concentration of the same. Low utilization rates provide a hedge against unexpected increases in substrate loading and reduce water pumping requirements but force use of elevated reactor volumes. Further ?? effects on R1 and R2 can be reduced through use of a tanks-in-series versus a single mixed-flow reactor configuration and by improving the solids removal efficiency of microscreen treatment.

Effect of mechanical disruption on the effectiveness of three reactors used for dilute acid pretreatment of corn stover Part 2: morphological and structural substrate analysis

PubMed Central

2014-01-01

Background Lignocellulosic biomass is a renewable, naturally mass-produced form of stored solar energy. Thermochemical pretreatment processes have been developed to address the challenge of biomass recalcitrance, however the optimization, cost reduction, and scalability of these processes remain as obstacles to the adoption of biofuel production processes at the industrial scale. In this study, we demonstrate that the type of reactor in which pretreatment is carried out can profoundly alter the micro- and nanostructure of the pretreated materials and dramatically affect the subsequent efficiency, and thus cost, of enzymatic conversion of cellulose. Results Multi-scale microscopy and quantitative image analysis was used to investigate the impact of different biomass pretreatment reactor configurations on plant cell wall structure. We identify correlations between enzymatic digestibility and geometric descriptors derived from the image data. Corn stover feedstock was pretreated under the same nominal conditions for dilute acid pretreatment (2.0 wt% H2SO4, 160°C, 5 min) using three representative types of reactors: ZipperClave® (ZC), steam gun (SG), and horizontal screw (HS) reactors. After 96 h of enzymatic digestion, biomass treated in the SG and HS reactors achieved much higher cellulose conversions, 88% and 95%, respectively, compared to the conversion obtained using the ZC reactor (68%). Imaging at the micro- and nanoscales revealed that the superior performance of the SG and HS reactors could be explained by reduced particle size, cellular dislocation, increased surface roughness, delamination, and nanofibrillation generated within the biomass particles during pretreatment. Conclusions Increased cellular dislocation, surface roughness, delamination, and nanofibrillation revealed by direct observation of the micro- and nanoscale change in accessibility explains the superior performance of reactors that augment pretreatment with physical energy. PMID:24690534

JPRS Report, Science & Technology, China: Energy

DTIC Science & Technology

1988-06-29

capacity. There are currently two types of HTGR reactor designs: the particle-bed core , which uses spherical fuel elements, and the rod type core , in...and trial operating experience with the HTGR reactor. Its main design features are as follows. 1. A particle-bed core , continuous fueling and...Favorable for Development of Small-Scale HTGR (Xu Jiming; HE DONGLI GONGCHENG, Feb 88) 47 ERRATUM: In JPRS-CEN-88-003 of 25 April 1988 in article

Flow photochemistry: Old light through new windows

PubMed Central

Knowles, Jonathan P; Elliott, Luke D

2012-01-01

Summary Synthetic photochemistry carried out in classic batch reactors has, for over half a century, proved to be a powerful but under-utilised technique in general organic synthesis. Recent developments in flow photochemistry have the potential to allow this technique to be applied in a more mainstream setting. This review highlights the use of flow reactors in organic photochemistry, allowing a comparison of the various reactor types to be made. PMID:23209538

Status report on the fusion breeder

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

Moir, R.W.

1980-12-12

The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of /sup 233/U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW m/sup -2/, and the hybrid should cost lessmore » than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are unusually rapid.« less

A mini-cavity probe reactor.

NASA Technical Reports Server (NTRS)

Hyland, R. E.

1971-01-01

The mini-cavity reactor is a rocket engine concept which combines the high specific impulse from a central gaseous fueled cavity (0.6 m diam) and NERVA type fuel elements in a driver region that is external to a moderator-reflector zone to produce a compact light weight reactor. The overall dimension including a pressure vessel that is located outside of the spherical reactor is approximately 1.21 m in diameter. Specific impulses up to 2000 sec are obtainable for 220 to 890 N of thrust with pressures less than 1000 atm. Powerplant weights including a radiator for disposing of the power in the driver region are between 4600 and 32,000 kg - less than payloads of the shuttle. This reactor could also be used as a test reactor for gas-core, MHD, breeding and materials research.

Bio-hydrogen production from molasses by anaerobic fermentation in continuous stirred tank reactor

NASA Astrophysics Data System (ADS)

Han, Wei; Li, Yong-feng; Chen, Hong; Deng, Jie-xuan; Yang, Chuan-ping

2010-11-01

A study of bio-hydrogen production was performed in a continuous flow anaerobic fermentation reactor (with an available volume of 5.4 L). The continuous stirred tank reactor (CSTR) for bio-hydrogen production was operated under the organic loading rates (OLR) of 8-32 kg COD/m3 reactor/d (COD: chemical oxygen demand) with molasses as the substrate. The maximum hydrogen production yield of 8.19 L/d was obtained in the reactor with the OLR increased from 8 kg COD/m3 reactor/d to 24 kg COD/m3 d. However, the hydrogen production and volatile fatty acids (VFAs) drastically decreased at an OLR of 32 kg COD/m3 reactor/d. Ethanoi, acetic, butyric and propionic were the main liquid fermentation products with the percentages of 31%, 24%, 20% and 18%, which formed the mixed-type fermentation.

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

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

NASA Astrophysics Data System (ADS)

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

2006-01-01

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

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

NUCLEAR REACTOR

DOEpatents

Miller, H.I.; Smith, R.C.

1958-01-21

This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

Compton suppression method and epithermal NAA in the determination of nutrients and heavy metals in Nigerian food and beverages.

PubMed

Ahmed, Y A; Landsberger, S; O'Kelly, D J; Braisted, J; Gabdo, H; Ewa, I O B; Umar, I M; Funtua, I I

2010-10-01

We used in this study Compton suppression method and epithermal neutron activation analysis to determine the concentration of nutrients and heavy metals in Nigerian food and beverages. The work was performed at the University of Texas TRIGA Reactor by short, medium, and long irradiation protocols, using thermal flux of 1.4x10(12)n cm(-2)s(-1) and epithermal flux of 1.4x10(11)n cm(-2)s(-1). Application of Compton suppression method has reduced interferences from Compton scattered photons thereby allowing easy evaluation of Na, Cl, Ca, Cu, Mn, Mg, Co, Cr, Rb, Fe, and Se. The epithermal NAA method has enabled determination of Cd, As, Ba, Sr, Br, I, and V with little turn-around time. Quality Control and Quality Assurance of the method was tested by analyzing four Standard Reference Materials (non-fat powdered milk, apple leaves, citrus leaves, and peach leaves) obtained from National Institute for Standards and Technology. Our results show that sorghum, millet, and maize have high values of Zn, Mn, Fe, low values of Cd, As, and Se. Powdered milks, rice, beans, and soybeans were found to have moderate amounts of all the elements. Tobacco recorded high content of Cd, Mn, and As, whereas tea, tsobo leaves, Baobab leaves, and okro seed have more As values than others. However, biscuits, macaroni, spaghetti, and noodles show lower concentrations of all the elements. The distribution of these nutrients and heavy metals in these food and beverages shows the need to fortify biscuits and pastas with micro and macro-nutrients and reduce the use of tobacco, tea, tsobo leaves, Baobab leaves, and Okro seed to avoid intake of heavy elements. Copyright 2010 Elsevier Ltd. All rights reserved.

Experimental demonstration of radiation effects on the performance of a stirling-alternator convertor and candidate materials evaluation

NASA Astrophysics Data System (ADS)

Mireles, Omar R.

Free-piston Stirling power convertors are under consideration by NASA for service in the Advanced Stirling Radioisotope Generator (ASRG) and Fission Surface Power (FSP) systems to enable aggressive exploration missions by providing a reliable and constant power supply. The ASRG must withstand environmental radiation conditions, while the FSP system must tolerate a mixed neutron and gamma-ray environment resulting from self-irradiation. Stirling-alternators utilize rare earth magnets and a variety of organic materials whose radiation limits dominate service life estimates and shielding requirements. The project objective was to demonstrate the performance of the alternator, identify materials that exhibit excessive radiation sensitivity, identify radiation tolerant substitutes, establish empirical dose limits, and demonstrate the feasibility of cost effective nuclear and radiation tests by selection of the appropriate personnel and test facilities as a function of hardware maturity. The Stirling Alternator Radiation Test Article (SARTA) was constructed from linear alternator components of a Stirling convertor and underwent significant pre-exposure characterization. The SARTA was operated at the Sandia National Laboratories Gamma Irradiation Facility to a dose of over 40 Mrad. Operating performance was within nominal variation, although modestly decreasing trends occurred in later runs as well as the detection of an electrical fault after the final exposure. Post-irradiation disassembly and internal inspection revealed minimal degradation of the majority of the organic components. Radiation testing of organic material coupons was conducted since the majority of the literature was inconsistent. These inconsistencies can be attributed to testing at environmental conditions vastly different than those Stirling-alternator organics will experience during operation. Samples were irradiated at the Texas A&M TRIGA reactor to above expected FSP neutron fluence. A thorough materials evaluation followed and results indicate that the majority of material properties experienced minimal statistically significant change.

Device fabrication, characterization, and thermal neutron detection response of LiZnP and LiZnAs semiconductor devices

NASA Astrophysics Data System (ADS)

Montag, Benjamin W.; Ugorowski, Philip B.; Nelson, Kyle A.; Edwards, Nathaniel S.; McGregor, Douglas S.

2016-11-01

Nowotny-Juza compounds continue to be explored as candidates for solid-state neutron detectors. Such a device would have greater efficiency, in a compact form, than present day gas-filled 3He and 10BF3 detectors. The 6Li(n,t)4He reaction yields a total Q-value of 4.78 MeV, larger than 10B, an energy easily identified above background radiations. Hence, devices fabricated from semiconductor compounds having either natural Li (nominally 7.5% 6Li) or enriched 6Li (usually 95% 6Li) as constituent atoms may provide a material for compact high efficiency neutron detectors. Starting material was synthesized by preparing equimolar portions of Li, Zn, and As sealed under vacuum (10-6 Torr) in quartz ampoules lined with boron nitride and subsequently reacted in a compounding furnace [1]. The raw synthesized material indicated the presence high impurity levels (material and electrical property characterizations). A static vacuum sublimation in quartz was performed to help purify the synthesized material [2,3]. Bulk crystalline samples were grown from the purified material [4,5]. Samples were cut using a diamond wire saw, and processed into devices. Bulk resistivity was determined from I-V curve measurements, ranging from 106-1011 Ω cm. Devices were characterized for sensitivity to 5.48 MeV alpha particles, 337 nm laser light, and neutron sensitivity in a thermal neutron diffracted beam at the Kansas State University TRIGA Mark II nuclear reactor. Thermal neutron reaction product charge induction was measured with a LiZnP device, and the reaction product spectral response was observed.

Using the nuclear activation AMS method for determining chlorine in solids at ppb-levels and below

NASA Astrophysics Data System (ADS)

Winkler, Stephan R.; Eigl, Rosmarie; Forstner, Oliver; Martschini, Martin; Steier, Peter; Sterba, Johannes H.; Golser, Robin

2015-10-01

Neutron activation analysis using decay counting of the activated element is a well-established method in elemental analysis. However, for chlorine there is a better alternative to measuring decay of the short-lived activation product chlorine-38 (t1/2 = 37.24 min) - accelerator mass spectrometry (AMS) of 36Cl: the relatively high neutron capture cross section of chlorine-35 for thermal neutrons (43.7 b) and combined the AMS technique for chlorine-36 (t1/2 = 301 ka) allow for determination of chlorine down to ppb-levels using practical sample sizes and common exposure durations. The combination of neutron activation and AMS can be employed for a few other elements (nitrogen, thorium, and uranium) as well. For bulk solid samples an advantage of the method is that lab contamination can be rendered irrelevant. The chlorine-35 in the sample is activated to chlorine-36, and surface chlorine can be removed after the irradiation. Subsequent laboratory contamination, however, will not carry a prominent chlorine-36 signature. After sample dissolution and addition of sufficient amounts of stable chlorine carrier the produced chlorine-36 and thus the original chlorine-35 of the sample can be determined using AMS. We have developed and applied the method for analysis of chlorine in steel samples. The chlorine content of steel is of interest to nuclear industry, precisely because of above mentioned high neutron capture cross section for chlorine-35, which leads to accumulation of chlorine-36 as long-term nuclear waste. The samples were irradiated at the TRIGA Mark II reactor of the Atominstitut in Vienna and the 36Cl-AMS setup at the Vienna Environmental Research Accelerator (VERA) was used for 36Cl/Cl analysis.

Development and performance of an alternative biofilter system.

PubMed

Lee, D H; Lau, A K; Pinder, K L

2001-01-01

Step tracer tests were carried out on lab-scale biofilters to determine the residence time distributions (RTDs) of gases passing through two types of biofilters: a standard biofilter with vertical gas flow and a modified biofilter with horizontal gas flow. Results were used to define the flow patterns in the reactors. "Non-ideal flow" indicates that the flow reactors did not behave like either type of ideal reactor: the perfectly stirred reactor [often called a "continuously stirred tank reactor" (CSTR)] or the plug-flow reactor. The horizontal biofilter with back-mixing was able to accommodate a shorter residence time without the usual requirement of greater biofilter surface area for increased biofiltration efficiency. Experimental results indicated that the first bed of the modified biofilter behaved like two CSTRs in series, while the second bed may be represented by two or three CSTRs in series. Because of the flow baffles used in the horizontal biofilter system, its performance was more similar to completely mixed systems, and hence, it could not be modeled as a plug-flow reactor. For the standard biofilter, the number of CSTRs was found to be between 2 and 9 depending on the airflow rate. In terms of NH3 removal efficiency and elimination capacity, the standard biofilter was not as good as the modified system; moreover, the second bed of the modified biofilter exhibited greater removal efficiency than the first bed. The elimination rate increased as biofilter load increased. An opposite trend was exhibited with respect to removal efficiency.

Synthesis of layered double hydroxide nanosheets by coprecipitation using a T-type microchannel reactor

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

Pang, Xiujiang; Sun, Meiyu; Ma, Xiuming

The synthesis of Mg{sub 2}Al–NO{sub 3} layered double hydroxide (LDH) nanosheets by coprecipitation using a T-type microchannel reactor is reported. Aqueous LDH nanosheet dispersions were obtained. The LDH nanosheets were characterized by X-ray diffraction, transmission electron microscopy, atomic force microscopy and particle size analysis, and the transmittance and viscosity of LDH nanosheet dispersions were examined. The two-dimensional LDH nanosheets consisted of 1–2 brucite-like layers and were stable for ca. 16 h at room temperature. In addition, the co-assembly between LDH nanosheets and dodecyl sulfate (DS) anions was carried out, and a DS intercalated LDH nanohybrid was obtained. To the bestmore » of our knowledge, this is the first report of LDH nanosheets being directly prepared in bulk aqueous solution. This simple, cheap method can provide naked LDH nanosheets in high quantities, which can be used as building blocks for functional materials. - Graphical abstract: Layered double hydroxide (LDH) nanosheets were synthesized by coprecipitation using a T-type microchannel reactor, and could be used as basic building blocks for LDH-based functional materials. Display Omitted - Highlights: • LDH nanosheets were synthesized by coprecipitation using a T-type microchannel reactor. • Naked LDH nanosheets were dispersed in aqueous media. • LDH nanosheets can be used as building blocks for functional materials.« less

NEUTRON REACTOR HAVING A Xe$sup 135$ SHIELD

DOEpatents

Stanton, H.E.

1957-10-29

Shielding for reactors of the type in which the fuel is a chain reacting liquid composition comprised essentially of a slurry of fissionable and fertile material suspended in a liquid moderator is discussed. The neutron reflector comprises a tank containing heavy water surrounding the reactor, a shield tank surrounding the reflector, a gamma ray shield surrounding said shield tank, and a means for conveying gaseous fission products, particularly Xe/sup 135/, from the reactor chamber to the shield tank, thereby serving as a neutron shield by capturing the thermalized neutrons that leak outwardly from the shield tank.

CONTROL FOR NEUTRONIC REACTOR

DOEpatents

Lichtenberger, H.V.; Cameron, R.A.

1959-03-31

S>A control rod operating device in a nuclear reactor of the type in which the control rod is gradually withdrawn from the reactor to a position desired during stable operation is described. The apparatus is comprised essentially of a stop member movable in the direction of withdrawal of the control rod, a follower on the control rod engageable with the stop and means urging the follower against the stop in the direction of withdrawal. A means responsive to disengagement of the follower from the stop is provided for actuating the control rod to return to the reactor shut-down position.

SP-100 Program: space reactor system and subsystem investigations

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

Harty, R.B.

1983-09-30

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

SP-100 program: Space reactor system and subsystem investigations

NASA Astrophysics Data System (ADS)

Harty, R. B.

1983-09-01

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

On the factors influencing the performance of solar reactors for water disinfection with photosensitized singlet oxygen.

PubMed

Manjón, Francisco; Villén, Laura; García-Fresnadillo, David; Orellana, Guillermo

2008-01-01

Two solar reactors based on compound parabolic collectors (CPCs) were optimized for water disinfection by photosensitized singlet oxygen (1O2) production in the heterogeneous phase. Sensitizing materials containing Ru(II) complexes immobilized on porous silicone were produced, photochemically characterized, and successfully tested for the inactivation of up to 10(4) CFU mL(-1) of waterborne Escherichia coli (gram-negative) or Enterococcus faecalis (gram-positive) bacteria. The main factors determining the performance of the solar reactors are the type of photosensitizing material, the sensitizer loading, the CPC collector geometry (fin- vs coaxial-type), the fluid rheology, and the balance between concurrent photothermal--photolytic and 1O2 effects on the microorganisms' inactivation. In this way, at the 40 degrees N latitude of Spain, water can be disinfected on a sunny day (0.6-0.8 MJ m(-2) L(-1) accumulated solar radiation dose in the 360-700 nm range, typically 5-6 h of sunlight) with a fin-type reactor containing 0.6 m2 of photosensitizing material saturated with tris(4,7-diphenyl-1,10-phenanthroline)ruthenium(II) (ca. 2.0 g m(-2)). The optimum rheological conditions require laminar-to-transitional water flow in both prototypes. The fin-type system showed better inactivation efficiency than the coaxial reactor due to a more important photolytic contribution. The durability of the sensitizing materials was tested and the operational lifetime of the photocatalyst is at least three months without any reduction in the bacteria inactivation efficiency. Solar water disinfection with 1O2-generating films is demonstrated to be an effective technique for use in isolated regions of developing countries with high yearly average sunshine.

Proposal of a neutron transmutation doping facility for n-type spherical silicon solar cell at high-temperature engineering test reactor.

PubMed

Ho, Hai Quan; Honda, Yuki; Motoyama, Mizuki; Hamamoto, Shimpei; Ishii, Toshiaki; Ishitsuka, Etsuo

2018-05-01

The p-type spherical silicon solar cell is a candidate for future solar energy with low fabrication cost, however, its conversion efficiency is only about 10%. The conversion efficiency of a silicon solar cell can be increased by using n-type silicon semiconductor as a substrate. This study proposed a new method of neutron transmutation doping silicon (NTD-Si) for producing the n-type spherical solar cell, in which the Si-particles are irradiated directly instead of the cylinder Si-ingot as in the conventional NTD-Si. By using a 'screw', an identical resistivity could be achieved for the Si-particles without a complicated procedure as in the NTD with Si-ingot. Also, the reactivity and neutron flux swing could be kept to a minimum because of the continuous irradiation of the Si-particles. A high temperature engineering test reactor (HTTR), which is located in Japan, was used as a reference reactor in this study. Neutronic calculations showed that the HTTR has a capability to produce about 40t/EFPY of 10Ωcm resistivity Si-particles for fabrication of the n-type spherical solar cell. Copyright © 2018 Elsevier Ltd. All rights reserved.

10 CFR 52.131 - Scope of subpart.

Code of Federal Regulations, 2010 CFR

2010-01-01

... REGULATORY COMMISSION (CONTINUED) LICENSES, CERTIFICATIONS, AND APPROVALS FOR NUCLEAR POWER PLANTS Standard... review, and referral to the Advisory Committee on Reactor Safeguards of standard designs for a nuclear power reactor of the type described in § 50.22 of this chapter or major portions thereof. ...

Count-doubling time safety circuit

DOEpatents

Rusch, Gordon K.; Keefe, Donald J.; McDowell, William P.

1981-01-01

There is provided a nuclear reactor count-factor-increase time monitoring circuit which includes a pulse-type neutron detector, and means for counting the number of detected pulses during specific time periods. Counts are compared and the comparison is utilized to develop a reactor scram signal, if necessary.

EFFECTS OF REACTOR CONDITIONS ON ELECTROCHEMICAL DECHLORINATION OF TRICHLOROETHYLENE USING GRANULAR-GRAPHITE ELECTRODE

EPA Science Inventory

Trichloroethylene (TCE) was electrochemically dechlorinated in aqueous environments using granular graphite cathode in a mixed reactor. Effects of pH, current, electrolyte type, and flow rate on TCE dechlorination rate were evaluated. TCE dechlorination rate constant and gas pr...

EFFECTS OF REACTOR CONDITIONS ON ELECTROCHEMICAL DECHLORINATION OF TRICHLOROETHYLENE USING GRANULAR-GRAPHITE ELECTRODE.

EPA Science Inventory

Trichloroethylene (TCE) was electrochemically dechlorinated in aqueous environments using granular graphite cathode in a mixed reactor. Effects of pH, current, electrolyte type, and flow rate on TCE dechlorination rate were evaluated. TCE dechlorination rate constant and gas pr...

REACTOR CONTROL

DOEpatents

Ruano, W.J.

1957-12-10

This patent relates to nuclear reactors of the type which utilize elongited rod type fuel elements immersed in a liquid moderator and shows a design whereby control of the chain reaction is obtained by varying the amount of moderator or reflector material. A central tank for containing liquid moderator and fuel elements immersed therein is disposed within a surrounding outer tank providing an annular space between the two tanks. This annular space is filled with liquid moderator which functions as a reflector to reflect neutrons back into the central reactor tank to increase the reproduction ratio. Means are provided for circulating and cooling the moderator material in both tanks and additional means are provided for controlling separately the volume of moderator in each tank, which latter means may be operated automatically by a neutron density monitoring device. The patent also shows an arrangement for controlling the chain reaction by injecting and varying an amount of poisoning material in the moderator used in the reflector portion of the reactor.

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

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

Not Available

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

Down-selection of candidate alloys for further testing of advanced replacement materials for LWR core internals

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

Was, Gary; Leonard, Keith J.; Tan, Lizhen

Life extension of the existing nuclear reactors imposes irradiation of high fluences to structural materials, resulting in significant challenges to the traditional reactor materials such as type 304 and 316 stainless steels. Advanced alloys with superior radiation resistance will increase safety margins, design flexibility, and economics for not only the life extension of the existing fleet but also new builds with advanced reactor designs. The Electric Power Research Institute (EPRI) teamed up with Department of Energy (DOE) Light Water Reactor Sustainability Program to initiate the Advanced Radiation Resistant Materials (ARRM) program, aiming to identify and develop advanced alloys with superiormore » degradation resistance in light water reactor (LWR)-relevant environments by 2024.« less

NEUTRONIC REACTOR CONSTRUCTION

DOEpatents

Vernon, H.C.; Goett, J.J.

1958-09-01

A cover device is described for the fuel element receiving tube of a neutronic reactor of the heterogeneous, water cooled type wherein said tubes are arranged in a moderator with their longitudinal axes vertical. The cover is provided with means to support a rod-type fuel element from the bottom thereof and means to lock the cover in place, the latter being adapted for remote operation. This cover device is easily removable and seals the opening in the upper end of the fuel tube against leakage of coolant.

Summary of Thermocouple Performance During Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor and Out-of-Pile Thermocouple Testing in Support of Such Experiments

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

A. J. Palmer; DC Haggard; J. W. Herter

High temperature gas reactor experiments create unique challenges for thermocouple based temperature measurements. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time dependent change in composition and, as a consequence, a time dependent drift of the thermocouple signal. This drift is particularly severe for high temperature platinum-rhodium thermocouples (Types S, R, and B); and tungsten-rhenium thermocouples (Types C and W). For lower temperature applications, previous experiences with type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly type Nmore » thermocouples are expected to be only slightly affected by neutron fluxes. Currently the use of these Nickel based thermocouples is limited when the temperature exceeds 1000°C due to drift related to phenomena other than nuclear irradiation. High rates of open-circuit failure are also typical. Over the past ten years, three long-term Advanced Gas Reactor (AGR) experiments have been conducted with measured temperatures ranging from 700oC – 1200oC. A variety of standard Type N and specialty thermocouple designs have been used in these experiments with mixed results. A brief summary of thermocouple performance in these experiments is provided. Most recently, out of pile testing has been conducted on a variety of Type N thermocouple designs at the following (nominal) temperatures and durations: 1150oC and 1200oC for 2000 hours at each temperature, followed by 200 hours at 1250oC, and 200 hours at 1300oC. The standard Type N design utilizes high purity crushed MgO insulation and an Inconel 600 sheath. Several variations on the standard Type N design were tested, including Haynes 214 alloy sheath, spinel (MgAl2O4) insulation instead of MgO, a customized sheath developed at the University of Cambridge, and finally a loose assembly thermocouple with hard fired alumina insulation and molybdenum sheath. The most current version of the High Temperature Irradiation Resistant Thermocouple (HTIR-TC) based on molybdenum/niobium alloys, and developed at Idaho National Laboratory, was also tested.« less

Summary of thermocouple performance during advanced gas reactor fuel irradiation experiments in the advanced test reactor and out-of-pile thermocouple testing in support of such experiments

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

Palmer, A. J.; Haggard, DC; Herter, J. W.

High temperature gas reactor experiments create unique challenges for thermocouple-based temperature measurements. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time-dependent change in composition and, as a consequence, a time-dependent drift of the thermocouple signal. This drift is particularly severe for high temperature platinum-rhodium thermocouples (Types S, R, and B) and tungsten-rhenium thermocouples (Type C). For lower temperature applications, previous experiences with Type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly, Type N thermocouples are expected to bemore » only slightly affected by neutron fluence. Currently, the use of these nickel-based thermocouples is limited when the temperature exceeds 1000 deg. C due to drift related to phenomena other than nuclear irradiation. High rates of open-circuit failure are also typical. Over the past 10 years, three long-term Advanced Gas Reactor experiments have been conducted with measured temperatures ranging from 700 deg. C - 1200 deg. C. A variety of standard Type N and specialty thermocouple designs have been used in these experiments with mixed results. A brief summary of thermocouple performance in these experiments is provided. Most recently, out-of-pile testing has been conducted on a variety of Type N thermocouple designs at the following (nominal) temperatures and durations: 1150 deg. C and 1200 deg. C for 2,000 hours at each temperature, followed by 200 hours at 1250 deg. C and 200 hours at 1300 deg. C. The standard Type N design utilizes high purity, crushed MgO insulation and an Inconel 600 sheath. Several variations on the standard Type N design were tested, including a Haynes 214 alloy sheath, spinel (MgAl{sub 2}O{sub 4}) insulation instead of MgO, a customized sheath developed at the University of Cambridge, and finally a loose assembly thermocouple with hard-fired alumina insulation and a molybdenum sheath. The most current version of the High Temperature Irradiation Resistant Thermocouple, based on molybdenum/niobium alloys and developed at Idaho National Laboratory, was also tested. (authors)« less

Small Modular Reactors: The Army’s Secure Source of Energy?

DTIC Science & Technology

2012-03-21

significant advantages of SMRs is the minimal amount of carbon dioxide (greenhouse gases) that is released in conjunction with the lifecycle operations...moderator in these reactors as well as the cooling agent and the means by which heat is removed to produce steam for turning the turbines of the...separate water system to generate steam to turn a turbine which then produces electricity. In the second type of light water reactors, the boiling water

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.

Axially staggered seed-blanket reactor-fuel-module construction. [LWBR

DOEpatents

Cowell, G.K.; DiGuiseppe, C.P.

1982-10-28

A heterogeneous nuclear reactor of the seed-blanket type is provided wherein the fissile (seed) and fertile (blanket) nuclear fuels are segregated axially within each fuel element such that fissile and fertile regions occur in an alternating pattern along the length of the fuel element. Further, different axial stacking patterns are used for the fuel elements of at least two module types such that when modules of different types are positioned adjacent to one another, the fertile regions of the modules are offset or staggered. Thus, when a module of one type is surrounded by modules of the second type the fertile regions thereof will be surrounded on all sides by fissile material. This provides enhanced neutron communication both radially and axially, thereby resulting in greater power oscillation stability than other axial arrangements.

Control of reactor coolant flow path during reactor decay heat removal

DOEpatents

Hunsbedt, Anstein N.

1988-01-01

An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

Graphite for the nuclear industry

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

Burchell, T.D.; Fuller, E.L.; Romanoski, G.R.

Graphite finds applications in both fission and fusion reactors. Fission reactors harness the energy liberated when heavy elements, such as uranium or plutonium, fragment or fission''. Reactors of this type have existed for nearly 50 years. The first nuclear fission reactor, Chicago Pile No. 1, was constructed of graphite under a football stand at Stagg Field, University of Chicago. Fusion energy devices will produce power by utilizing the energy produced when isotopes of the element hydrogen are fused together to form helium, the same reaction that powers our sun. The role of graphite is very different in these two reactormore » systems. Here we summarize the function of the graphite in fission and fusion reactors, detailing the reasons for their selection and discussing some of the challenges associated with their application in nuclear fission and fusion reactors. 10 refs., 15 figs., 1 tab.« less

SELF-REACTIVATING NEUTRON SOURCE FOR A NEUTRONIC REACTOR

DOEpatents

Newson, H.W.

1959-02-01

Reactors of the type employing beryllium in a reflector region around the active portion and to a neutron source for use therewith are discussed. The neutron source is comprised or a quantity of antimony permanently incorporated in, and as an integral part of, the reactor in or near the beryllium reflector region. During operation of the reactor the natural occurring antimony isotope of atomic weight 123 absorbs neutrons and is thereby transformed to the antimony isotope of atomic weight 124, which is radioactive and emits gamma rays. The gamma rays react with the beryllium to produce neutrons. The beryllium and antimony thus cooperate to produce a built in neutron source which is automatically reactivated by the operation of the reactor itself and which is of sufficient strength to maintain the slow neutron flux at a sufficiently high level to be reliably measured during periods when the reactor is shut down.

Hanging core support system for a nuclear reactor. [LMFBR

DOEpatents

Burelbach, J.P.; Kann, W.J.; Pan, Y.C.; Saiveau, J.G.; Seidensticker, R.W.

1984-04-26

For holding the reactor core in the confining reactor vessel, a support is disclosed that is structurally independent of the vessel, that is dimensionally accurate and stable, and that comprises tandem tension linkages that act redundantly of one another to maintain stabilized core support even in the unlikely event of the complete failure of one of the linkages. The core support has a mounting platform for the reactor core, and unitary structure including a flange overlying the top edge of the reactor vessels, and a skirt and box beams between the flange and platform for establishing one of the linkages. A plurality of tension rods connect between the deck closing the reactor vessel and the platform for establishing the redundant linkage. Loaded Belleville springs flexibly hold the tension rods at the deck and separable bayonet-type connections hold the tension rods at the platform.

Development of a Model and Computer Code to Describe Solar Grade Silicon Production Processes

NASA Technical Reports Server (NTRS)

Srivastava, R.; Gould, R. K.

1979-01-01

Mathematical models and computer codes based on these models, which allow prediction of the product distribution in chemical reactors for converting gaseous silicon compounds to condensed-phase silicon were developed. The following tasks were accomplished: (1) formulation of a model for silicon vapor separation/collection from the developing turbulent flow stream within reactors of the Westinghouse (2) modification of an available general parabolic code to achieve solutions to the governing partial differential equations (boundary layer type) which describe migration of the vapor to the reactor walls, (3) a parametric study using the boundary layer code to optimize the performance characteristics of the Westinghouse reactor, (4) calculations relating to the collection efficiency of the new AeroChem reactor, and (5) final testing of the modified LAPP code for use as a method of predicting Si(1) droplet sizes in these reactors.

Application of point kinetics equations to the design of a reactivity meter

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

Binney, S.E.; Bakir, A.J.M.

1988-01-01

The time-dependent reactivity of a nuclear reactor is obviously one of the most important reactor parameters that describes the state of the reactor. Although several different types of techniques exist to measure reactivity, only the kinetic method is described here. The paper illustrates the measured reactor power and calculated reactivity for a 70 cents step change in reactivity. These data were taken at 1-s time intervals. It is seen that the reactivity, initially at zero, rises rapidly to a predetermined value (determined by the reactivity change induced in the system) and then returns to zero as the reactor is reestablishedmore » in a critical situation by insertion of another control rod. It is concluded that the method of Tuttle has been adapted to produce a reliable, on-line calculation of reactivity from a time-dependent reactor power signal.« less

A systematic reactor design approach for the synthesis of active pharmaceutical ingredients.

PubMed

Emenike, Victor N; Schenkendorf, René; Krewer, Ulrike

2018-05-01

Today's highly competitive pharmaceutical industry is in dire need of an accelerated transition from the drug development phase to the drug production phase. At the heart of this transition are chemical reactors that facilitate the synthesis of active pharmaceutical ingredients (APIs) and whose design can affect subsequent processing steps. Inspired by this challenge, we present a model-based approach for systematic reactor design. The proposed concept is based on the elementary process functions (EPF) methodology to select an optimal reactor configuration from existing state-of-the-art reactor types or can possibly lead to the design of novel reactors. As a conceptual study, this work summarizes the essential steps in adapting the EPF approach to optimal reactor design problems in the field of API syntheses. Practically, the nucleophilic aromatic substitution of 2,4-difluoronitrobenzene was analyzed as a case study of pharmaceutical relevance. Here, a small-scale tubular coil reactor with controlled heating was identified as the optimal set-up reducing the residence time by 33% in comparison to literature values. Copyright © 2017 Elsevier B.V. All rights reserved.

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

Central waste processing system

NASA Technical Reports Server (NTRS)

Kester, F. L.

1973-01-01

A new concept for processing spacecraft type wastes has been evaluated. The feasibility of reacting various waste materials with steam at temperatures of 538 - 760 C in both a continuous and batch reactor with residence times from 3 to 60 seconds has been established. Essentially complete gasification is achieved. Product gases are primarily hydrogen, carbon dioxide, methane, and carbon monoxide. Water soluble synthetic wastes are readily processed in a continuous tubular reactor at concentrations up to 20 weight percent. The batch reactor is able to process wet and dry wastes at steam to waste weight ratios from 2 to 20. Feces, urine, and synthetic wastes have been successfully processed in the batch reactor.

COOLED NEUTRONIC REACTOR

DOEpatents

Binner, C.R.; Wilkie, C.B.

1958-03-18

This patent relates to a design for a reactor of the type in which a fluid coolant is flowed through the active portion of the reactor. This design provides for the cooling of the shielding material as well as the reactor core by the same fluid coolant. The core structure is a solid moderator having coolant channels in which are disposed the fuel elements in rod or slug form. The coolant fluid enters the chamber in the shield, in which the core is located, passes over the inner surface of said chamber, enters the core structure at the center, passes through the coolant channels over the fuel elements and out through exhaust ducts.

METHOD OF SUSTAINING A NEUTRONIC CHAIN REACTING SYSTEM

DOEpatents

Fermi, E.; Leverett, M.C.

1957-11-12

This patent relates to neutronic reactors and a method of sustainlng a chain reaction. The reactor shown in the patent for carrying out the method is the gas-cooled type comprised of a solid moderator having a plurality of passages therethrough for receiving bodies of fissionable material. In carrying out the method, the reactor is loaded by inserting in the passages fuel elements and moderator material in a proportion to sustain a chain reaction As the reproduction ratio decreases below the desired fiiaire due to impurities formed during operation of the reactor, the moderator material is gradually replaced with additional fuel material to maintain the reproduction ratio above unity.

Turbulence coefficients and stability studies for the coaxial flow or dissimiliar fluids. [gaseous core nuclear reactors

NASA Technical Reports Server (NTRS)

Weinstein, H.; Lavan, Z.

1975-01-01

Analytical investigations of fluid dynamics problems of relevance to the gaseous core nuclear reactor program are presented. The vortex type flow which appears in the nuclear light bulb concept is analyzed along with the fluid flow in the fuel inlet region for the coaxial flow gaseous core nuclear reactor concept. The development of numerical methods for the solution of the Navier-Stokes equations for appropriate geometries is extended to the case of rotating flows and almost completes the gas core program requirements in this area. The investigations demonstrate that the conceptual design of the coaxial flow reactor needs further development.

Achieving ethanol-type fermentation for hydrogen production in a granular sludge system by aeration.

PubMed

Zhang, Song; Liu, Min; Chen, Ying; Pan, Yu-Ting

2017-01-01

To investigate the effects of aeration on hydrogen-producing granular system, experiments were performed in two laboratory-scale anaerobic internal circulation hydrogen production (AICHP) reactors. The preliminary experiment of Reactor 1 showed that direct aeration was beneficial to enhancing hydrogen production. After the direct aeration was implied in Reactor 2, hydrogen production rate (HPR) and hydrogen content were increased by 100% and 60%, respectively. In addition, mixed-acid fermentation was transformed into typical ethanol-type fermentation (ETF). Illumina MiSeq sequencing shows that the direct aeration did not change the species of hydrogen-producing bacteria but altered their abundance. Hydrogen-producing bacteria and ethanol-type fermentative bacteria were increased by 24.5% and 146.3%, respectively. Ethanoligenens sp. sharply increased by 162.2% and turned into predominant bacteria in the system. These findings indicated that appropriate direct aeration might be a novel and promising way to obtain ETF and enhance hydrogen production in practical use. Copyright © 2016 Elsevier Ltd. All rights reserved.

48 CFR 2009.570-3 - Criteria for recognizing contractor organizational conflicts of interest.

Code of Federal Regulations, 2011 CFR

2011-10-01

... reactor component that is unique to one type of advanced reactor. As is the case with other technically... contractor prepares plans for specific approaches or methodologies that are to be incorporated into competitive procurements using the approaches or methodologies. (iii) Where the offeror or contractor is...

48 CFR 2009.570-3 - Criteria for recognizing contractor organizational conflicts of interest.

Code of Federal Regulations, 2010 CFR

2010-10-01

... reactor component that is unique to one type of advanced reactor. As is the case with other technically... contractor prepares plans for specific approaches or methodologies that are to be incorporated into competitive procurements using the approaches or methodologies. (iii) Where the offeror or contractor is...

48 CFR 2009.570-3 - Criteria for recognizing contractor organizational conflicts of interest.

Code of Federal Regulations, 2013 CFR

2013-10-01

... reactor component that is unique to one type of advanced reactor. As is the case with other technically... contractor prepares plans for specific approaches or methodologies that are to be incorporated into competitive procurements using the approaches or methodologies. (iii) Where the offeror or contractor is...

48 CFR 2009.570-3 - Criteria for recognizing contractor organizational conflicts of interest.

Code of Federal Regulations, 2014 CFR

2014-10-01

... reactor component that is unique to one type of advanced reactor. As is the case with other technically... contractor prepares plans for specific approaches or methodologies that are to be incorporated into competitive procurements using the approaches or methodologies. (iii) Where the offeror or contractor is...

48 CFR 2009.570-3 - Criteria for recognizing contractor organizational conflicts of interest.

Code of Federal Regulations, 2012 CFR

2012-10-01

... reactor component that is unique to one type of advanced reactor. As is the case with other technically... contractor prepares plans for specific approaches or methodologies that are to be incorporated into competitive procurements using the approaches or methodologies. (iii) Where the offeror or contractor is...

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.

Modeling and Comparison of Options for the Disposal of Excess Weapons Plutonium in Russia

DTIC Science & Technology

2002-04-01

fuel LWR cooling time LWR Pu load rate LWR net destruction frac ~ LWR reactors op life mox core frac Excess Separated Pu HTGR Cycle Pu in Waste LWR MOX...reflecting the cycle used in this type of reactor. For the HTGR , the entire core consists of plutonium fuel , therefore a core fraction is not specified...cooling time Time spent fuel unloaded from HTGR reactor must cool before permanently stored 3 years Mox core fraction Fraction of

FUEL ASSEMBLY FOR A NEUTRONIC REACTOR

DOEpatents

Wigner, E.P.

1958-04-29

A fuel assembly for a nuclear reactor of the type wherein liquid coolant is circulated through the core of the reactor in contact with the external surface of the fuel elements is described. In this design a plurality of parallel plates containing fissionable material are spaced about one-tenth of an inch apart and are supported between a pair of spaced parallel side members generally perpendicular to the plates. The plates all have a small continuous and equal curvature in the same direction between the side members.

Development of High Quality 4H-SiC Thick Epitaxy for Reliable High Power Electronics Using Halogenated Precursors

DTIC Science & Technology

2016-08-02

epitaxy platform, it is essential that malignant defects, such as in-grown stacking faults (IGSFs) and basal plane dislocations (BPDs), be...crystal quality. (5) Even though the inlet C/Si ratio is kept fixed , the C/Si ratio at the growth surface varies depending on the different gas...morphology, and quality (generation of additional defects). Two CVD reactor types, a chimney reactor and an inverted chimney reactor, are assembled; the

A Programmable Liquid Collimator for Both Coded Aperture Adaptive Imaging and Multiplexed Compton Scatter Tomography

DTIC Science & Technology

2012-03-01

environments where a source is either weak or shielded. A vehicle of this type could survey large areas after a nuclear attack or a nuclear reactor accident...to prevent its detection by γ-rays. The best application for unmanned vehicles is the detection of radioactive material after a nuclear reactor ...accident or a nuclear weapon detonation [70]. Whether by a nuclear detonation or a nuclear reactor accident, highly radioactive substances could be dis

Pyrolysis of aseptic packages (tetrapak) in a laboratory screw type reactor and secondary thermal/catalytic tar decomposition.

PubMed

Haydary, J; Susa, D; Dudáš, J

2013-05-01

Pyrolysis of aseptic packages (tetrapak cartons) in a laboratory apparatus using a flow screw type reactor and a secondary catalytic reactor for tar cracking was studied. The pyrolysis experiments were realized at temperatures ranging from 650 °C to 850 °C aimed at maximizing of the amount of the gas product and reducing its tar content. Distribution of tetrapak into the product yields at different conditions was obtained. The presence of H2, CO, CH4, CO2 and light hydrocarbons, HCx, in the gas product was observed. The Aluminum foil was easily separated from the solid product. The rest part of char was characterized by proximate and elemental analysis and calorimetric measurements. The total organic carbon in the tar product was estimated by elemental analysis of tars. Two types of catalysts (dolomite and red clay marked AFRC) were used for catalytic thermal tar decomposition. Three series of experiments (without catalyst in a secondary cracking reactor, with dolomite and with AFRC) at temperatures of 650, 700, 750, 800 and 850 °C were carried out. Both types of catalysts have significantly affected the content of tars and other components in pyrolytic gases. The effect of catalyst on the tetrapack distribution into the product yield on the composition of gas and on the total organic carbon in the tar product is presented in this work. Copyright © 2013 Elsevier Ltd. All rights reserved.

Structural materials issues for the next generation fission reactors

NASA Astrophysics Data System (ADS)

Chant, I.; Murty, K. L.

2010-09-01

Generation-IV reactor design concepts envisioned thus far cater to a common goal of providing safer, longer lasting, proliferation-resistant, and economically viable nuclear power plants. The foremost consideration in the successful development and deployment of Gen-W reactor systems is the performance and reliability issues involving structural materials for both in-core and out-of-core applications. The structural materials need to endure much higher temperatures, higher neutron doses, and extremely corrosive environments, which are beyond the experience of the current nuclear power plants. Materials under active consideration for use in different reactor components include various ferritic/martensitic steels, austenitic stainless steels, nickel-base superalloys, ceramics, composites, etc. This article addresses the material requirements for these advanced fission reactor types, specifically addressing structural materials issues depending on the specific application areas.

Component and System Sensitivity Considerations for Design of a Lunar ISRU Oxygen Production Plant

NASA Technical Reports Server (NTRS)

Linne, Diane L.; Gokoglu, Suleyman; Hegde, Uday G.; Balasubramaniam, Ramaswamy; Santiago-Maldonado, Edgardo

2009-01-01

Component and system sensitivities of some design parameters of ISRU system components are analyzed. The differences between terrestrial and lunar excavation are discussed, and a qualitative comparison of large and small excavators is started. The effect of excavator size on the size of the ISRU plant's regolith hoppers is presented. Optimum operating conditions of both hydrogen and carbothermal reduction reactors are explored using recently developed analytical models. Design parameters such as batch size, conversion fraction, and maximum particle size are considered for a hydrogen reduction reactor while batch size, conversion fraction, number of melt zones, and methane flow rate are considered for a carbothermal reduction reactor. For both reactor types the effect of reactor operation on system energy and regolith delivery requirements is presented.

Development of Cross Section Library and Application Programming Interface (API)

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

Lee, C. H.; Marin-Lafleche, A.; Smith, M. A.

2014-04-09

The goal of NEAMS neutronics is to develop a high-fidelity deterministic neutron transport code termed PROTEUS for use on all reactor types of interest, but focused primarily on sodium-cooled fast reactors. While PROTEUS-SN has demonstrated good accuracy for homogeneous fast reactor problems and partially heterogeneous fast reactor problems, the simulation results were not satisfactory when applied on fully heterogeneous thermal problems like the Advanced Test Reactor (ATR). This is mainly attributed to the quality of cross section data for heterogeneous geometries since the conventional cross section generation approach does not work accurately for such irregular and complex geometries. Therefore, onemore » of the NEAMS neutronics tasks since FY12 has been the development of a procedure to generate appropriate cross sections for a heterogeneous geometry core.« less

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

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

Kang, Jung Kil, E-mail: jkkang@email.kings.ac.kr; Hah, Chang Joo, E-mail: changhah@kings.ac.kr; Cho, Sung Ju, E-mail: sungju@knfc.co.kr

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

Biodiesel production process intensification using a rotor-stator type generator of hydrodynamic cavitation.

PubMed

Crudo, Daniele; Bosco, Valentina; Cavaglià, Giuliano; Grillo, Giorgio; Mantegna, Stefano; Cravotto, Giancarlo

2016-11-01

Triglyceride transesterification for biodiesel production is a model reaction which is used to compare the conversion efficiency, yield, reaction time, energy consumption, scalability and cost estimation of different reactor technology and energy source. This work describes an efficient, fast and cost-effective procedure for biodiesel preparation using a rotating generator of hydrodynamic cavitation (HC). The base-catalyzed transesterification (methanol/sodium hydroxide) has been carried out using refined and bleached palm oil and waste vegetable cooking oil. The novel HC unit is a continuous rotor-stator type reactor in which reagents are directly fed into the controlled cavitation chamber. The high-speed rotation of the reactor creates micron-sized droplets of the immiscible reacting mixture leading to outstanding mass and heat transfer and enhancing the kinetics of the transesterification reaction which completes much more quickly than traditional methods. All the biodiesel samples obtained respect the ASTM standard and present fatty acid methyl ester contents of >99% m/m in both feedstocks. The electrical energy consumption of the HC reactor is 0.030kWh per L of produced crude biodiesel, making this innovative technology really quite competitive. The reactor can be easily scaled-up, from producing a few hundred to thousands of liters of biodiesel per hour while avoiding the risk of orifices clogging with oil impurities, which may occur in conventional HC reactors. Furthermore it requires minimal installation space due to its compact design, which enhances overall security. Copyright © 2016 Elsevier B.V. All rights reserved.

Helium Leak Detection of Vessels in Fuel Transfer Cell (FTC) of Prototype Fast Breeder Reactor (PFBR)

NASA Astrophysics Data System (ADS)

Dutta, N. G.

2012-11-01

Bharatiya Nabhikiya Vidyut Nigam (BHAVINI) is engaged in construction of 500MW Prototype Fast Breeder Reactor (PFBR) at Kalpak am, Chennai. In this very important and prestigious national programme Special Product Division (SPD) of M/s Kay Bouvet Engg.pvt. ltd. (M/s KBEPL) Satara is contributing in a major way by supplying many important sub-assemblies like- Under Water trolley (UWT), Airlocks (PAL, EAL) Container and Storage Rack (CSR) Vessels in Fuel Transfer Cell (FTC) etc for PFBR. SPD of KBEPL caters to the requirements of Government departments like - Department of Atomic Energy (DAE), BARC, Defense, and Government undertakings like NPCIL, BHAVINI, BHEL etc. and other precision Heavy Engg. Industries. SPD is equipped with large size Horizontal Boring Machines, Vertical Boring Machines, Planno milling, Vertical Turret Lathe (VTL) & Radial drilling Machine, different types of welding machines etc. PFBR is 500 MWE sodium cooled pool type reactor in which energy is produced by fissions of mixed oxides of Uranium and Plutonium pellets by fast neutrons and it also breeds uranium by conversion of thorium, put along with fuel rod in the reactor. In the long run, the breeder reactor produces more fuel then it consumes. India has taken the lead to go ahead with Fast Breeder Reactor Programme to produce electricity primarily because India has large reserve of Thorium. To use Thorium as further fuel in future, thorium has to be converted in Uranium by PFBR Technology.

View of Pakistan Atomic Energy Commission towards SMPR's in the light of KANUPP performance

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

Huseini, S.D.

1985-01-01

The developing countries in general do not have grid capacities adequate enough to incorporate standard size, economic but rather large nuclear power plants for maximum advantage. Therefore, small and medium size reactors (SMPR) have been and still are, of particular interest to the developing countries in spite of certain known problems with these reactors. Pakistan Atomic Energy Commission (PAEC) has been operating a CANDU type of a small PHWR plant since 1971 when it was connected to the local Karachi grid. This paper describes PAEC's view in the light of KANUPP performance with respect to such factors associated with SMPR'smore » as selection of suitable reactor size and type, its operation in a grid of small capacity, flexibility of operation and its role as a reliable source of electrical power.« less

NUCLEAR REACTOR

DOEpatents

Treshow, M.

1958-08-19

A neuclear reactor is described of the heterogeneous type and employing replaceable tubular fuel elements and heavy water as a coolant and moderator. A pluraltty of fuel tubesa having their axes parallel, extend through a tank type pressure vessel which contatns the liquid moderator. The fuel elements are disposed within the fuel tubes in the reaetive portion of the pressure vessel during normal operation and the fuel tubes have removable plug members at each end to permit charging and discharging of the fuel elements. The fuel elements are cylindrical strands of jacketed fissionable material having helical exterior ribs. A bundle of fuel elements are held within each fuel tube with their longitudinal axes parallel, the ribs serving to space them apart along their lengths. Coolant liquid is circulated through the fuel tubes between the spaced fuel elements. Suitable control rod and monitoring means are provided for controlling the reactor.

POWER GENERATING NEUTRONIC REACTOR SYSTEM

DOEpatents

Vernon, H.C.

1958-03-01

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

Taming The Next Set of Strategic Weapons Threats

DTIC Science & Technology

2006-06-01

Reactors Victor Gilinsky . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 61 6. Coping with Biological Threats after...Regime (MTCR) is not yet optimized to cope with these challenges. Finally, nuclear technologies have become much more difficult to control. New...resistance of the most popular type of power reactor concludes that the current international nuclear safeguards system needs to be modified to cope

Biocatalytic methanation of hydrogen and carbon dioxide in an anaerobic three-phase system.

PubMed

Burkhardt, M; Koschack, T; Busch, G

2015-02-01

A new type of anaerobic trickle-bed reactor was used for biocatalytic methanation of hydrogen and carbon dioxide under mesophilic temperatures and ambient pressure in a continuous process. The conversion of gaseous substrates through immobilized hydrogenotrophic methanogenic archaea in a biofilm is a unique feature of this type of reactor. Due to the formation of a three-phase system on the carrier surface and operation as a plug flow reactor without gas recirculation, a complete reaction could be observed. With a methane concentration higher than c(CH4) = 98%, the product gas exhibits a very high quality. A specific methane production of P(CH4) = 1.49 Nm(3)/(m(3)(SV) d) was achieved at a hydraulic loading rate of LR(H2) = 6.0 Nm(3)/(m(3)(SV) d). The relation between trickle flow through the reactor and productivity could be shown. An application for methane enrichment in combination with biogas facilities as a source of carbon dioxide has also been positively proven. Copyright © 2014 Elsevier Ltd. All rights reserved.

Continuous production of pectinase by immobilized yeast cells on spent grains.

PubMed

Almeida, Catarina; Brányik, Tomás; Moradas-Ferreira, Pedro; Teixeira, José

2003-01-01

A yeast strain secreting endopolygalacturonase was used in this work to study the possibility of continuous production of this enzyme. It is a feasible and interesting alternative to fungal batch production essentially due to the specificity of the type of pectinase excreted by Kluyveromyces marxianus CCT 3172, to the lower broth viscosity and to the easier downstream operations. In order to increase the reactors' productivity, a cellulosic carrier obtained from barley spent grains was tested as an immobilization support. Two types of reactors were studied for pectinase production using glucose as a carbon and energy source--a continuous stirred tank reactor (CSTR) and a packed bed reactor (PBR) with recycled flow. The highest value for pectinase volumetric productivity (P(V)=0.98 U ml(-1) h(-1)) was achieved in the PBR for D=0.40 h(-1), a glucose concentration on the inlet of S(in)=20 g l(-1), and a biomass load in the support of X(i)=0.225 g g(-1). The results demonstrate the attractiveness of the packed bed system for pectinase production.

Technology, safety, and costs of decommissioning reference nuclear research and test reactors: sensitivity of decommissioning radiation exposure and costs to selected parameters

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

Konzek, G.J.

1983-07-01

Additional analyses of decommissioning at the reference research and test (R and T) reactors and analyses of five recent reactor decommissionings are made that examine some parameters not covered in the initial study report (NUREG/CR-1756). The parameters examined for decommissioning are: (1) the effect on costs and radiation exposure of plant size and/or type; (2) the effects on costs of increasing disposal charges and of unavailability of waste disposal capacity at licensed waste disposal facilities; and (3) the costs of and the available alternatives for the disposal of nuclear R and T reactor fuel assemblies.

Development of a Novel Catalytic Membrane Reactor for Heterogeneous Catalysis in Supercritical CO2

PubMed Central

Islam, Nazrul M.; Chatterjee, Maya; Ikushima, Yutaka; Yokoyama, Toshiro; Kawanami, Hajime

2010-01-01

A novel type of high-pressure membrane reactor has been developed for hydrogenation in supercritical carbon dioxide (scCO2). The main objectives of the design of the reactor are the separate feeding of hydrogen and substrate in scCO2 for safe reactions in a continuous flow process, and to reduce the reaction time. By using this new reactor, hydrogenation of cinnamaldehyde into hydrocinnamaldehyde has been successfully carried out with 100% selectivity at 50 °C in 10 MPa (H2: 1 MPa, CO2: 9 MPa) with a flow rate of substrate ranging from 0.05 to 1.0 mL/min. PMID:20162008

Reverse-Bumpy-Ball-Type-Nanoreactor-Loaded Nylon Membranes as Peroxidase-Mimic Membrane Reactors for a Colorimetric Assay for H₂O₂.

PubMed

Tong, Ying; Jiao, Xiangyu; Yang, Hankun; Wen, Yongqiang; Su, Lei; Zhang, Xueji

2016-04-01

Herein we report for the first time fabrication of reverse bumpy ball (RBB)-type-nanoreactor-based flexible peroxidase-mimic membrane reactors (MRs). The RBB-type nanoreactors with gold nanoparticles embedded in the inner walls of carbon shells were loaded on nylon membranes through a facile filtration approach. The as-prepared flexible catalytic membrane was studied as a peroxidase-mimic MR. It was found that the obtained peroxidase-mimic MR could exhibit several advantages over natural enzymes, such as facile and good recyclability, long-term stability and easy storage. Moreover, the RBB NS-modified nylon MRs as a peroxidase mimic provide a useful colorimetric assay for H₂O₂.

Flat-plate collector research area: Silicon material task

NASA Technical Reports Server (NTRS)

Lutwack, R.

1982-01-01

Silane decomposition in a fluidized-bed reactor (FBR) process development unit (PDU) to make semiconductor-grade Si is reviewed. The PDU was modified by installation of a new heating system to provide the required temperature profile and better control, and testing was resumed. A process for making trichlorosilane by the hydrochlorination of metallurgical-grade Si and silicon tetrachloride is reported. Fabrication and installation of the test system employing a new 2-in.-dia reactor was completed. A process that converts trichlorosilane to dichlorosilane (DCS), which is reduced by hydrogen to make Si by a chemical vapor deposition step in a Siemens-type reactor is described. Testing of the DCS PDU integraled with Si deposition reactors continued. Experiments in a 2-in.-dia reactor to define the operating window and to investigate the Si deposition kinetics were completed.

A simple, space constrained NIRIM type reactor for chemical vapour deposition of diamond

NASA Astrophysics Data System (ADS)

Thomas, Evan L. H.; Ginés, Laia; Mandal, Soumen; Klemencic, Georgina M.; Williams, Oliver A.

2018-03-01

In this paper the design of a simple, space constrained chemical vapour deposition reactor for diamond growth is detailed. Based on the design by NIRIM, the reactor is composed of a quartz discharge tube placed within a 2.45 GHz waveguide to create the conditions required for metastable growth of diamond. Utilising largely off-the-shelf components and a modular design, the reactor allows for easy modification, repair, and cleaning between growth runs. The elements of the reactor design are laid out with the CAD files, parts list, and control files made easily available to enable replication. Finally, the quality of nanocrystalline diamond films produced are studied with SEM and Raman spectroscopy, with the observation of clear faceting and a large diamond fraction suggesting the design offers deposition of diamond with minimal complexity.

Effect of small-scale biomass gasification at the state of refractory lining the fixed bed reactor

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

Janša, Jan, E-mail: jan.jansa@vsb.cz; Peer, Vaclav, E-mail: vaclav.peer@vsb.cz; Pavloková, Petra, E-mail: petra.pavlokova@vsb.cz

The article deals with the influence of biomass gasification on the condition of the refractory lining of a fixed bed reactor. The refractory lining of the gasifier is one part of the device, which significantly affects the operational reliability and durability. After removing the refractory lining of the gasifier from the experimental reactor, there was done an assessment how gasification of different kinds of biomass reflected on its condition in terms of the main factors affecting its life. Gasification of biomass is reflected on the lining, especially through sticking at the bottom of the reactor. Measures for prolonging the lifemore » of lining consist in the reduction of temperature in the reactor, in this case, in order to avoid ash fusion biomass which it is difficult for this type of gasifier.« less

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.

Current and prospective safety issues at the HFBR

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

Tichler, P.R.

The Brookhaven high-flux beam reactor (HFBR) was designed primarily to produce external neutron beams for experimental research. It is cooled, moderated, and reflected by heavy water and uses materials test reactor and engineering test reactor type of fuel elements containing enriched uranium. The reactor power when operation began in 1965 was 40 MW, was raised to 60 MW in 1982 after a number of plant modifications, and operated at that level until 1989. Since that time, safety questions have been raised that resulted in extended shutdowns and a reduction in operating power to 30 MW. This paper discusses the principalmore » safety issues and plans for their resolution and return to 60-MW operation. In addition, radiation embrittlement of the reactor vessel and thermal shield and its effect on the life of the facility are briefly discussed.« less

Thermal Neutron Tomography for Cultural Heritage at INR

NASA Astrophysics Data System (ADS)

Dinca, Marin; Mandescu, Dragos

The neutron and gamma imaging facility placed at the tangential channel of the TRIGA-ACPR from INR was used for tomography investigations on a test object with good results and shortly followed its involvement for tomography investigations on prehistoric statues of clay from the Arges County Museum. This activity was performed in connection with a research contract with IAEA with title ;The neutron and gamma imaging method combined with neutron-based analytical methods for cultural heritage research;, in the frame of a current CRP, that helps curators to reveal the internal structure and composition of the objects. The detector system has been developed based on two interchangeable scintillators, one for thermal neutrons and the other one for gamma radiations, a mirror of float glass coated with aluminum and two interchangeable CCD cameras. Experiments of tomography imaging for two prehistoric statues of clay with CCD STARLIGHT XPRESS SXV-H9 camera with XD-4 type image intensifier are presented in this paper. The tomography reconstructions with Octopus software have shown the potential of good results even for 100 projections/1800. This was a good opportunity for the dissemination of the investigation methods based on neutrons for cultural heritage and beyond this area.

Pyrolysis of waste tyres: A review

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

Williams, Paul T., E-mail: p.t.williams@leeds.ac.uk

2013-08-15

Graphical abstract: - Highlights: • Pyrolysis of waste tyres produces oil, gas and char, and recovered steel. • Batch, screw kiln, rotary kiln, vacuum and fluidised-bed are main reactor types. • Product yields are influenced by reactor type, temperature and heating rate. • Pyrolysis oils are complex and can be used as chemical feedstock or fuel. • Research into higher value products from the tyre pyrolysis process is reviewed. - Abstract: Approximately 1.5 billion tyres are produced each year which will eventually enter the waste stream representing a major potential waste and environmental problem. However, there is growing interest inmore » pyrolysis as a technology to treat tyres to produce valuable oil, char and gas products. The most common reactors used are fixed-bed (batch), screw kiln, rotary kiln, vacuum and fluidised-bed. The key influence on the product yield, and gas and oil composition, is the type of reactor used which in turn determines the temperature and heating rate. Tyre pyrolysis oil is chemically very complex containing aliphatic, aromatic, hetero-atom and polar fractions. The fuel characteristics of the tyre oil shows that it is similar to a gas oil or light fuel oil and has been successfully combusted in test furnaces and engines. The main gases produced from the pyrolysis of waste tyres are H{sub 2}, C{sub 1}–C{sub 4} hydrocarbons, CO{sub 2}, CO and H{sub 2}S. Upgrading tyre pyrolysis products to high value products has concentrated on char upgrading to higher quality carbon black and to activated carbon. The use of catalysts to upgrade the oil to a aromatic-rich chemical feedstock or the production of hydrogen from waste tyres has also been reported. Examples of commercial and semi-commercial scale tyre pyrolysis systems show that small scale batch reactors and continuous rotary kiln reactors have been developed to commercial scale.« less

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

A comparison of mass transfer coefficients between trickle-bed, hollow fiber membrane and stirred tank reactors.

PubMed

Orgill, James J; Atiyeh, Hasan K; Devarapalli, Mamatha; Phillips, John R; Lewis, Randy S; Huhnke, Raymond L

2013-04-01

Trickle-bed reactor (TBR), hollow fiber membrane reactor (HFR) and stirred tank reactor (STR) can be used in fermentation of sparingly soluble gasses such as CO and H2 to produce biofuels and bio-based chemicals. Gas fermenting reactors must provide high mass transfer capabilities that match the kinetic requirements of the microorganisms used. The present study compared the volumetric mass transfer coefficient (K(tot)A/V(L)) of three reactor types; the TBR with 3 mm and 6 mm beads, five different modules of HFRs, and the STR. The analysis was performed using O2 as the gaseous mass transfer agent. The non-porous polydimethylsiloxane (PDMS) HFR provided the highest K(tot)A/V(L) (1062 h(-1)), followed by the TBR with 6mm beads (421 h(-1)), and then the STR (114 h(-1)). The mass transfer characteristics in each reactor were affected by agitation speed, and gas and liquid flow rates. Furthermore, issues regarding the comparison of mass transfer coefficients are discussed. Copyright © 2013 Elsevier Ltd. All rights reserved.

A Parametric Sizing Model for Molten Regolith Electrolysis Reactors to Produce Oxygen from Lunar Regolith

NASA Technical Reports Server (NTRS)

Schreiner, Samuel S.; Dominguez, Jesus A.; Sibille, Laurent; Hoffman, Jeffrey A.

2015-01-01

We present a parametric sizing model for a Molten Electrolysis Reactor that produces oxygen and molten metals from lunar regolith. The model has a foundation of regolith material properties validated using data from Apollo samples and simulants. A multiphysics simulation of an MRE reactor is developed and leveraged to generate a vast database of reactor performance and design trends. A novel design methodology is created which utilizes this database to parametrically design an MRE reactor that 1) can sustain the required mass of molten regolith, current, and operating temperature to meet the desired oxygen production level, 2) can operate for long durations via joule heated, cold wall operation in which molten regolith does not touch the reactor side walls, 3) can support a range of electrode separations to enable operational flexibility. Mass, power, and performance estimates for an MRE reactor are presented for a range of oxygen production levels. The effects of several design variables are explored, including operating temperature, regolith type/composition, batch time, and the degree of operational flexibility.

The U.S. Geological Survey's TRIGA® reactor

USGS Publications Warehouse

DeBey, Timothy M.; Roy, Brycen R.; Brady, Sally R.

2012-01-01

The U.S. Geological Survey (USGS) operates a low-enriched uranium-fueled, pool-type reactor located at the Federal Center in Denver, Colorado. The mission of the Geological Survey TRIGA® Reactor (GSTR) is to support USGS science by providing information on geologic, plant, and animal specimens to advance methods and techniques unique to nuclear reactors. The reactor facility is supported by programs across the USGS and is organizationally under the Associate Director for Energy and Minerals, and Environmental Health. The GSTR is the only facility in the United States capable of performing automated delayed neutron analyses for detecting fissile and fissionable isotopes. Samples from around the world are submitted to the USGS for analysis using the reactor facility. Qualitative and quantitative elemental analyses, spatial elemental analyses, and geochronology are performed. Few research reactor facilities in the United States are equipped to handle the large number of samples processed at the GSTR. Historically, more than 450,000 sample irradiations have been performed at the USGS facility. Providing impartial scientific information to resource managers, planners, and other interested parties throughout the world is an integral part of the research effort of the USGS.

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

Reforming results of a novel radial reactor for a solid oxide fuel cell system with anode off-gas recirculation

NASA Astrophysics Data System (ADS)

Bosch, Timo; Carré, Maxime; Heinzel, Angelika; Steffen, Michael; Lapicque, François

2017-12-01

A novel reactor of a natural gas (NG) fueled, 1 kW net power solid oxide fuel cell (SOFC) system with anode off-gas recirculation (AOGR) is experimentally investigated. The reactor operates as pre-reformer, is of the type radial reactor with centrifugal z-flow, has the shape of a hollow cylinder with a volume of approximately 1 L and is equipped with two different precious metal wire-mesh catalyst packages as well as with an internal electric heater. Reforming investigations of the reactor are done stand-alone but as if the reactor would operate within the total SOFC system with AOGR. For the tests presented here it is assumed that the SOFC system runs on pure CH4 instead of NG. The manuscript focuses on the various phases of reactor operation during the startup process of the SOFC system. Startup process reforming experiments cover reactor operation points at which it runs on an oxygen to carbon ratio at the reactor inlet (ϕRI) of 1.2 with air supplied, up to a ϕRI of 2.4 without air supplied. As confirmed by a Monte Carlo simulation, most of the measured outlet gas concentrations are in or close to equilibrium.

A Single-Granule-Level Approach Reveals Ecological Heterogeneity in an Upflow Anaerobic Sludge Blanket Reactor

PubMed Central

Mei, Ran; Narihiro, Takashi; Bocher, Benjamin T. W.; Yamaguchi, Takashi; Liu, Wen-Tso

2016-01-01

Upflow anaerobic sludge blanket (UASB) reactor has served as an effective process to treat industrial wastewater such as purified terephthalic acid (PTA) wastewater. For optimal UASB performance, balanced ecological interactions between syntrophs, methanogens, and fermenters are critical. However, much of the interactions remain unclear because UASB have been studied at a “macro”-level perspective of the reactor ecosystem. In reality, such reactors are composed of a suite of granules, each forming individual micro-ecosystems treating wastewater. Thus, typical approaches may be oversimplifying the complexity of the microbial ecology and granular development. To identify critical microbial interactions at both macro- and micro- level ecosystem ecology, we perform community and network analyses on 300 PTA–degrading granules from a lab-scale UASB reactor and two full-scale reactors. Based on MiSeq-based 16S rRNA gene sequencing of individual granules, different granule-types co-exist in both full-scale reactors regardless of granule size and reactor sampling depth, suggesting that distinct microbial interactions occur in different granules throughout the reactor. In addition, we identify novel networks of syntrophic metabolic interactions in different granules, perhaps caused by distinct thermodynamic conditions. Moreover, unseen methanogenic relationships (e.g. “Candidatus Aminicenantes” and Methanosaeta) are observed in UASB reactors. In total, we discover unexpected microbial interactions in granular micro-ecosystems supporting UASB ecology and treatment through a unique single-granule level approach. PMID:27936088

Numerical study of the effects of lamp configuration and reactor wall roughness in an open channel water disinfection UV reactor.

PubMed

Sultan, Tipu

2016-07-01

This article describes the assessment of a numerical procedure used to determine the UV lamp configuration and surface roughness effects on an open channel water disinfection UV reactor. The performance of the open channel water disinfection UV reactor was numerically analyzed on the basis of the performance indictor reduction equivalent dose (RED). The RED values were calculated as a function of the Reynolds number to monitor the performance. The flow through the open channel UV reactor was modelled using a k-ε model with scalable wall function, a discrete ordinate (DO) model for fluence rate calculation, a volume of fluid (VOF) model to locate the unknown free surface, a discrete phase model (DPM) to track the pathogen transport, and a modified law of the wall to incorporate the reactor wall roughness effects. The performance analysis was carried out using commercial CFD software (ANSYS Fluent 15.0). Four case studies were analyzed based on open channel UV reactor type (horizontal and vertical) and lamp configuration (parallel and staggered). The results show that lamp configuration can play an important role in the performance of an open channel water disinfection UV reactor. The effects of the reactor wall roughness were Reynolds number dependent. The proposed methodology is useful for performance optimization of an open channel water disinfection UV reactor. Copyright © 2016 Elsevier Ltd. All rights reserved.

FALCON reactor-pumped laser description and program overview

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

None

1989-12-01

The FALCON (Fission Activated Laser CONcept) reactor-pumped laser program at Sandia National Laboratories is examining the feasibility of high-power systems pumped directly by the energy from a nuclear reactor. In this concept we use the highly energetic fission fragments from neutron induced fission to excite a large volume laser medium. This technology has the potential to scale to extremely large optical power outputs in a primarily self-powered device. A laser system of this type could also be relatively compact and capable of long run times without refueling.

Analysis of fluid fuel flow to the neutron kinetics on molten salt reactor FUJI-12

NASA Astrophysics Data System (ADS)

Aji, Indarta Kuncoro; Waris, Abdul; Permana, Sidik

2015-09-01

Molten Salt Reactor is a reactor are operating with molten salt fuel flowing. This condition interpret that the neutron kinetics of this reactor is affected by the flow rate of the fuel. This research analyze effect by the alteration velocity of the fuel by MSR type Fuji-12, with fuel composition LiF-BeF2-ThF4-233UF4 respectively 71.78%-16%-11.86%-0.36%. Calculation process in this study is performed numerically by SOR and finite difference method use C programming language. Data of reactivity, neutron flux, and the macroscopic fission cross section for calculation process obtain from SRAC-CITATION (Standard thermal Reactor Analysis Code) and JENDL-4.0 data library. SRAC system designed and developed by JAEA (Japan Atomic Energy Agency). This study aims to observe the effect of the velocity of fuel salt to the power generated from neutron precursors at fourth year of reactor operate (last critical condition) with number of multiplication effective; 1.0155.

Study of the diversity of microbial communities in a sequencing batch reactor oxic-settling-anaerobic process and its modified process.

PubMed

Sun, Lianpeng; Chen, Jianfan; Wei, Xiange; Guo, Wuzhen; Lin, Meishan; Yu, Xiaoyu

2016-05-01

To further reveal the mechanism of sludge reduction in the oxic-settling-anaerobic (OSA) process, the polymerase chain reaction - denaturing gradient gel electrophoresis protocol was used to study the possible difference in the microbial communities between a sequencing batch reactor (SBR)-OSA process and its modified process, by analyzing the change in the diversity of the microbial communities in each reactor of both systems. The results indicated that the structure of the microbial communities in aerobic reactors of the 2 processes was very different, but the predominant microbial populations in anaerobic reactors were similar. The predominant microbial population in the aerobic reactor of the SBR-OSA belonged to Burkholderia cepacia, class Betaproteobacteria, while those of the modified process belonged to the classes Alphaproteobacteria, Betaproteobacteria, and Gammaproteobacteria. These 3 types of microbes had a cryptic growth characteristic, which was the main cause of a greater sludge reduction efficiency achieved by the modified process.

MEANS FOR SHIELDING AND COOLING REACTORS

DOEpatents

Wigner, E.P.; Ohlinger, L.A.; Young, G.J.; Weinberg, A.M.

1959-02-10

Reactors of the water-cooled type and a means for shielding such a rcactor to protect operating personnel from harmful radiation are discussed. In this reactor coolant tubes which contain the fissionable material extend vertically through a mass of moderator. Liquid coolant enters through the bottom of the coolant tubes and passes upwardly over the fissionable material. A shield tank is disposed over the top of the reactor and communicates through its bottom with the upper end of the coolant tubes. A hydrocarbon shielding fluid floats on the coolant within the shield tank. With this arrangements the upper face of the reactor can be opened to the atmosphere through the two superimposed liquid layers. A principal feature of the invention is that in the event radioactive fission products enter thc coolant stream. imposed layer of hydrocarbon reduces the intense radioactivity introduced into the layer over the reactors and permits removal of the offending fuel material by personnel shielded by the uncontaminated hydrocarbon layer.

Development and optimization of water treatment reactors using TiO2-modified polymer beads with a refractive index identical to that of water

NASA Astrophysics Data System (ADS)

Myoga, Arata; Iwashita, Ryutaro; Unno, Noriyuki; Satake, Shin-ichi; Taniguchi, Jun; Yuki, Kazuhisa; Seki, Yohji

2018-03-01

Various water purification reactors were constructed using beads of TiO2-coated MEXFLON, which is a fluoropolymer exhibiting a refractive index identical to that of water. The performance of these reactors was evaluated in a recirculation experiment utilizing an aqueous solution of methylene blue. Reactor pipes (length = 150 mm, internal diameter = 10 mm) were made of a fluorinated ethylene polymer with a refractive index of 1.338 and contained 206-bead clusters. A UV lamp was used to irradiate eight reactor pipes surrounding it. The above-mentioned eight bead-packed pipes were connected both in series and in parallel, and the performances of these two reactor types were compared. A pseudo-first-order rate constant of 0.70 h- 1 was obtained for the series connection, whereas the corresponding value for the parallel connection was 1.5 times smaller, confirming the effectiveness of increasing the reaction surface by employing a larger number of beads.

Modeling residence-time distribution in horizontal screw hydrolysis reactors

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

Sievers, David A.; Stickel, Jonathan J.

The dilute-acid thermochemical hydrolysis step used in the production of liquid fuels from lignocellulosic biomass requires precise residence-time control to achieve high monomeric sugar yields. Difficulty has been encountered reproducing residence times and yields when small batch reaction conditions are scaled up to larger pilot-scale horizontal auger-tube type continuous reactors. A commonly used naive model estimated residence times of 6.2-16.7 min, but measured mean times were actually 1.4-2.2 the estimates. Here, this study investigated how reactor residence-time distribution (RTD) is affected by reactor characteristics and operational conditions, and developed a method to accurately predict the RTD based on key parameters.more » Screw speed, reactor physical dimensions, throughput rate, and process material density were identified as major factors affecting both the mean and standard deviation of RTDs. The general shape of RTDs was consistent with a constant value determined for skewness. The Peclet number quantified reactor plug-flow performance, which ranged between 20 and 357.« less

Modeling residence-time distribution in horizontal screw hydrolysis reactors

DOE PAGES

Sievers, David A.; Stickel, Jonathan J.

2017-10-12

The dilute-acid thermochemical hydrolysis step used in the production of liquid fuels from lignocellulosic biomass requires precise residence-time control to achieve high monomeric sugar yields. Difficulty has been encountered reproducing residence times and yields when small batch reaction conditions are scaled up to larger pilot-scale horizontal auger-tube type continuous reactors. A commonly used naive model estimated residence times of 6.2-16.7 min, but measured mean times were actually 1.4-2.2 the estimates. Here, this study investigated how reactor residence-time distribution (RTD) is affected by reactor characteristics and operational conditions, and developed a method to accurately predict the RTD based on key parameters.more » Screw speed, reactor physical dimensions, throughput rate, and process material density were identified as major factors affecting both the mean and standard deviation of RTDs. The general shape of RTDs was consistent with a constant value determined for skewness. The Peclet number quantified reactor plug-flow performance, which ranged between 20 and 357.« less

Hanging core support system for a nuclear reactor

DOEpatents

Burelbach, James P.; Kann, William J.; Pan, Yen-Cheng; Saiveau, James G.; Seidensticker, Ralph W.

1987-01-01

For holding the reactor core in the confining reactor vessel, a support is disclosed that is structurally independent of the vessel, that is dimensionally accurate and stable, and that comprises tandem tension linkages that act redundantly of one another to maintain stabilized core support even in the unlikely event of the complete failure of one of the linkages. The core support has a mounting platform for the reactor core, and unitary structure including a flange overlying the top edge of the reactor vessels, and a skirt and box beams between the flange and platform for establishing one of the linkages. A plurality of tension rods connect between the deck closing the reactor vessel and the platform for establishing the redundant linkage. Loaded Belleville springs flexibly hold the tension rods at the deck and separable bayonet-type connections hold the tension rods at the platform. Motion or radiation sensing detectors can be provide at the lower ends of the tension rods for obtaining pertinent readings proximate the core.

Development and optimization of water treatment reactors using TiO2-modified polymer beads with a refractive index identical to that of water

NASA Astrophysics Data System (ADS)

Myoga, Arata; Iwashita, Ryutaro; Unno, Noriyuki; Satake, Shin-ichi; Taniguchi, Jun; Yuki, Kazuhisa; Seki, Yohji

2018-06-01

Various water purification reactors were constructed using beads of TiO2-coated MEXFLON, which is a fluoropolymer exhibiting a refractive index identical to that of water. The performance of these reactors was evaluated in a recirculation experiment utilizing an aqueous solution of methylene blue. Reactor pipes (length = 150 mm, internal diameter = 10 mm) were made of a fluorinated ethylene polymer with a refractive index of 1.338 and contained 206-bead clusters. A UV lamp was used to irradiate eight reactor pipes surrounding it. The above-mentioned eight bead-packed pipes were connected both in series and in parallel, and the performances of these two reactor types were compared. A pseudo-first-order rate constant of 0.70 h- 1 was obtained for the series connection, whereas the corresponding value for the parallel connection was 1.5 times smaller, confirming the effectiveness of increasing the reaction surface by employing a larger number of beads.

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

NASA Astrophysics Data System (ADS)

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

2013-03-01

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

Hardness of AISI type 410 martensitic steels after high temperature irradiation via nanoindentation

NASA Astrophysics Data System (ADS)

Waseem, Owais Ahmed; Jeong, Jong-Ryul; Park, Byong-Guk; Maeng, Cheol-Soo; Lee, Myoung-Goo; Ryu, Ho Jin

2017-11-01

The hardness of irradiated AISI type 410 martensitic steel, which is utilized in structural and magnetic components of nuclear power plants, is investigated in this study. Proton irradiation of AISI type 410 martensitic steel samples was carried out by exposing the samples to 3 MeV protons up to a 1.0 × 1017 p/cm2 fluence level at a representative nuclear reactor coolant temperature of 350 °C. The assessment of deleterious effects of irradiation on the micro-structure and mechanical behavior of the AISI type 410 martensitic steel samples via transmission electron microscopy-energy dispersive spectroscopy and cross-sectional nano-indentation showed no significant variation in the microscopic or mechanical characteristics. These results ensure the integrity of the structural and magnetic components of nuclear reactors made of AISI type 410 martensitic steel under high-temperature irradiation damage levels up to approximately 5.2 × 10-3 dpa.

Energy from the Atom. A Basic Teaching Unit on Energy. Revised.

ERIC Educational Resources Information Center

McDermott, Hugh, Ed.; Scharmann, Larry, Ed.

Recommended for grades 9-12 social studies and/or physical science classes, this 4-8 day unit focuses on four topics: (1) the background and history of atomic development; (2) two common types of nuclear reactors (boiling water and pressurized water reactors); (3) disposal of radioactive waste; and (4) the future of nuclear energy. Each topic…

Analysis of standard reference materials by absolute INAA

NASA Astrophysics Data System (ADS)

Heft, R. E.; Koszykowski, R. F.

1981-07-01

Three standard reference materials: flyash, soil, and ASI 4340 steel, are analyzed by a method of absolute instrumental neutron activation analysis. Two different light water pool-type reactors were used to produce equivalent analytical results even though the epithermal to thermal flux ratio in one reactor was higher than that in the other by a factor of two.

NUCLEAR REACTOR FUEL ELEMENT

DOEpatents

Currier, E.L. Jr.; Nicklas, J.H.

1963-06-11

A fuel plate is designed for incorporation into control rods of the type utilized in high-flux test reactors. The fuel plate is designed so that the portion nearest the poison section of the control rod contains about one-half as much fissionable material as in the rest of the plate, thereby eliminating dangerous flux peaking in that portion. (AEC)

On the equilibrium isotopic composition of the thorium-uranium-plutonium fuel cycle

NASA Astrophysics Data System (ADS)

Marshalkin, V. Ye.; Povyshev, V. M.

2016-12-01

The equilibrium isotopic compositions and the times to equilibrium in the process of thorium-uranium-plutonium oxide fuel recycling in VVER-type reactors using heavy water mixed with light water are estimated. It is demonstrated thEhfat such reactors have a capacity to operate with self-reproduction of active isotopes in the equilibrium mode.

On the equilibrium isotopic composition of the thorium–uranium–plutonium fuel cycle

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

Marshalkin, V. Ye., E-mail: marshalkin@vniief.ru; Povyshev, V. M.

2016-12-15

The equilibrium isotopic compositions and the times to equilibrium in the process of thorium–uranium–plutonium oxide fuel recycling in VVER-type reactors using heavy water mixed with light water are estimated. It is demonstrated thEhfat such reactors have a capacity to operate with self-reproduction of active isotopes in the equilibrium mode.

Comparison of the microstructure, deformation and crack initiation behavior of austenitic stainless steel irradiated in-reactor or with protons

NASA Astrophysics Data System (ADS)

Stephenson, Kale J.; Was, Gary S.

2015-01-01

The objective of this study was to compare the microstructures, microchemistry, hardening, susceptibility to IASCC initiation, and deformation behavior resulting from proton or reactor irradiation. Two commercial purity and six high purity austenitic stainless steels with various solute element additions were compared. Samples of each alloy were irradiated in the BOR-60 fast reactor at 320 °C to doses between approximately 4 and 12 dpa or by a 3.2 MeV proton beam at 360 °C to a dose of 5.5 dpa. Irradiated microstructures consisted mainly of dislocation loops, which were similar in size but lower in density after proton irradiation. Both irradiation types resulted in the formation of Ni-Si rich precipitates in a high purity alloy with added Si, but several other high purity neutron irradiated alloys showed precipitation that was not observed after proton irradiation, likely due to their higher irradiation dose. Low densities of small voids were observed in several high purity proton irradiated alloys, and even lower densities in neutron irradiated alloys, implying void nucleation was in process. Elemental segregation at grain boundaries was very similar after each irradiation type. Constant extension rate tensile experiments on the alloys in simulated light water reactor environments showed excellent agreement in terms of the relative amounts of intergranular cracking, and an analysis of localized deformation after straining showed a similar response of cracking to surface step height after both irradiation types. Overall, excellent agreement was observed after proton and reactor irradiation, providing additional evidence that proton irradiation is a useful tool for accelerated testing of irradiation effects in austenitic stainless steel.

Application of ATHLET/DYN3D coupled codes system for fast liquid metal cooled reactor steady state simulation

NASA Astrophysics Data System (ADS)

Ivanov, V.; Samokhin, A.; Danicheva, I.; Khrennikov, N.; Bouscuet, J.; Velkov, K.; Pasichnyk, I.

2017-01-01

In this paper the approaches used for developing of the BN-800 reactor test model and for validation of coupled neutron-physic and thermohydraulic calculations are described. Coupled codes ATHLET 3.0 (code for thermohydraulic calculations of reactor transients) and DYN3D (3-dimensional code of neutron kinetics) are used for calculations. The main calculation results of reactor steady state condition are provided. 3-D model used for neutron calculations was developed for start reactor BN-800 load. The homogeneous approach is used for description of reactor assemblies. Along with main simplifications, the main reactor BN-800 core zones are described (LEZ, MEZ, HEZ, MOX, blankets). The 3D neutron physics calculations were provided with 28-group library, which is based on estimated nuclear data ENDF/B-7.0. Neutron SCALE code was used for preparation of group constants. Nodalization hydraulic model has boundary conditions by coolant mass-flow rate for core inlet part, by pressure and enthalpy for core outlet part, which can be chosen depending on reactor state. Core inlet and outlet temperatures were chosen according to reactor nominal state. The coolant mass flow rate profiling through the core is based on reactor power distribution. The test thermohydraulic calculations made with using of developed model showed acceptable results in coolant mass flow rate distribution through the reactor core and in axial temperature and pressure distribution. The developed model will be upgraded in future for different transient analysis in metal-cooled fast reactors of BN type including reactivity transients (control rods withdrawal, stop of the main circulation pump, etc.).

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

Biological hydrogen production by Clostridium acetobutylicum in an unsaturated flow reactor.

PubMed

Zhang, Husen; Bruns, Mary Ann; Logan, Bruce E

2006-02-01

A mesophilic unsaturated flow (trickle bed) reactor was designed and tested for H2 production via fermentation of glucose. The reactor consisted of a column packed with glass beads and inoculated with a pure culture (Clostridium acetobutylicum ATCC 824). A defined medium containing glucose was fed at a flow rate of 1.6 mL/min (0.096 L/h) into the capped reactor, producing a hydraulic retention time of 2.1 min. Gas-phase H2 concentrations were constant, averaging 74 +/- 3% for all conditions tested. H2 production rates increased from 89 to 220 mL/hL of reactor when influent glucose concentrations were varied from 1.0 to 10.5 g/L. Specific H2 production rate ranged from 680 to 1270 mL/g glucose per liter of reactor (total volume). The H2 yield was 15-27%, based on a theoretical limit by fermentation of 4 moles of H2 from 1 mole of glucose. The major fermentation by-products in the liquid effluent were acetate and butyrate. The reactor rapidly (within 60-72 h) became clogged with biomass, requiring manual cleaning of the system. In order to make long-term operation of the reactor feasible, biofilm accumulation in the reactor will need to be controlled through some process such as backwashing. These tests using an unsaturated flow reactor demonstrate the feasibility of the process to produce high H2 gas concentrations in a trickle-bed type of reactor. A likely application of this reactor technology could be H2 gas recovery from pre-treatment of high carbohydrate-containing wastewaters.

Neutronic safety parameters and transient analyses for Poland's MARIA research reactor.

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

Bretscher, M. M.; Hanan, N. A.; Matos, J. E.

1999-09-27

Reactor kinetic parameters, reactivity feedback coefficients, and control rod reactivity worths have been calculated for the MARIA Research Reactor (Swierk, Poland) for M6-type fuel assemblies with {sup 235}U enrichments of 80% and 19.7%. Kinetic parameters were evaluated for family-dependent effective delayed neutron fractions, decay constants, and prompt neutron lifetimes and neutron generation times. Reactivity feedback coefficients were determined for fuel Doppler coefficients, coolant (H{sub 2}O) void and temperature coefficients, and for in-core and ex-core beryllium temperature coefficients. Total and differential control rod worths and safety rod worths were calculated for each fuel type. These parameters were used to calculate genericmore » transients for fast and slow reactivity insertions with both HEU and LEU fuels. The analyses show that the HEU and LEU cores have very similar responses to these transients.« less

Preparation of dilute magnetic semiconductor films by metalorganic chemical vapor deposition

NASA Technical Reports Server (NTRS)

Nouhi, Akbar (Inventor); Stirn, Richard J. (Inventor)

1988-01-01

A method for preparation of a dilute magnetic semiconductor (DMS) film is provided, in which a Group II metal source, a Group VI metal source and a transition metal magnetic ion source are pyrolyzed in the reactor of a metalorganic chemical vapor deposition (MOCVD) system by contact with a heated substrate. As an example, the preparation of films of Cd(sub 1-x)Mn(sub x)Te, in which 0 is less than or equal to x less than or equal to 0.7, on suitable substrates (e.g., GaAs) is described. As a source of manganese, tricarbonyl (methylcyclopentadienyl) manganese (TCPMn) is employed. To prevent TCPMn condensation during its introduction into the reactor, the gas lines, valves and reactor tubes are heated. A thin-film solar cell of n-i-p structure, in which the i-type layer comprises a DMS, is also described; the i-type layer is suitably prepared by MOCVD.

Preparation of dilute magnetic semiconductor films by metalorganic chemical vapor deposition

NASA Technical Reports Server (NTRS)

Nouhi, Akbar (Inventor); Stirn, Richard J. (Inventor)

1990-01-01

A method for preparation of a dilute magnetic semiconductor (DMS) film is provided, wherein a Group II metal source, a Group VI metal source and a transition metal magnetic ion source are pyrolyzed in the reactor of a metalorganic chemical vapor deposition (MOCVD) system by contact with a heated substrate. As an example, the preparation of films of Cd.sub.1-x Mn.sub.x Te, wherein 0.ltoreq..times..ltoreq.0.7, on suitable substrates (e.g., GaAs) is described. As a source of manganese, tricarbonyl (methylcyclopentadienyl) maganese (TCPMn) is employed. To prevent TCPMn condensation during the introduction thereof int the reactor, the gas lines, valves and reactor tubes are heated. A thin-film solar cell of n-i-p structure, wherein the i-type layer comprises a DMS, is also described; the i-type layer is suitably prepared by MOCVD.

Scale Effects on Magnet Systems of Heliotron-Type Reactors

NASA Astrophysics Data System (ADS)

S, Imagawa; A, Sagara

2005-02-01

For power plants heliotron-type reactors have attractive advantages, such as no current-disruptions, no current-drive, and wide space between helical coils for the maintenance of in-vessel components. However, one disadvantage is that a major radius has to be large enough to obtain large Q-value or to produce sufficient space for blankets. Although the larger radius is considered to increase the construction cost, the influence has not been understood clearly, yet. Scale effects on superconducting magnet systems have been estimated under the conditions of a constant energy confinement time and similar geometrical parameters. Since the necessary magnetic field with a larger radius becomes lower, the increase rate of the weight of the coil support to the major radius is less than the square root. The necessary major radius will be determined mainly by the blanket space. The appropriate major radius will be around 13 m for a reactor similar to the Large Helical Device (LHD).

Experimental Criticality Benchmarks for SNAP 10A/2 Reactor Cores

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

Krass, A.W.

2005-12-19

This report describes computational benchmark models for nuclear criticality derived from descriptions of the Systems for Nuclear Auxiliary Power (SNAP) Critical Assembly (SCA)-4B experimental criticality program conducted by Atomics International during the early 1960's. The selected experimental configurations consist of fueled SNAP 10A/2-type reactor cores subject to varied conditions of water immersion and reflection under experimental control to measure neutron multiplication. SNAP 10A/2-type reactor cores are compact volumes fueled and moderated with the hydride of highly enriched uranium-zirconium alloy. Specifications for the materials and geometry needed to describe a given experimental configuration for a model using MCNP5 are provided. Themore » material and geometry specifications are adequate to permit user development of input for alternative nuclear safety codes, such as KENO. A total of 73 distinct experimental configurations are described.« less

Vibration and acoustic noise emitted by dry-type air-core reactors under PWM voltage excitation

NASA Astrophysics Data System (ADS)

Li, Jingsong; Wang, Shanming; Hong, Jianfeng; Yang, Zhanlu; Jiang, Shengqian; Xia, Shichong

2018-05-01

According to coupling way between the magnetic field and the structural order, structure mode is discussed by engaging finite element (FE) method and both natural frequency and modal shape for a dry-type air-core reactor (DAR) are obtained in this paper. On the basis of harmonic response analysis, electromagnetic force under PWM (Pulse Width Modulation) voltage excitation is mapped with the structure mesh, the vibration spectrum is gained and the consequences represents that the whole structure vibration predominates in the radial direction, with less axial vibration. Referring to the test standard of reactor noise, the rules of emitted noise of the DAR are measured and analyzed at chosen switching frequency matches the sample resonant frequency and the methods of active vibration and noise reduction are put forward. Finally, the low acoustic noise emission of a prototype DAR is verified by measurement.

The slow and fast pyrolysis of cherry seed.

PubMed

Duman, Gozde; Okutucu, Cagdas; Ucar, Suat; Stahl, Ralph; Yanik, Jale

2011-01-01

The slow and fast pyrolysis of cherry seeds (CWS) and cherry seeds shells (CSS) was studied in fixed-bed and fluidized bed reactors at different pyrolysis temperatures. The effects of reactor type and temperature on the yields and composition of products were investigated. In the case of fast pyrolysis, the maximum bio-oil yield was found to be about 44 wt% at pyrolysis temperature of 500 °C for both CWS and CSS, whereas the bio yields were of 21 and 15 wt% obtained at 500 °C from slow pyrolysis of CWS and CSS, respectively. Both temperature and reactor type affected the composition of bio-oils. The results showed that bio-oils obtained from slow pyrolysis of CWS and CSS can be used as a fuel for combustion systems in industry and the bio-oil produced from fast pyrolysis can be evaluated as a chemical feedstock. Copyright © 2010 Elsevier Ltd. All rights reserved.

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

NASA Astrophysics Data System (ADS)

Doshi, Bharat; Reddy, D. Chenna

2017-04-01

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

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

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

Skutnik, Steven E.

The goal of this project, “ORIGEN-based Nuclear Fuel Depletion Module for Fuel Cycle Assessment" is to create a physics-based reactor depletion and decay module for the Cyclus nuclear fuel cycle simulator in order to assess nuclear fuel inventories over a broad space of reactor operating conditions. The overall goal of this approach is to facilitate evaluations of nuclear fuel inventories for a broad space of scenarios, including extended used nuclear fuel storage and cascading impacts on fuel cycle options such as actinide recovery in used nuclear fuel, particularly for multiple recycle scenarios. The advantages of a physics-based approach (compared tomore » a recipe-based approach which has been typically employed for fuel cycle simulators) is in its inherent flexibility; such an approach can more readily accommodate the broad space of potential isotopic vectors that may be encountered under advanced fuel cycle options. In order to develop this flexible reactor analysis capability, we are leveraging the Origen nuclear fuel depletion and decay module from SCALE to produce a standalone “depletion engine” which will serve as the kernel of a Cyclus-based reactor analysis module. The ORIGEN depletion module is a rigorously benchmarked and extensively validated tool for nuclear fuel analysis and thus its incorporation into the Cyclus framework can bring these capabilities to bear on the problem of evaluating long-term impacts of fuel cycle option choices on relevant metrics of interest, including materials inventories and availability (for multiple recycle scenarios), long-term waste management and repository impacts, etc. Developing this Origen-based analysis capability for Cyclus requires the refinement of the Origen analysis sequence to the point where it can reasonably be compiled as a standalone sequence outside of SCALE; i.e., wherein all of the computational aspects of Origen (including reactor cross-section library processing and interpolation, input and output processing, and depletion/decay solvers) can be self-contained into a single executable sequence. Further, to embed this capability into other software environments (such as the Cyclus fuel cycle simulator) requires that Origen’s capabilities be encapsulated into a portable, self-contained library which other codes can then call directly through function calls, thereby directly accessing the solver and data processing capabilities of Origen. Additional components relevant to this work include modernization of the reactor data libraries used by Origen for conducting nuclear fuel depletion calculations. This work has included the development of new fuel assembly lattices not previously available (such as for CANDU heavy-water reactor assemblies) as well as validation of updated lattices for light-water reactors updated to employ modern nuclear data evaluations. The CyBORG reactor analysis module as-developed under this workscope is fully capable of dynamic calculation of depleted fuel compositions from all commercial U.S. reactor assembly types as well as a number of international fuel types, including MOX, VVER, MAGNOX, and PHWR CANDU fuel assemblies. In addition, the Origen-based depletion engine allows for CyBORG to evaluate novel fuel assembly and reactor design types via creation of Origen reactor data libraries via SCALE. The establishment of this new modeling capability affords fuel cycle modelers a substantially improved ability to model dynamically-changing fuel cycle and reactor conditions, including recycled fuel compositions from fuel cycle scenarios involving material recycle into thermal-spectrum systems.« less

Integration of Nine Steps into One Membrane Reactor To Produce Synthesis Gases for Ammonia and Liquid Fuel.

PubMed

Li, Wenping; Zhu, Xuefeng; Chen, Shuguang; Yang, Weishen

2016-07-18

The synthesis of ammonia and liquid fuel are two important chemical processes in which most of the energy is consumed in the production of H2 /N2 and H2 /CO synthesis gases from natural gas (methane). Here, we report a membrane reactor with a mixed ionic-electronic conducting membrane, in which the nine steps for the production of the two types of synthesis gases are shortened to one step by using water, air, and methane as feeds. In the membrane reactor, there is no direct CO2 emission and no CO or H2 S present in the ammonia synthesis gas. The energy consumption for the production of the two synthesis gases can be reduced by 63 % by using this membrane reactor. This promising membrane reactor process has been successfully demonstrated by experiment. © 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

CALANDRIA TYPE SODIUM GRAPHITE REACTOR

DOEpatents

Peterson, R.M.; Mahlmeister, J.E.; Vaughn, N.E.; Sanders, W.J.; Williams, A.C.

1964-02-11

A sodium graphite power reactor in which the unclad graphite moderator and fuel elements are contained within a core tank is described. The core tank is submersed in sodium within the reactor vessel. Extending longitudinally through the core thnk are process tubes with fuel elements positioned therein. A bellows sealing means allows axial expansion and construction of the tubes. Within the core tank, a leakage plenum is located below the graphite, and above the graphite is a gas space. A vent line regulates the gas pressure in the space, and another line removes sodium from the plenum. The sodium coolant flows from the lower reactor vessel through the annular space between the fuel elements and process tubes and out into the reactor vessel space above the core tank. From there, the heated coolant is drawn off through an outlet line and sent to the heat exchange. (AEC)

Observed Changes in As-Fabricated U-10Mo Monolithic Fuel Microstructures After Irradiation in the Advanced Test Reactor

NASA Astrophysics Data System (ADS)

Keiser, Dennis; Jue, Jan-Fong; Miller, Brandon; Gan, Jian; Robinson, Adam; Madden, James

2017-12-01

A low-enriched uranium U-10Mo monolithic nuclear fuel is being developed by the Material Management and Minimization Program, earlier known as the Reduced Enrichment for Research and Test Reactors Program, for utilization in research and test reactors around the world that currently use high-enriched uranium fuels. As part of this program, reactor experiments are being performed in the Advanced Test Reactor. It must be demonstrated that this fuel type exhibits mechanical integrity, geometric stability, and predictable behavior to high powers and high fission densities in order for it to be a viable fuel for qualification. This paper provides an overview of the microstructures observed at different regions of interest in fuel plates before and after irradiation for fuel samples that have been tested. These fuel plates were fabricated using laboratory-scale fabrication methods. Observations regarding how microstructural changes during irradiation may impact fuel performance are discussed.

Ceramic membrane microfilter as an immobilized enzyme reactor.

PubMed

Harrington, T J; Gainer, J L; Kirwan, D J

1992-10-01

This study investigated the use of a ceramic microfilter as an immobilized enzyme reactor. In this type of reactor, the substrate solution permeates the ceramic membrane and reacts with an enzyme that has been immobilized within its porous interior. The objective of this study was to examine the effect of permeation rate on the observed kinetic parameters for the immobilized enzyme in order to assess possible mass transfer influences or shear effects. Kinetic parameters were found to be independent of flow rate for immobilized penicillinase and lactate dehydrogenase. Therefore, neither mass transfer nor shear effects were observed for enzymes immobilized within the ceramic membrane. Both the residence time and the conversion in the microfilter reactor could be controlled simply by regulating the transmembrane pressure drop. This study suggests that a ceramic microfilter reactor can be a desirable alternative to a packed bed of porous particles, especially when an immobilized enzyme has high activity and a low Michaelis constant.

Fast Breeder Reactors in Sweden: Vision and Reality.

PubMed

Fjaestad, Maja

2015-01-01

The fast breeder is a type of nuclear reactor that aroused much attention in the 1950s and '60s. Its ability to produce more nuclear fuel than it consumes offered promises of cheap and reliable energy. Sweden had advanced plans for a nuclear breeder program, but canceled them in the middle of the 1970s with the rise of nuclear skepticism. The article investigates the nuclear breeder as a technological vision. The nuclear breeder reactor is an example of a technological future that did not meet its industrial expectations. But that does not change the fact that the breeder was an influential technology. Decisions about the contemporary reactors were taken with the idea that in a foreseeable future they would be replaced with the efficient breeder. The article argues that general themes in the history of the breeder reactor can deepen our understanding of the mechanisms behind technological change.

NEUTRONIC REACTOR

DOEpatents

Metcalf, H.E.

1957-10-01

A reactor of the type which preferably uses plutonium as the fuel and a liquid moderator, preferably ordinary water, and which produces steam within the reactor core due to the heat of the chain reaction is described. In the reactor shown the fuel elements are essentially in the form of trays and are ventically stacked in spaced relationship. The water moderator is continuously supplied to the trays to maintain a constant level on the upper surfaces of the fuel element as it is continually evaporated by the heat. The steam passes out through the spaces between the fuel elements and is drawn off at the top of the core. The fuel elements are clad in aluminum to prevent deterioration thereof with consequent contamimation of the water.

Shuttle APS propellant thermal conditioner study

NASA Technical Reports Server (NTRS)

Fulton, D. L.

1971-01-01

The conditioner design concept selected for evaluation consists of an integral reactor and baffle-type heat exchanger. Heat exchange is accomplished by flowing reactor hot gases past a series of slotted and formed plates, through which the conditioned propellant flows. Heat transfer analysis has resulted in the selection of a reactor hot gas nominal mixture ratio of 1.0, giving a combustion temperature of 1560 F with a hydrogen inlet temperature of 275 R. Worst case conditions result in a combustion gas temperature of 2060 F, satisfying the condition of no damage to the conditioner in case of failure to flow cold fluid. In addition, evaluation of hot gas flow requirements and conditioner weight has resulted in the selection of a reactor hot gas exhaust temperature of 750 R.

Shift Verification and Validation

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

Pandya, Tara M.; Evans, Thomas M.; Davidson, Gregory G

2016-09-07

This documentation outlines the verification and validation of Shift for the Consortium for Advanced Simulation of Light Water Reactors (CASL). Five main types of problems were used for validation: small criticality benchmark problems; full-core reactor benchmarks for light water reactors; fixed-source coupled neutron-photon dosimetry benchmarks; depletion/burnup benchmarks; and full-core reactor performance benchmarks. We compared Shift results to measured data and other simulated Monte Carlo radiation transport code results, and found very good agreement in a variety of comparison measures. These include prediction of critical eigenvalue, radial and axial pin power distributions, rod worth, leakage spectra, and nuclide inventories over amore » burn cycle. Based on this validation of Shift, we are confident in Shift to provide reference results for CASL benchmarking.« less

Assessment of the Technical Maturity of Generation IV Concepts for Test or Demonstration Reactor Applications, Revision 2

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

Gougar, Hans David

2015-10-01

The United States Department of Energy (DOE) commissioned a study the suitability of different advanced reactor concepts to support materials irradiations (i.e. a test reactor) or to demonstrate an advanced power plant/fuel cycle concept (demonstration reactor). As part of the study, an assessment of the technical maturity of the individual concepts was undertaken to see which, if any, can support near-term deployment. A Working Group composed of the authors of this document performed the maturity assessment using the Technical Readiness Levels as defined in DOE’s Technology Readiness Guide . One representative design was selected for assessment from of each ofmore » the six Generation-IV reactor types: gas-cooled fast reactor (GFR), lead-cooled fast reactor (LFR), molten salt reactor (MSR), supercritical water-cooled reactor (SCWR), sodium-cooled fast reactor (SFR), and very high temperature reactor (VHTR). Background information was obtained from previous detailed evaluations such as the Generation-IV Roadmap but other technical references were also used including consultations with concept proponents and subject matter experts. Outside of Generation IV activity in which the US is a party, non-U.S. experience or data sources were generally not factored into the evaluations as one cannot assume that this data is easily available or of sufficient quality to be used for licensing a US facility. The Working Group established the scope of the assessment (which systems and subsystems needed to be considered), adapted a specific technology readiness scale, and scored each system through discussions designed to achieve internal consistency across concepts. In general, the Working Group sought to determine which of the reactor options have sufficient maturity to serve either the test or demonstration reactor missions.« less

Preliminary study on aerobic granular biomass formation with aerobic continuous flow reactor

NASA Astrophysics Data System (ADS)

Yulianto, Andik; Soewondo, Prayatni; Handajani, Marissa; Ariesyady, Herto Dwi

2017-03-01

A paradigm shift in waste processing is done to obtain additional benefits from treated wastewater. By using the appropriate processing, wastewater can be turned into a resource. The use of aerobic granular biomass (AGB) can be used for such purposes, particularly for the processing of nutrients in wastewater. During this time, the use of AGB for processing nutrients more reactors based on a Sequencing Batch Reactor (SBR). Studies on the use of SBR Reactor for AGB demonstrate satisfactory performance in both formation and use. SBR reactor with AGB also has been applied on a full scale. However, the use use of SBR reactor still posses some problems, such as the need for additional buffer tank and the change of operation mode from conventional activated sludge to SBR. This gives room for further reactor research with the use of a different type, one of which is a continuous reactor. The purpose of this study is to compare AGB formation using continuous reactor and SBR with same operation parameter. Operation parameter are Organic Loading Rate (OLR) set to 2,5 Kg COD/m3.day with acetate as substrate, aeration rate 3 L/min, and microorganism from Hospital WWTP as microbial source. SBR use two column reactor with volumes 2 m3, and continuous reactor uses continuous airlift reactor, with two compartments and working volume of 5 L. Results from preliminary research shows that although the optimum results are not yet obtained, AGB can be formed on the continuous reactor. When compared with AGB generated by SBR, then the characteristics of granular diameter showed similarities, while the sedimentation rate and Sludge Volume Index (SVI) characteristics showed lower yields.

Cultivation of methanogenic community from subseafloor sediments using a continuous-flow bioreactor

PubMed Central

Imachi, Hiroyuki; Aoi, Ken; Tasumi, Eiji; Saito, Yumi; Yamanaka, Yuko; Saito, Yayoi; Yamaguchi, Takashi; Tomaru, Hitoshi; Takeuchi, Rika; Morono, Yuki; Inagaki, Fumio; Takai, Ken

2011-01-01

Microbial methanogenesis in subseafloor sediments is a key process in the carbon cycle on the Earth. However, the cultivation-dependent evidences have been poorly demonstrated. Here we report the cultivation of a methanogenic microbial consortium from subseafloor sediments using a continuous-flow-type bioreactor with polyurethane sponges as microbial habitats, called down-flow hanging sponge (DHS) reactor. We anaerobically incubated methane-rich core sediments collected from off Shimokita Peninsula, Japan, for 826 days in the reactor at 10 °C. Synthetic seawater supplemented with glucose, yeast extract, acetate and propionate as potential energy sources was provided into the reactor. After 289 days of operation, microbiological methane production became evident. Fluorescence in situ hybridization analysis revealed the presence of metabolically active microbial cells with various morphologies in the reactor. DNA- and RNA-based phylogenetic analyses targeting 16S rRNA indicated the successful growth of phylogenetically diverse microbial components during cultivation in the reactor. Most of the phylotypes in the reactor, once it made methane, were more closely related to culture sequences than to the subsurface environmental sequence. Potentially methanogenic phylotypes related to the genera Methanobacterium, Methanococcoides and Methanosarcina were predominantly detected concomitantly with methane production, while uncultured archaeal phylotypes were also detected. Using the methanogenic community enrichment as subsequent inocula, traditional batch-type cultivations led to the successful isolation of several anaerobic microbes including those methanogens. Our results substantiate that the DHS bioreactor is a useful system for the enrichment of numerous fastidious microbes from subseafloor sediments and will enable the physiological and ecological characterization of pure cultures of previously uncultivated subseafloor microbial life. PMID:21654849

Cultivation of methanogenic community from subseafloor sediments using a continuous-flow bioreactor.

PubMed

Imachi, Hiroyuki; Aoi, Ken; Tasumi, Eiji; Saito, Yumi; Yamanaka, Yuko; Saito, Yayoi; Yamaguchi, Takashi; Tomaru, Hitoshi; Takeuchi, Rika; Morono, Yuki; Inagaki, Fumio; Takai, Ken

2011-12-01

Microbial methanogenesis in subseafloor sediments is a key process in the carbon cycle on the Earth. However, the cultivation-dependent evidences have been poorly demonstrated. Here we report the cultivation of a methanogenic microbial consortium from subseafloor sediments using a continuous-flow-type bioreactor with polyurethane sponges as microbial habitats, called down-flow hanging sponge (DHS) reactor. We anaerobically incubated methane-rich core sediments collected from off Shimokita Peninsula, Japan, for 826 days in the reactor at 10 °C. Synthetic seawater supplemented with glucose, yeast extract, acetate and propionate as potential energy sources was provided into the reactor. After 289 days of operation, microbiological methane production became evident. Fluorescence in situ hybridization analysis revealed the presence of metabolically active microbial cells with various morphologies in the reactor. DNA- and RNA-based phylogenetic analyses targeting 16S rRNA indicated the successful growth of phylogenetically diverse microbial components during cultivation in the reactor. Most of the phylotypes in the reactor, once it made methane, were more closely related to culture sequences than to the subsurface environmental sequence. Potentially methanogenic phylotypes related to the genera Methanobacterium, Methanococcoides and Methanosarcina were predominantly detected concomitantly with methane production, while uncultured archaeal phylotypes were also detected. Using the methanogenic community enrichment as subsequent inocula, traditional batch-type cultivations led to the successful isolation of several anaerobic microbes including those methanogens. Our results substantiate that the DHS bioreactor is a useful system for the enrichment of numerous fastidious microbes from subseafloor sediments and will enable the physiological and ecological characterization of pure cultures of previously uncultivated subseafloor microbial life.

Level monitoring system with pulsating sensor—Application to online level monitoring of dashpots in a fast breeder reactor

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

Malathi, N.; Sahoo, P., E-mail: sahoop@igcar.gov.in; Ananthanarayanan, R.

2015-02-15

An innovative continuous type liquid level monitoring system constructed by using a new class of sensor, viz., pulsating sensor, is presented. This device is of industrial grade and it is exclusively used for level monitoring of any non conducting liquid. This instrument of unique design is suitable for high resolution online monitoring of oil level in dashpots of a sodium-cooled fast breeder reactor. The sensing probe is of capacitance type robust probe consisting of a number of rectangular mirror polished stainless steel (SS-304) plates separated with uniform gaps. The performance of this novel instrument has been thoroughly investigated. The precision,more » sensitivity, response time, and the lowest detection limit in measurement using this device are <0.01 mm, ∼100 Hz/mm, ∼1 s, and ∼0.03 mm, respectively. The influence of temperature on liquid level is studied and the temperature compensation is provided in the instrument. The instrument qualified all recommended tests, such as environmental, electromagnetic interference and electromagnetic compatibility, and seismic tests prior to its deployment in nuclear reactor. With the evolution of this level measurement approach, it is possible to provide dashpot oil level sensors in fast breeder reactor for the first time for continuous measurement of oil level in dashpots of Control and Safety Rod Drive Mechanism during reactor operation.« less

Flowsheets and source terms for radioactive waste projections

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

Forsberg, C.W.

1985-03-01

Flowsheets and source terms used to generate radioactive waste projections in the Integrated Data Base (IDB) Program are given. Volumes of each waste type generated per unit product throughput have been determined for the following facilities: uranium mining, UF/sub 6/ conversion, uranium enrichment, fuel fabrication, boiling-water reactors (BWRs), pressurized-water reactors (PWRs), and fuel reprocessing. Source terms for DOE/defense wastes have been developed. Expected wastes from typical decommissioning operations for each facility type have been determined. All wastes are also characterized by isotopic composition at time of generation and by general chemical composition. 70 references, 21 figures, 53 tables.

Irradiation tests of ITER candidate Hall sensors using two types of neutron spectra.

PubMed

Ďuran, I; Bolshakova, I; Viererbl, L; Sentkerestiová, J; Holyaka, R; Lahodová, Z; Bém, P

2010-10-01

We report on irradiation tests of InSb based Hall sensors at two irradiation facilities with two distinct types of neutron spectra. One was a fission reactor neutron spectrum with a significant presence of thermal neutrons, while another one was purely fast neutron field. Total neutron fluence of the order of 10(16) cm(-2) was accumulated in both cases, leading to significant drop of Hall sensor sensitivity in case of fission reactor spectrum, while stable performance was observed at purely fast neutron spectrum. This finding suggests that performance of this particular type of Hall sensors is governed dominantly by transmutation. Additionally, it further stresses the need to test ITER candidate Hall sensors under neutron flux with ITER relevant spectrum.

Assessment of Nuclear Fuels using Radiographic Thickness Measurement Method

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

Muhammad Abir; Fahima Islam; Hyoung Koo Lee

2014-11-01

The Convert branch of the National Nuclear Security Administration (NNSA) Global Threat Reduction Initiative (GTRI) focuses on the development of high uranium density fuels for research and test reactors for nonproliferation. This fuel is aimed to convert low density high enriched uranium (HEU) based fuel to high density low enriched uranium (LEU) based fuel for high performance research reactors (HPRR). There are five U.S. reactors that fall under the HPRR category, including: the Massachusetts Institute of Technology Reactor (MITR), the National Bureau of Standards Reactor (NBSR), the Missouri University Research Reactor (UMRR), the Advanced Test Reactor (ATR), and the Highmore » Flux Isotope Reactor (HFIR). U-Mo alloy fuel phase in the form of either monolithic or dispersion foil type fuels, such as ATR Full-size In center flux trap Position (AFIP) and Reduced Enrichment for Research and Test Reactor (RERTR), are being designed for this purpose. The fabrication process1 of RERTR is susceptible to introducing a variety of fuel defects. A dependable quality control method is required during fabrication of RERTR miniplates to maintain the allowable design tolerances, therefore evaluating and analytically verifying the fabricated miniplates for maintaining quality standards as well as safety. The purpose of this work is to analyze the thickness of the fabricated RERTR-12 miniplates using non-destructive technique to meet the fuel plate specification for RERTR fuel to be used in the ATR.« less

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

Analysis on Reactor Criticality Condition and Fuel Conversion Capability Based on Different Loaded Plutonium Composition in FBR Core

NASA Astrophysics Data System (ADS)

Permana, Sidik; Saputra, Geby; Suzuki, Mitsutoshi; Saito, Masaki

2017-01-01

Reactor criticality condition and fuel conversion capability are depending on the fuel arrangement schemes, reactor core geometry and fuel burnup process as well as the effect of different fuel cycle and fuel composition. Criticality condition of reactor core and breeding ratio capability have been investigated in this present study based on fast breeder reactor (FBR) type for different loaded fuel compositions of plutonium in the fuel core regions. Loaded fuel of Plutonium compositions are based on spent nuclear fuel (SNF) of light water reactor (LWR) for different fuel burnup process and cooling time conditions of the reactors. Obtained results show that different initial fuels of plutonium gives a significant chance in criticality conditions and fuel conversion capability. Loaded plutonium based on higher burnup process gives a reduction value of criticality condition or less excess reactivity. It also obtains more fuel breeding ratio capability or more breeding gain. Some loaded plutonium based on longer cooling time of LWR gives less excess reactivity and in the same time, it gives higher breeding ratio capability of the reactors. More composition of even mass plutonium isotopes gives more absorption neutron which affects to decresing criticality or less excess reactivity in the core. Similar condition that more absorption neutron by fertile material or even mass plutonium will produce more fissile material or odd mass plutonium isotopes to increase the breeding gain of the reactor.

Phenomena Important in Molten Salt Reactor Simulations

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

Diamond, David J.; Brown, Nicholas R.; Denning, Richard

The U.S. Nuclear Regulatory Commission (NRC) is preparing for the future licensing of advanced reactors that will be very different from current light water reactors. Part of the NRC preparation strategy is to identify the simulation tools that will be used for confirmatory safety analysis of normal operation and abnormal situations in those reactors. This report advances that strategy for reactors that will use molten salts (MSRs). This includes reactors with the fuel within the salt as well as reactors using solid fuel. Although both types are discussed in this report, the emphasis is on those reactors with liquid fuelmore » because of the perception that solid-fuel MSRs will be significantly easier to simulate. These liquid-fuel reactors include thermal and fast neutron spectrum alternatives. The specific designs discussed in the report are a subset of many designs being considered in the U.S. and elsewhere but they are considered the most likely to submit information to the NRC in the near future. The objective herein, is to understand the design of proposed molten salt reactors, how they will operate under normal or transient/accident conditions, and what will be the corresponding modeling needs of simulation tools that consider neutronics, heat transfer, fluid dynamics, and material composition changes in the molten salt. These tools will enable the NRC to eventually carry out confirmatory analyses that examine the validity and accuracy of applicant’s calculations and help determine the margin of safety in plant design.« less

Nuclear fuel requirements for the American economy - A model

NASA Astrophysics Data System (ADS)

Curtis, Thomas Dexter

A model is provided to determine the amounts of various fuel streams required to supply energy from planned and projected nuclear plant operations, including new builds. Flexible, user-defined scenarios can be constructed with respect to energy requirements, choices of reactors and choices of fuels. The model includes interactive effects and extends through 2099. Outputs include energy provided by reactors, the number of reactors, and masses of natural Uranium and other fuels used. Energy demand, including electricity and hydrogen, is obtained from US DOE historical data and projections, along with other studies of potential hydrogen demand. An option to include other energy demand to nuclear power is included. Reactor types modeled include (thermal reactors) PWRs, BWRs and MHRs and (fast reactors) GFRs and SFRs. The MHRs (VHTRs), GFRs and SFRs are similar to those described in the 2002 DOE "Roadmap for Generation IV Nuclear Energy Systems." Fuel source choices include natural Uranium, self-recycled spent fuel, Plutonium from breeder reactors and existing stockpiles of surplus HEU, military Plutonium, LWR spent fuel and depleted Uranium. Other reactors and fuel sources can be added to the model. Fidelity checks of the model's results indicate good agreement with historical Uranium use and number of reactors, and with DOE projections. The model supports conclusions that substantial use of natural Uranium will likely continue to the end of the 21st century, though legacy spent fuel and depleted uranium could easily supply all nuclear energy demand by shifting to predominant use of fast reactors.

APPARATUS FOR CONTROL OF A BOILING REACTOR RESPONSIVE TO STEAM DEMAND

DOEpatents

Treshow, M.

1963-07-23

A method of controlling a fuel-rod-in-tube-type boilingwater reactor having nozzles at the point of water entry into the tube is described. Water is pumped into the nozzles by an auxiliary pump operated by steam from an interstage position of the associated turbine, so that the pumping speed is responsive to turbine demand. (AEC)

Removal properties of diesel exhaust particles by a dielectric barrier discharge reactor.

PubMed

Suzuki, Ken-ichiro; Takeuchi, Naomi; Madokoro, Kazuhiko; Fushimi, Chihiro; Yao, Shuiliang; Fujioka, Yuichi; Nihei, Yoshimasa

2008-02-01

The removal properties of diesel exhaust particles (DEP) were investigated using an engine exhaust particle size spectrometer (EEPS), field emission-type scanning electron microscopy (FE-SEM) and time-of-flight secondary ion mass spectrometry (TOF-SIMS). DEP were treated using a dielectric barrier discharge (DBD) reactor installed in the tail pipe of a diesel engine, and a model DBD reactor fed with DEP in the mixture of N(2) and O(2). When changing the experimental parameters of both the plasma conditions and the engine load conditions, we obtained characteristic information of DEP treated with plasma discharges from the particle diameter and the composition. In evaluating the model DBD reactor, it became clear that there were two types of plasma processes (reactions with active oxygen species to yield CO(2) and reactions with active nitrogen species to yield nitrogen containing compounds). Moreover, from the result of a TOF-SIMS analysis, the characteristic secondary ions, such as C(2)H(6)N(+), C(4)H(12)N(+), and C(10)H(20)N(2)(+), were strongly detected from the DEP surfaces during the plasma discharges. This indicates that the nitrogen contained hydrocarbons were generated by plasma reactions.

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

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

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

1991-09-01

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

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

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

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

Study Neutronic of Small Pb-Bi Cooled Non-Refuelling Nuclear Power Plant Reactor (SPINNOR) with Hexagonal Geometry Calculation

NASA Astrophysics Data System (ADS)

Nur Krisna, Dwita; Su'ud, Zaki

2017-01-01

Nuclear reactor technology is growing rapidly, especially in developing Nuclear Power Plant (NPP). The utilization of nuclear energy in power generation systems has been progressing phase of the first generation to the fourth generation. This final project paper discusses the analysis neutronic one-cooled fast reactor type Pb-Bi, which is capable of operating up to 20 years without refueling. This reactor uses Thorium Uranium Nitride as fuel and operating on power range 100-500MWtNPPs. The method of calculation used a computer simulation program utilizing the SRAC. SPINNOR reactor is designed with the geometry of hexagonal shaped terrace that radially divided into three regions, namely the outermost regions with highest percentage of fuel, the middle regions with medium percentage of fuel, and most in the area with the lowest percentage. SPINNOR fast reactor operated for 20 years with variations in the percentage of Uranium-233 by 7%, 7.75%, and 8.5%. The neutronic calculation and analysis show that the design can be optimized in a fast reactor for thermal power output SPINNOR 300MWt with a fuel fraction 60% and variations of Uranium-233 enrichment of 7%-8.5%.

Operating manual for the Bulk Shielding Reactor

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

Not Available

1983-04-01

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

Operating manual for the Bulk Shielding Reactor

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

Not Available

1987-03-01

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

MEANS FOR SHIELDING REACTORS

DOEpatents

Garrison, W.M.; McClinton, L.T.; Burton, M.

1959-03-10

A reactor of the heterageneous, heavy water moderated type is described. The reactor is comprised of a plurality of vertically disposed fuel element tubes extending through a tank of heavy water moderator and adapted to accommodate a flow of coolant water in contact with the fuel elements. A tank containing outgoing coolant water is disposed above the core to function is a radiation shield. Unsaturated liquid hydrocarbon is floated on top of the water in the shield tank to reduce to a minimum the possibility of the occurrence of explosive gaseous mixtures resulting from the neutron bombardment of the water in the shield tank.

Satellite nuclear power station: An engineering analysis

NASA Technical Reports Server (NTRS)

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

1973-01-01

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

NEUTRONIC REACTOR SHIELD AND SPACER CONSTRUCTION

DOEpatents

Wigner, E.P.; Ohlinger, L.A.

1958-11-18

Reactors of the heterogeneous, graphite moderated, fluid cooled type and shielding and spacing plugs for the coolant channels thereof are reported. In this design, the coolant passages extend horizontally through the moderator structure, accommodating the fuel elements in abutting end-to-end relationship, and have access openings through the outer shield at one face of the reactor to facilitate loading of the fuel elements. In the outer ends of the channels which extend through the shields are provided spacers and shielding plugs designed to offer minimal reslstance to coolant fluid flow while preventing emanation of harmful radiation through the access openings when closed between loadings.

Multiscale Aspects of Modeling Gas-Phase Nanoparticle Synthesis

PubMed Central

Buesser, B.; Gröhn, A.J.

2013-01-01

Aerosol reactors are utilized to manufacture nanoparticles in industrially relevant quantities. The development, understanding and scale-up of aerosol reactors can be facilitated with models and computer simulations. This review aims to provide an overview of recent developments of models and simulations and discuss their interconnection in a multiscale approach. A short introduction of the various aerosol reactor types and gas-phase particle dynamics is presented as a background for the later discussion of the models and simulations. Models are presented with decreasing time and length scales in sections on continuum, mesoscale, molecular dynamics and quantum mechanics models. PMID:23729992

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

Haydary, J., E-mail: juma.haydary@stuba.sk; Susa, D.; Dudáš, J.

Highlights: ► Pyrolysis of aseptic packages was carried out in a laboratory flow reactor. ► Distribution of tetrapak into the product yields was obtained. ► Composition of the pyrolysis products was estimated. ► Secondary thermal and catalytic decomposition of tars was studied. ► Two types of catalysts (dolomite and red clay marked AFRC) were used. - Abstract: Pyrolysis of aseptic packages (tetrapak cartons) in a laboratory apparatus using a flow screw type reactor and a secondary catalytic reactor for tar cracking was studied. The pyrolysis experiments were realized at temperatures ranging from 650 °C to 850 °C aimed at maximizingmore » of the amount of the gas product and reducing its tar content. Distribution of tetrapak into the product yields at different conditions was obtained. The presence of H{sub 2}, CO, CH{sub 4}, CO{sub 2} and light hydrocarbons, HCx, in the gas product was observed. The Aluminum foil was easily separated from the solid product. The rest part of char was characterized by proximate and elemental analysis and calorimetric measurements. The total organic carbon in the tar product was estimated by elemental analysis of tars. Two types of catalysts (dolomite and red clay marked AFRC) were used for catalytic thermal tar decomposition. Three series of experiments (without catalyst in a secondary cracking reactor, with dolomite and with AFRC) at temperatures of 650, 700, 750, 800 and 850 °C were carried out. Both types of catalysts have significantly affected the content of tars and other components in pyrolytic gases. The effect of catalyst on the tetrapack distribution into the product yield on the composition of gas and on the total organic carbon in the tar product is presented in this work.« less

The use of plasma technology for the treatment of noxious waste

NASA Astrophysics Data System (ADS)

Wilman, Jonathan James

This thesis begins by describing the common types of air pollution and the main effects of these pollutants. Natural and man-made sources are discussed as well as the current types of technology used for reduction of common pollutants. The use of atmospheric pressure non-thermal plasma reactors for the control of pollutants is introduced at this stage. The second chapter describes the different types of atmospheric pressure non-thermal reactor designs and their modes of operation. The fundamental processes behind the production of plasmas are discussed and the chemistry of some simple discharges is also presented. The third chapter begins the experimental and modelling work done at Manchester on the destruction of volatile organic compounds (VOCs) using packed bed reactors and pulsed corona reactors. This chapter is concerned with the destruction of toluene and its behaviour as the oxygen content of the carrier gas, flowing through the reactor, is changed. Work using a pulsed corona reactor is also presented showing the destruction of toluene as a function of the applied specific energy. A model is constructed using mainly atmospheric reactions and the predictions are compared with experimental values. The fourth chapter discusses the destruction of dichloromethane (DCM) as a function of the oxygen content of the carrier gas. A model is constructed, again from mainly atmospheric reactions, and the predictions compared with the experimental data obtained. Methane is chosen as a molecule to study in the fifth chapter. A model is constructed and compared with experimental findings, showing that the chemistry of non-thermal plasmas can be effectively represented using neutral gas phase chemistry. Finally chapter six is concerned with the use of a large scale pulsed corona system for the reduction of NO[x] in industrial flue gas. This system has been tested on a modem incinerator, showing encouraging results. The workings of a modem incinerator are described together with those of the corona facility and any instruments used in these tests. Some experimental results are discussed. The aim of this chapter is to show that plasma reactors can be scaled up for industrial use. This section also discusses the difficulty of analysing and working with industrial gases and large scale apparatus as opposed to laboratory scale experiments.

Interim waste storage for the Integral Fast Reactor

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

Benedict, R.W.; Phipps, R.D.; Condiff, D.W.

1991-01-01

The Integral Fast Reactor (IFR), which Argonne National Laboratory is developing, is an innovative liquid metal breeder reactor that uses metallic fuel and has a close coupled fuel recovery process. A pyrochemical process is used to separate the fission products from the actinide elements. These actinides are used to make new fuel for the reactor. As part of the overall IFR development program, Argonne has refurbished an existing Fuel Cycle Facility at ANL-West and is installing new equipment to demonstrate the remote reprocessing and fabrication of fuel for the Experimental Breeder Reactor II (EBR-II). During this demonstration the wastes thatmore » are produced will be treated and packaged to produce waste forms that would be typical of future commercial operations. These future waste forms would, assuming Argonne development goals are fulfilled, be essentially free of long half-life transuranic isotopes. Promising early results indicate that actinide extraction processes can be developed to strip these isotopes from waste stream and return them to the IFR type reactors for fissioning. 1 fig.« less

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.

Alternative nuclear technologies

NASA Astrophysics Data System (ADS)

Schubert, E.

1981-10-01

The lead times required to develop a select group of nuclear fission reactor types and fuel cycles to the point of readiness for full commercialization are compared. Along with lead times, fuel material requirements and comparative costs of producing electric power were estimated. A conservative approach and consistent criteria for all systems were used in estimates of the steps required and the times involved in developing each technology. The impact of the inevitable exhaustion of the low- or reasonable-cost uranium reserves in the United States on the desirability of completing the breeder reactor program, with its favorable long-term result on fission fuel supplies, is discussed. The long times projected to bring the most advanced alternative converter reactor technologies the heavy water reactor and the high-temperature gas-cooled reactor into commercial deployment when compared to the time projected to bring the breeder reactor into equivalent status suggest that the country's best choice is to develop the breeder. The perceived diversion-proliferation problems with the uranium plutonium fuel cycle have workable solutions that can be developed which will enable the use of those materials at substantially reduced levels of diversion risk.

Analysis of fluid fuel flow to the neutron kinetics on molten salt reactor FUJI-12

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

Aji, Indarta Kuncoro, E-mail: indartaaji@s.itb.ac.id; Waris, Abdul, E-mail: awaris@fi.itb.ac.id; Permana, Sidik

Molten Salt Reactor is a reactor are operating with molten salt fuel flowing. This condition interpret that the neutron kinetics of this reactor is affected by the flow rate of the fuel. This research analyze effect by the alteration velocity of the fuel by MSR type Fuji-12, with fuel composition LiF-BeF{sub 2}-ThF{sub 4}-{sup 233}UF{sub 4} respectively 71.78%-16%-11.86%-0.36%. Calculation process in this study is performed numerically by SOR and finite difference method use C programming language. Data of reactivity, neutron flux, and the macroscopic fission cross section for calculation process obtain from SRAC-CITATION (Standard thermal Reactor Analysis Code) and JENDL-4.0 datamore » library. SRAC system designed and developed by JAEA (Japan Atomic Energy Agency). This study aims to observe the effect of the velocity of fuel salt to the power generated from neutron precursors at fourth year of reactor operate (last critical condition) with number of multiplication effective; 1.0155.« less

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.

Emissivity of Candidate Materials for VHTR Applicationbs: Role of Oxidation and Surface Modification Treatments

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

Sridharan, Kumar; Allen, Todd; Anderson, Mark

The Generation IV (GEN IV) Nuclear Energy Systems Initiative was instituted by the Department of Energy (DOE) with the goal of researching and developing technologies and materials necessary for various types of future reactors. These GEN IV reactors will employ advanced fuel cycles, passive safety systems, and other innovative systems, leading to significant differences between these future reactors and current water-cooled reactors. The leading candidate for the Next Generation Nuclear Plant (NGNP) to be built at Idaho National Lab (INL) in the United States is the Very High Temperature Reactor (VHTR). Due to the high operating temperatures of the VHTR,more » the Reactor Pressure Vessel (RPV) will partially rely on heat transfer by radiation for cooling. Heat expulsion by radiation will become all the more important during high temperature excursions during off-normal accident scenarios. Radiant power is dictated by emissivity, a material property. The NGNP Materials Research and Development Program Plan [1] has identified emissivity and the effects of high temperature oxide formation on emissivity as an area of research towards the development of the VHTR.« less

Modeling and simulation of enzymatic gluconic acid production using immobilized enzyme and CSTR-PFTR circulation reaction system.

PubMed

Li, Can; Lin, Jianqun; Gao, Ling; Lin, Huibin; Lin, Jianqiang

2018-04-01

Production of gluconic acid by using immobilized enzyme and continuous stirred tank reactor-plug flow tubular reactor (CSTR-PFTR) circulation reaction system. A production system is constructed for gluconic acid production, which consists of a continuous stirred tank reactor (CSTR) for pH control and liquid storage and a plug flow tubular reactor (PFTR) filled with immobilized glucose oxidase (GOD) for gluconic acid production. Mathematical model is developed for this production system and simulation is made for the enzymatic reaction process. The pH inhibition effect on GOD is modeled by using a bell-type curve. Gluconic acid can be efficiently produced by using the reaction system and the mathematical model developed for this system can simulate and predict the process well.

Rotating Fluidized Bed Reactor for Space Nuclear Propulsion. Annual Report; Design Studies and Experimental Results

NASA Technical Reports Server (NTRS)

1971-01-01

The rotating fluidized bed reactor concept is being investigated for possible application in nuclear propulsion systems. Physics calculations show U-233 to be superior to U-235 as a fuel for a cavity reactor of this type. Preliminary estimates of the effect of hydrogen in the reactor, reflector material, and power peaking are given. A preliminary engineering analysis was made for U-235 and U-233 fueled systems. An evaluation of the parameters affecting the design of the system is given, along with the thrust-to-weight ratios. The experimental equipment is described, as are the special photographic techniques and procedures. Characteristics of the fluidized bed and experimental results are given, including photographic evidence of bed fluidization at high rotational velocities.

NUCLEAR REACTOR

DOEpatents

Christy, R.F.

1958-07-15

A nuclear reactor of the homogeneous liquid fuel type is described wherein the fissionable isotope is suspended or dissolved in a liquid moderator such as water. The reactor core is comprised essentially of a spherical vessel for containing the reactive composition surrounded by a reflector, preferably of beryllium oxide. The reactive composition may be an ordinary water solution of a soluble salt of uranium, the quantity of fissionable isotope in solution being sufficient to provide a critical mass in the vessel. The liquid fuel is stored in a tank of non-crtttcal geometry below the reactor vessel and outside of the reflector and is passed from the tank to the vessel through a pipe connecting the two by air pressure means. Neutron absorbing control and safety rods are operated within slots in the reflector adjacent to the vessel.

Design of virtual SCADA simulation system for pressurized water reactor

NASA Astrophysics Data System (ADS)

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

2016-02-01

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

Systems and methods for managing shared-path instrumentation and irradiation targets in a nuclear reactor

DOEpatents

Heinold, Mark R.; Berger, John F.; Loper, Milton H.; Runkle, Gary A.

2015-12-29

Systems and methods permit discriminate access to nuclear reactors. Systems provide penetration pathways to irradiation target loading and offloading systems, instrumentation systems, and other external systems at desired times, while limiting such access during undesired times. Systems use selection mechanisms that can be strategically positioned for space sharing to connect only desired systems to a reactor. Selection mechanisms include distinct paths, forks, diverters, turntables, and other types of selectors. Management methods with such systems permits use of the nuclear reactor and penetration pathways between different systems and functions, simultaneously and at only distinct desired times. Existing TIP drives and other known instrumentation and plant systems are useable with access management systems and methods, which can be used in any nuclear plant with access restrictions.

Designing a SCADA system simulator for fast breeder reactor

NASA Astrophysics Data System (ADS)

Nugraha, E.; Abdullah, A. G.; Hakim, D. L.

2016-04-01

SCADA (Supervisory Control and Data Acquisition) system simulator is a Human Machine Interface-based software that is able to visualize the process of a plant. This study describes the results of the process of designing a SCADA system simulator that aims to facilitate the operator in monitoring, controlling, handling the alarm, accessing historical data and historical trend in Nuclear Power Plant (NPP) type Fast Breeder Reactor (FBR). This research used simulation to simulate NPP type FBR Kalpakkam in India. This simulator was developed using Wonderware Intouch software 10 and is equipped with main menu, plant overview, area graphics, control display, set point display, alarm system, real-time trending, historical trending and security system. This simulator can properly simulate the principle of energy flow and energy conversion process on NPP type FBR. This SCADA system simulator can be used as training media for NPP type FBR prospective operators.

Simulation of an integrated system for the production of methane and single cell protein from biomass

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

Thomas, M.V.

1989-01-01

A numerical model was developed to simulate the operation of an integrated system for the production of methane and single-cell algal protein from a variety of biomass energy crops or waste streams. Economic analysis was performed at the end of each simulation. The model was capable of assisting in the determination of design parameters by providing relative economic information for various strategies. Three configurations of anaerobic reactors were simulated. These included fed-bed reactors, conventional stirred tank reactors, and continuously expanding reactors. A generic anaerobic digestion process model, using lumped substrate parameters, was developed for use by type-specific reactor models. Themore » generic anaerobic digestion model provided a tool for the testing of conversion efficiencies and kinetic parameters for a wide range of substrate types and reactor designs. Dynamic growth models were used to model the growth of algae and Eichornia crassipes was modeled as a function of daily incident radiation and temperature. The growth of Eichornia crassipes was modeled for the production of biomass as a substrate for digestion. Computer simulations with the system model indicated that tropical or subtropical locations offered the most promise for a viable system. The availability of large quantities of digestible waste and low land prices were found to be desirable in order to take advantage of the economies of scale. Other simulations indicated that poultry and swine manure produced larger biogas yields than cattle manure. The model was created in a modular fashion to allow for testing of a wide variety of unit operations. Coding was performed in the Pascal language for use on personal computers.« less

SPERT I DESTRUCTIVE TEST PROGRAM SAFETY ANALYSIS REPORT

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

Spano, A.H.; Miller, R.W.

1962-06-15

The water-moderated core used for destructive experiments is mounted in the Spent I open-type reactor vessel, which has no provision for pressurization or forced coolant flow. The core is an array of highly enriched aluminum clad, plate-type fuel assemblies, using four bladetype, gang-operated control rods. Reactor transients are initiated at ambient temperature by step-insentions of reactivity, using a control rod which can be quickly ejected from the core. Following an initial series of static measurements to determine the basic- reactor properties of the test core, a series of nondestructive, self-limiting power excursion tests was performed, which covered a reactor periodmore » range down to the point where minor fuel plate damage first occurred -approximately for a 10- msec period test. These tests provided power, temperature, and pressure data. Additional kinetic teste in the period region between 10 and 5 msec were completed to explore the region of limited core damage. Fuel plate damage results included plate distortion, cladding cracking, and fuel melting. These exploratory tests were valuable in revealing unexpected changes in the dependence of pressure, temperature, burst energy, and burst shape parameters on reactor period, although the dependence of peak power on reactor period was not significantly changed. An evaluation of hazards involved in conducting the 2- msec test, based on pessimistic assumptions regarding fission product release and weather conditions, indicates that with the procedural controls normally exercised in the conduct of any transient test at Spent and the special controls to be in effect during the destructive test series, no significant hazard to personnel or to the general public will be obtained. All nuclear operation is conducted remotely approximately 1/2 mile from the reactor building. Discussion is also given of the supervision and control of personnel during and after each destructive test, and of the plans for re-entry, cleanup, and restoration of the facility. (auth)« less

Effect of mechanical disruption on the effectiveness of three reactors used for dilute acid pretreatment of corn stover Part 1: chemical and physical substrate analysis

PubMed Central

2014-01-01

Background There is considerable interest in the conversion of lignocellulosic biomass to liquid fuels to provide substitutes for fossil fuels. Pretreatments, conducted to reduce biomass recalcitrance, usually remove at least some of the hemicellulose and/or lignin in cell walls. The hypothesis that led to this research was that reactor type could have a profound effect on the properties of pretreated materials and impact subsequent cellulose hydrolysis. Results Corn stover was dilute-acid pretreated using commercially relevant reactor types (ZipperClave® (ZC), Steam Gun (SG) and Horizontal Screw (HS)) under the same nominal conditions. Samples produced in the SG and HS achieved much higher cellulose digestibilities (88% and 95%, respectively), compared to the ZC sample (68%). Characterization, by chemical, physical, spectroscopic and electron microscopy methods, was used to gain an understanding of the effects causing the digestibility differences. Chemical differences were small; however, particle size differences appeared significant. Sum-frequency generation vibrational spectra indicated larger inter-fibrillar spacing or randomization of cellulose microfibrils in the HS sample. Simons’ staining indicated increased cellulose accessibility for the SG and HS samples. Electron microscopy showed that the SG and HS samples were more porous and fibrillated because of mechanical grinding and explosive depressurization occurring with these two reactors. These structural changes most likely permitted increased cellulose accessibility to enzymes, enhancing saccharification. Conclusions Dilute-acid pretreatment of corn stover using three different reactors under the same nominal conditions gave samples with very different digestibilities, although chemical differences in the pretreated substrates were small. The results of the physical and chemical analyses of the samples indicate that the explosive depressurization and mechanical grinding with these reactors increased enzyme accessibility. Pretreatment reactors using physical force to disrupt cell walls increase the effectiveness of the pretreatment process. PMID:24713111

Conceptual design of a thermalhydraulic loop for multiple test geometries at supercritical conditions named Supercritical Phenomena Experimental Test Apparatus (SPETA)

NASA Astrophysics Data System (ADS)

Adenariwo, Adepoju

The efficiency of nuclear reactors can be improved by increasing the operating pressure of current nuclear reactors. Current CANDU-type nuclear reactors use heavy water as coolant at an outlet pressure of up to 11.5 MPa. Conceptual SuperCritical Water Reactors (SCWRs) will operate at a higher coolant outlet pressure of 25 MPa. Supercritical water technology has been used in advanced coal plants and its application proves promising to be employed in nuclear reactors. To better understand how supercritical water technology can be applied in nuclear power plants, supercritical water loops are used to study the heat transfer phenomena as it applies to CANDU-type reactors. A conceptual design of a loop known as the Supercritical Phenomena Experimental Apparatus (SPETA) has been done. This loop has been designed to fit in a 9 m by 2 m by 2.8 m enclosure that will be installed at the University of Ontario Institute of Technology Energy Research Laboratory. The loop include components to safely start up and shut down various test sections, produce a heat source to the test section, and to remove reject heat. It is expected that loop will be able to investigate the behaviour of supercritical water in various geometries including bare tubes, annulus tubes, and multi-element-type bundles. The experimental geometries are designed to match the fluid properties of Canadian SCWR fuel channel designs so that they are representative of a practical application of supercritical water technology in nuclear plants. This loop will investigate various test section orientations which are the horizontal, vertical, and inclined to investigate buoyancy effects. Frictional pressure drop effects and satisfactory methods of estimating hydraulic resistances in supercritical fluid shall also be estimated with the loop. Operating limits for SPETA have been established to be able to capture the important heat transfer phenomena at supercritical conditions. Heat balance and flow calculations have been done to appropriately size components in the loop. Sensitivity analysis has been done to find the optimum design for the loop.

Reactor Materials Program - Baseline Material Property Handbook - Mechanical Properties of 1950's Vintage Stainless Steel Weldment Components, Task Number 89-23-A-1

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

Stoner, K.J.

1999-11-05

The Process Water System (primary coolant) piping of the nuclear production reactors constructed in the 1950''s at Savannah River Site is comprised primarily of Type 304 stainless steel with Type 308 stainless steel weld filler. A program to measure the mechanical properties of archival PWS piping and weld materials (having approximately six years of service at temperatures between 25 and 100 degrees C) has been completed. The results from the mechanical testing has been synthesized to provide a mechanical properties database for structural analyses of the SRS piping.

FABRICATION OF TUBE TYPE FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Loeb, E.; Nicklas, J.H.

1959-02-01

A method of fabricating a nuclear reactor fuel element is given. It consists essentially of fixing two tubes in concentric relationship with respect to one another to provide an annulus therebetween, filling the annulus with a fissionablematerial-containing powder, compacting the powder material within the annulus and closing the ends thereof. The powder material is further compacted by swaging the inner surface of the inner tube to increase its diameter while maintaining the original size of the outer tube. This process results in reduced fabrication costs of powdered fissionable material type fuel elements and a substantial reduction in the peak core temperatures while materially enhancing the heat removal characteristics.

Issues of intergranular embrittlement of VVER-type nuclear reactors pressure vessel materials

NASA Astrophysics Data System (ADS)

Zabusov, O.

2016-04-01

In light of worldwide tendency to extension of service life of operating nuclear power plants - VVER-type in the first place - recently a special attention is concentrated on phenomena taking place in reactor pressure vessel materials that are able to lead to increased level of mechanical characteristics degradation (resistibility to brittle fracture) during long term of operation. Formerly the hardening mechanism of degradation (increase in the yield strength under influence of irradiation) mainly had been taken into consideration to assess pressure vessel service life limitations, but when extending the service life up to 60 years and more the non-hardening mechanism (intergranular embrittlement of the steels) must be taken into account as well. In this connection NRC “Kurchatov Institute” has initiated a number of works on investigations of this mechanism contribution to the total embrittlement of reactor pressure vessel steels. The main results of these investigations are described in this article. Results of grain boundary phosphorus concentration measurements in specimens made of first generation of VVER-type pressure vessels materials as well as VVER-1000 surveillance specimens are presented. An assessment of non-hardening mechanism contribution to the total ductile-to- brittle transition temperature shift is given.

Degradation of Acid Orange 7 Dye in Two Hybrid Plasma Discharge Reactors

NASA Astrophysics Data System (ADS)

Shen, Yongjun; Lei, Lecheng; Zhang, Xingwang; Ding, Jiandong

2014-11-01

To get an optimized pulsed electrical plasma discharge reactor and to increase the energy utilization efficiency in the removal of pollutants, two hybrid plasma discharge reactors were designed and optimized. The reactors were compared via the discharge characteristics, energy transfer efficiency, the yields of the active species and the energy utilization in dye wastewater degradation. The results showed that under the same AC input power, the characteristics of the discharge waveform of the point-to-plate reactor were better. Under the same AC input power, the two reactors both had almost the same peak voltage of 22 kV. The peak current of the point-to-plate reactor was 146 A, while that of the wire-to-cylinder reactor was only 48.8 A. The peak powers of the point-to-plate reactor and the wire-to-cylinder reactor were 1.38 MW and 1.01 MW, respectively. The energy per pulse of the point-to-plate reactor was 0.2221 J, which was about 29.4% higher than that of the wire-to-cylinder reactor (0.1716 J). To remove 50% Acid Orange 7 (AO7), the energy utilizations of the point-to-plate reactor and the wire-to-cylinder reactor were 1.02 × 10-9 mol/L and 0.61 × 10-9 mol/L, respectively. In the point-to-plate reactor, the concentration of hydrogen peroxide in pure water was 3.6 mmol/L after 40 min of discharge, which was higher than that of the wire-to-cylinder reactor (2.5 mmol/L). The concentration of liquid phase ozone in the point-to-plate reactor (5.7 × 10-2 mmol/L) was about 26.7% higher than that in the wire-to-cylinder reactor (4.5 × 10-2 mmol/L). The analysis results of the variance showed that the type of reactor and reaction time had significant impacts on the yields of the hydrogen peroxide and ozone. The main degradation intermediates of AO7 identified by gas chromatography and mass spectrometry (GCMS) were acetic acid, maleic anhydride, p-benzoquinone, phenol, benzoic acid, phthalic anhydride, coumarin and 2-naphthol. Proposed degradation pathways were elucidated in light of the analyzed degradation products.

Steady-State Thermal-Hydraulics Analyses for the Conversion of the BR2 Reactor to LEU

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

Licht, J. R.; Bergeron, A.; Dionne, B.

BR2 is a research reactor used for radioisotope production and materials testing. It’s a tank-in-pool type reactor cooled by light water and moderated by beryllium and light water. The reactor core consists of a beryllium moderator forming a matrix of 79 hexagonal prisms in a hyperboloid configuration; each having a central bore that can contain a variety of different components such as a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Based on a series of tests, the BR2 operation is currently limited to a maximum allowable heat flux of 470 W/cmmore » 2 to ensure fuel plate integrity during steady-state operation and after a loss-of-flow/loss-of-pressure accident. A feasibility study for the conversion of the BR2 reactor from highly-enriched uranium (HEU) to low-enriched uranium (LEU) fuel was previously performed to verify it can operate safely at the same maximum nominal steady-state heat flux. An assessment was also performed to quantify the heat fluxes at which the onset of flow instability and critical heat flux occur for each fuel type. This document updates and expands these results for the current representative core configuration (assuming a fresh beryllium matrix) by evaluating the onset of nucleate boiling (ONB), onset of fully developed nucleate boiling (FDNB), onset of flow instability (OFI) and critical heat flux (CHF).« less

Development of a Reactor Model for Chemical Conversion of Lunar Regolith

NASA Technical Reports Server (NTRS)

Hegde, U.; Balasubramaniam, R.; Gokoglu, S.

2009-01-01

Lunar regolith will be used for a variety of purposes such as oxygen and propellant production and manufacture of various materials. The design and development of chemical conversion reactors for processing lunar regolith will require an understanding of the coupling among the chemical, mass and energy transport processes occurring at the length and time scales of the overall reactor with those occurring at the corresponding scales of the regolith particles. To this end, a coupled transport model is developed using, as an example, the reduction of ilmenite-containing regolith by a continuous flow of hydrogen in a flow-through reactor. The ilmenite conversion occurs on the surface and within the regolith particles. As the ilmenite reduction proceeds, the hydrogen in the reactor is consumed, and this, in turn, affects the conversion rate of the ilmenite in the particles. Several important quantities are identified as a result of the analysis. Reactor scale parameters include the void fraction (i.e., the fraction of the reactor volume not occupied by the regolith particles) and the residence time of hydrogen in the reactor. Particle scale quantities include the time for hydrogen to diffuse into the pores of the regolith particles and the chemical reaction time. The paper investigates the relationships between these quantities and their impact on the regolith conversion. Application of the model to various chemical reactor types, such as fluidized-bed, packed-bed, and rotary-bed configurations, are discussed.

Development of a Reactor Model for Chemical Conversion of Lunar Regolith

NASA Technical Reports Server (NTRS)

Hedge, uday; Balasubramaniam, R.; Gokoglu, S.

2007-01-01

Lunar regolith will be used for a variety of purposes such as oxygen and propellant production and manufacture of various materials. The design and development of chemical conversion reactors for processing lunar regolith will require an understanding of the coupling among the chemical, mass and energy transport processes occurring at the length and time scales of the overall reactor with those occurring at the corresponding scales of the regolith particles. To this end, a coupled transport model is developed using, as an example, the reduction of ilmenite-containing regolith by a continuous flow of hydrogen in a flow-through reactor. The ilmenite conversion occurs on the surface and within the regolith particles. As the ilmenite reduction proceeds, the hydrogen in the reactor is consumed, and this, in turn, affects the conversion rate of the ilmenite in the particles. Several important quantities are identified as a result of the analysis. Reactor scale parameters include the void fraction (i.e., the fraction of the reactor volume not occupied by the regolith particles) and the residence time of hydrogen in the reactor. Particle scale quantities include the time for hydrogen to diffuse into the pores of the regolith particles and the chemical reaction time. The paper investigates the relationships between these quantities and their impact on the regolith conversion. Application of the model to various chemical reactor types, such as fluidized-bed, packed-bed, and rotary-bed configurations, are discussed.

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.

Axially staggered seed-blanket reactor fuel module construction

DOEpatents

Cowell, Gary K.; DiGuiseppe, Carl P.

1985-01-01

A heterogeneous nuclear reactor of the seed-blanket type is provided wher the fissile (seed) and fertile (blanket) nuclear fuels are segregated axially within each fuel element such that fissile and fertile regions occur in an alternating pattern along the length of the fuel element. Further, different axial stacking patterns are used for the fuel elements of at least two module types such that when modules of different types are positioned adjacent to one another, the fertile regions of the modules are offset or staggered. Thus, when a module of one type is surrounded by modules of the second type the fertile regions thereof will be surrounded on all sides by fissile material. This provides enhanced neutron communication both radially and axially, thereby resulting in greater power oscillation stability than other axial arrangements. The arrangements of the fissile and fertile regions in an alternating axial manner minimizes the radial power peaking factors and provides a more optional thermal-hydraulic design than is afforded by radial arrangements.

Evaluation of performance with small and scale-up rotating and flat reactors; photocatalytic degradation of bisphenol A, 17β-estradiol, and 17α-ethynyl estradiol under solar irradiation.

PubMed

Kim, Saewon; Cho, Hyekyung; Joo, Hyunku; Her, Namguk; Han, Jonghun; Yi, Kwangbok; Kim, Jong-Oh; Yoon, Jaekyung

2017-08-15

In this study, the performances of photocatalytic reactors of the small and scale-up rotating and flat types were evaluated to investigate the treatment of new emerging contaminants such as bisphenol A (BPA), 17α-ethynyl estradiol (EE2), and 17β-estradiol (E2) that are known as endocrine disrupting compounds (EDCs). In the laboratory tests with the small-scale rotating and flat reactors, the degradation efficiencies of the mixed EDCs were significantly influenced by the change of the hydraulic retention time (HRT). In particular, considering the effective two-dimensional reaction area with light and nanotubular TiO 2 (NTT) on a Ti substrate, the rotating reactors showed the more effective performance than the flat reactor because the degradation efficiencies are similar in the small effective area. In addition, the major parameters affecting the photocatalytic activities of the NTT were evaluated for the rotating reactors according to the effects of single and mixed EDCs, the initial concentrations of the EDCs, the UV intensity, and dissolved oxygen. In the extended outdoor tests with the scale-up photocatalytic reactors and NTT, it was confirmed from the four representative demonstrations that an excellent rotating-reactor performance is consistently shown in terms of the degradation of the target pollutants under solar irradiation. Copyright © 2017 Elsevier B.V. All rights reserved.

Co-cultivation of Lactobacillus zeae and Veillonella cricetifor the production of propionic acid

PubMed Central

2013-01-01

In this work a defined co-culture of the lactic acid bacterium Lactobacillus zeae and the propionate producer Veillonella criceti has been studied in continuous stirred tank reactor (CSTR) and in a dialysis membrane reactor. It is the first time that this reactor type is used for a defined co-culture fermentation. This reactor allows high mixing rates and working with high cell densities, making it ideal for co-culture investigations. In CSTR experiments the co-culture showed over a broad concentration range an almost linear correlation in consumption and production rates to the supply with complex nutrients. In CSTR and dialysis cultures a strong growth stimulation of L. zeae by V. criceti was shown. In dialysis cultures very high propionate production rates (0.61 g L-1h-1) with final titers up to 28 g L-1 have been realized. This reactor allows an individual, intracellular investigation of the co-culture partners by omic-technologies to provide a better understanding of microbial communities. PMID:23705662

Studies of PuF sub 6 and transplutonic materials' critical properties for space high power nuclear pumped lasers

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

Gu, A.G.; Miller, M.S.

1991-01-01

All space missions require a reliable, compact source of energy. This paper describes preliminary neutronics studies of pocket'' reactor concepts employing PuF{sub 6} and transplutonic materials as fuels for space high power/energy Nuclear Pumped Lasers (NPLs). Previous research has studied NPL reactor concepts with thin fuel layers, aerosol fuels and gaseous UF{sub 6}. The total reactor volumes for compact reactors with these types of fuels typically range from 3 m{sup 3} to 50 m{sup 3}. By employing PuF{sub 6} and transplutonic fuels at the same low densities, a calculated value for Keff of 1.2 has been achieved for conditions ofmore » 900 K and 5 atm, with total reactor volumes of 1.5 m{sup 3} for PuF{sub 6}, 0.51 m{sup 3} for Am-242m, 0.58 m{sup 3} for Cm-245 and 0.63 m{sup 3} for Cf-249.« less

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

PubMed

Fontana, Roselei Claudete; da Silveira, Maurício Moura

2012-11-01

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

Use of Nitrogen Trifluoride To Purify Molten Salt Reactor Coolant and Heat Transfer Fluoride Salts

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

Scheele, Randall D.; Casella, Andrew M.; McNamara, Bruce K.

2017-05-02

Abstract: The molten salt cooled nuclear reactor is included as one of the Generation IV reactor types. One of the challenges with the implementation of this reactor is purifying and maintaining the purity of the various molten fluoride salts that will be used as coolants. The method used for Oak Ridge National Laboratory’s molten salt experimental test reactor was to treat the coolant with a mixture of H2 and HF at 600°C. In this article we evaluate thermal NF3 treatment for purifying molten fluoride salt coolant candidates based on NF3’s 1) past use to purify fluoride salts, 2) other industrialmore » uses, 3) commercial availability, 4) operational, chemical, and health hazards, 5) environmental effects and environmental risk management methods, 6) corrosive properties, and 7) thermodynamic potential to eliminate impurities that could arise due to exposure to water and oxygen. Our evaluation indicates that nitrogen trifluoride is a viable and safer alternative to the previous method.« less

Analysis of a bio-electrochemical reactor containing carbon fiber textiles for the anaerobic digestion of tomato plant residues.

PubMed

Hirano, Shin-Ichi; Matsumoto, Norio

2018-02-01

A bio-electrochemical system packed with supporting material can promote anaerobic digestion for several types of organic waste. To expand the target organic matters of a BES, tomato plant residues (TPRs), generated year-round as agricultural and cellulosic waste, were treated using three methanogenic reactors: a continuous stirred tank reactor (CSTR), a carbon fiber textile (CFT) reactor, and a bio-electrochemical reactor (BER) including CFT with electrochemical regulation (BER + CFT). CFT had positive effects on methane fermentation and methanogen abundance. The microbial population stimulated by electrochemical regulation, including hydrogenotrophic methanogens, cellulose-degrading bacteria, and acetate-degrading bacteria, suppressed acetate accumulation, as evidenced by the low acetate concentration in the suspended fraction in the BER + CFT. These results indicated that the microbial community in the BER + CFT facilitated the efficient decomposition of TPR and its intermediates such as acetate to methane. Copyright © 2017 Elsevier Ltd. All rights reserved.

Investigation of the effects of radiolytic-gas bubbles on the long-term operation of solution reactors for medical-isotope production

NASA Astrophysics Data System (ADS)

Souto Mantecon, Francisco Javier

One of the most common and important medical radioisotopes is 99Mo, which is currently produced using the target irradiation technology in heterogeneous nuclear reactors. The medical isotope 99Mo can also be produced from uranium fission using aqueous homogeneous solution reactors. In solution reactors, 99Mo is generated directly in the fuel solution, resulting in potential advantages when compared with the target irradiation process in heterogeneous reactors, such as lower reactor power, less waste heat, and reduction by a factor of about 100 in the generation of spent fuel. The commercial production of medical isotopes in solution reactors requires steady-state operation at about 200 kW. At this power regime, the formation of radiolytic-gas bubbles creates a void volume in the fuel solution that introduces a negative coefficient of reactivity, resulting in power reduction and instabilities that may impede reactor operation for medical-isotope production. A model has been developed considering that reactivity effects are due to the increase in the fuel-solution temperature and the formation of radiolytic-gas bubbles. The model has been validated against experimental results from the Los Alamos National Laboratory uranyl fluoride Solution High-Energy Burst Assembly (SHEBA), and the SILENE uranyl nitrate solution reactor, commissioned at the Commissariat a l'Energie Atomique, in Valduc, France. The model shows the feasibility of solution reactors for the commercial production of medical isotopes and reveals some of the important parameters to consider in their design, including the fuel-solution type, 235U enrichment, uranium concentration, reactor vessel geometry, and neutron reflectors surrounding the reactor vessel. The work presented herein indicates that steady-state operation at 200 kW can be achieved with a solution reactor consisting of 120 L of uranyl nitrate solution enriched up to 20% with 235U and a uranium concentration of 145 kg/m3 in a graphite-reflected cylindrical geometry.

Management of fresh water weeds (macrophytes) by vermicomposting using Eisenia fetida.

PubMed

Najar, Ishtiyaq Ahmed; Khan, Anisa B

2013-09-01

In the present study, potential of Eisenia fetida to recycle the different types of fresh water weeds (macrophytes) used as substrate in different reactors (Azolla pinnata reactor, Trapa natans reactor, Ceratophyllum demersum reactor, free-floating macrophytes mixture reactor, and submerged macrophytes mixture reactor) during 2 months experiment is investigated. E. fetida showed significant variation in number and weight among the reactors and during the different fortnights (P <0.05) with maximum in A. pinnata reactor (number 343.3 ± 10.23 %; weight 98.62 ± 4.23 % ) and minimum in submerged macrophytes mixture reactor (number 105 ± 5.77 %; weight 41.07 ± 3.97 % ). ANOVA showed significant variation in cocoon production (F4 = 15.67, P <0.05) and mean body weight (F4 = 13.49, P <0.05) among different reactors whereas growth rate (F3 = 23.62, P <0.05) and relative growth rate (F3 = 4.91, P <0.05) exhibited significant variation during different fortnights. Reactors showed significant variation (P <0.05) in pH, Electrical conductivity (EC), Organic carbon (OC), Organic nitrogen (ON), and C/N ratio during different fortnights with increase in pH, EC, N, and K whereas decrease in OC and C/N ratio. Hierarchical cluster analysis grouped five substrates (weeds) into three clusters-poor vermicompost substrates, moderate vermicompost substrate, and excellent vermicompost substrate. Two principal components (PCs) have been identified by factor analysis with a cumulative variance of 90.43 %. PC1 accounts for 47.17 % of the total variance represents "reproduction factor" and PC2 explaining 43.26 % variance representing "growth factor." Thus, the nature of macrophyte affects the growth and reproduction pattern of E. fetida among the different reactors, further the addition of A. pinnata in other macrophytes reactors can improve their recycling by E. fetida.

Postirradiation thermocyclic loading of ferritic-martensitic structural materials

NASA Astrophysics Data System (ADS)

Belyaeva, L.; Orychtchenko, A.; Petersen, C.; Rybin, V.

Thermonuclear fusion reactors of the Tokamak-type will be unique power engineering plants to operate in thermocyclic mode only. Ferritic-martensitic stainless steels are prime candidate structural materials for test blankets of the ITER fusion reactor. Beyond the radiation damage, thermomechanical cyclic loading is considered as the most detrimental lifetime limiting phenomenon for the above structure. With a Russian and a German facility for thermal fatigue testing of neutron irradiated materials a cooperation has been undertaken. Ampule devices to irradiate specimens for postirradiation thermal fatigue tests have been developed by the Russian partner. The irradiation of these ampule devices loaded with specimens of ferritic-martensitic steels, like the European MANET-II, the Russian 05K12N2M and the Japanese Low Activation Material F82H-mod, in a WWR-M-type reactor just started. A description of the irradiation facility, the qualification of the ampule device and the modification of the German thermal fatigue facility will be presented.

Proposal for a novel type of small scale aneutronic fusion reactor

NASA Astrophysics Data System (ADS)

Gruenwald, J.

2017-02-01

The aim of this work is to propose a novel scheme for a small scale aneutronic fusion reactor. This new reactor type makes use of the advantages of combining laser driven plasma acceleration and electrostatic confinement fusion. An intense laser beam is used to create a lithium-proton plasma with high density, which is then collimated and focused into the centre of the fusion reaction chamber. The basic concept presented here is based on the 7Li-proton fusion reaction. However, the physical and technological fundamentals may generally as well be applied to 11B-proton fusion. The former fusion reaction path offers higher energy yields while the latter has larger fusion cross sections. Within this paper a technological realisation of such a fusion device, which allows a steady state operation with highly energetic, well collimated ion beam, is presented. It will be demonstrated that the energetic break even can be reached with this device by using a combination of already existing technologies.