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1

Source-Term and building-Wake Consequence Modeling for the Godiva IV Reactor at Los Alamos National Laboratory  

SciTech Connect

The objectives of this work were to evaluate the consequences of a postulated accident to onsite security personnel stationed near the facility during operations of the Godiva IV critical assembly and to identify controls needed to protect these personnel in case of an extreme criticality excursion equivalent to the design-basis accident (DBA). This paper presents the methodology and results of the source-term calculations, building ventilation rates, air concentrations, and consequence calculations that were performed using a multidisciplinary approach with several phenomenology models. Identification of controls needed to mitigate the consequences to near-field receptors is discussed.

Letellier, B.C.; McClure, P.; Restrepo, L.

1999-06-13

2

Godiva and Juliet Diagnostics CED-1 (IER-176)  

SciTech Connect

A suite of diagnostics are being proposed for use in the Juliet experiment (IER-128). In order to calibrate and test the diagnostics prior to use, the LLNL calibration facility and Godiva pulsed reactor will be used to provide intense sources of neutrons and gammas. Due to the similarities of the Godiva and Juliet radiation fields, the diagnostics being developed and tested for Juliet can also play an on-going role in diagnostics for Godiva as well as, perhaps, other critical assembly experiments. Similar work is also being conducted for IER-147 for the purpose of characterizing the Godiva radiation field in support of an upcoming international nuclear accident dosimetry exercise. Diagnostics developed and fielded under IER-147 can provide valuable data with respect to the neutron and gamma energy spectrums in the vicinity of Godiva which is relevant to the calibration of Juliet diagnostics.

Scorby, J C

2011-12-21

3

Validation of energy moments from the one-dimensional energy dependent neutron diffusion equation, MCNP5 and Attila-7.1.0 with the GODIVA experiment  

SciTech Connect

Normalized neutron energy moments (moments) from the one-dimensional energy dependent neutron diffusion equation (EDNDE), Monte Carlo N Particle 5 version 1.40 (MCNP5) and Attila-7.1.0-beta version (Attila) are validated with the GODIVA experiment (GODIVA). Energy moments 0–5 for all three methods are compared to GODIVA moments. GODIVA moments are measured with two methods. The 1st method is a time of flight (T-O-F) measurement of the average energy (moment 1) of the leaking neutrons from the surface of GODIVA and the 2nd method is from back calculating moments from foil activation analysis of various metal foils at the center of GODIVA. The error range of the EDNDE normalized moments compared to GODIVA is from 0% to 24%. The MCNP5 error range compared to GODIVA is 0–12% and the Attila error range is 0–79%. The method of moments is shown to be a fast reliable method, compared to either Monte Carlo methods (MCNP5) or 30 multi-energy group methods (Attila) with regard to the GODIVA experiment.

Douglas S. Crawford; Terry A. Ring

2012-12-01

4

Analysis of Godiva-IV delayed-critical and static super-prompt-critical conditions  

SciTech Connect

Super-prompt-critical burst experiments were conducted on the Godiva-IV assembly at Los Alamos National Laboratory from the 1960s through 2005. Detailed and simplified benchmark models have been constructed for four delayed-critical experiments and for the static phase of a super-prompt-critical burst experiment. In addition, a two-dimensional cylindrical model has been developed for the super-prompt-critical condition. Criticality calculations have been performed for all of those models with four modern nuclear data libraries: ENDFIB-VI, ENDF/8-VII.0, JEFF-3.1 , and JENDL-3.3. Overall, JENDL-3.3 produces the best agreement with the reference values for k{sub eff}.

Mosteller, Russell D [Los Alamos National Laboratory; Goda, Joetta M [Los Alamos National Laboratory

2009-01-01

5

Godiva Rim Member: A new stratigraphic unit of the Green River Formation in southwest Wyoming and northwest Colorado. Geology of the Eocene Wasatch, Green River, and Bridger (Washakie) Formations, Greater Green River Basin, Wyoming, Utah, and Colorado. Professional paper  

SciTech Connect

The report names and describes the Godiva Rim Member of the Green River Formation in the eastern part of the Washakie basin in southwest Wyoming and the central part of the Sand Wash basin in northwest Colorado. The Godiva Rim Member comprises lithofacies of mixed mudflat and lacustrine origin situated between the overlying lacustrine Laney Member of the Green River Formation and the underlying fluvial Cathedral Bluffs Tongue of the Wasatch Formation. The Godiva Rim Member is laterally equivalent to and grades westward into the LaClede Bed of the Laney Member. The Godiva Rim Member of the Green River Formation was deposited along the southeast margins of Lake Gosiute and is correlated to similar lithologic units that were deposited along the northeast margins of Lake Uinta in the Parachute Creek Member of the Green River Formation. The stratigraphic data presented provide significant evidence that the two lakes were periodically connected around the east end of the Uinta Mountains during the middle Eocene.

Roehler, H.W.

1991-01-01

6

Reevaluation of an individual's radiation exposure at NTS in 1963-64. [FRAN reactor  

SciTech Connect

The FRAN prompt burst reactor began operation at NTS on November 1, 1962 and continued in use until April 1965. From January 2, 1963 to August 12, 1964, an individual periodically performed maintenance and troubleshooting functions on various components of the FRAN reactor system. In June, 1980, the individual requested a review of the radiation dose that he received from his involvement with the FRAN reactor. An evaluation of the individual's radiation dose associated with the FRAN reactor operation was performed. This report details the reevaluation of the individual's estimated radiation dose from the FRAN reactor assembly, as derived from computer calculations, GODIVA-IV measurements, personnel dosimetry results, and a reconstruction of work scenarios.

Myers, D.S.

1983-02-25

7

Feasibility study of noise analysis methods on virtual thermal reactor subcriticality monitoring  

SciTech Connect

This paper presents the analysis results of Rossi-alpha, cross-correlation, Feynman-alpha, and Feynman difference methods applied to the subcriticality monitoring of nuclear reactors. A thermal spectrum Godiva model has been designed for the analysis of the four methods. This Godiva geometry consists of a spherical core containing the isotopes of H-l, U-235 and U-238, and the H{sub 2}O reflector outside the core. A Monte Carlo code, McCARD, is used in real time mode to generate virtual detector signals to analyze the feasibility of the four methods. The analysis results indicate that the four methods can be used with high accuracy for the continuous monitoring of subcriticality. In addition to that, in order to analyze the impact of the random noise contamination on the accuracy of the noise analysis, the McCARD-generated signals are contaminated with arbitrary noise. It is noticed that, even when the detector signals are contaminated, the four methods can predict the subcriticality with reasonable accuracy. Nonetheless, in order to reduce the adverse impact of the random noise, eight detector signals, rather than a single signal, are generated from the core, one signal from each equally divided eighth part of the core. The preliminary analysis with multiple virtual detector signals indicates that the approach of using many detectors is promising to improve the accuracy of criticality prediction and further study will be performed in this regard. (authors)

Kong, C.; Lee, D. [Ulsan National Institute of Science and Technology UNIST-gil, 50, Eonyang-eup, Ulju-gun, Ulsan, 689-798 (Korea, Republic of)] [Ulsan National Institute of Science and Technology UNIST-gil, 50, Eonyang-eup, Ulju-gun, Ulsan, 689-798 (Korea, Republic of); Lee, E. [Korea Hydro and Nuclear Power Co., 1312-70, Yuseong-daero, Yuseong-gu, Daejeon, 305-343 (Korea, Republic of)] [Korea Hydro and Nuclear Power Co., 1312-70, Yuseong-daero, Yuseong-gu, Daejeon, 305-343 (Korea, Republic of)

2013-07-01

8

Verification of Unstructured Mesh Capabilities in MCNP6 for Reactor Physics Problems  

SciTech Connect

New unstructured mesh capabilities in MCNP6 (developmental version during summer 2012) show potential for conducting multi-physics analyses by coupling MCNP to a finite element solver such as Abaqus/CAE[2]. Before these new capabilities can be utilized, the ability of MCNP to accurately estimate eigenvalues and pin powers using an unstructured mesh must first be verified. Previous work to verify the unstructured mesh capabilities in MCNP was accomplished using the Godiva sphere [1], and this work attempts to build on that. To accomplish this, a criticality benchmark and a fuel assembly benchmark were used for calculations in MCNP using both the Constructive Solid Geometry (CSG) native to MCNP and the unstructured mesh geometry generated using Abaqus/CAE. The Big Ten criticality benchmark [3] was modeled due to its geometry being similar to that of a reactor fuel pin. The C5G7 3-D Mixed Oxide (MOX) Fuel Assembly Benchmark [4] was modeled to test the unstructured mesh capabilities on a reactor-type problem.

Burke, Timothy P. [Los Alamos National Laboratory; Martz, Roger L. [Los Alamos National Laboratory; Kiedrowski, Brian C. [Los Alamos National Laboratory; Martin, William R. [Los Alamos National Laboratory

2012-08-22

9

Thirty-five years at Pajarito Canyon Site  

SciTech Connect

A history of the research activities performed at the Pajarito Canyon Site from 1946 to 1981 is presented. Critical assemblies described include: the Topsy assembly; Lady Godiva; Godiva 2; Jezebel; Flattop; the Honeycomb assembly for Rover studies; Kiwi-TNT; PARKA reactor; Big Ten; and Plasma Cavity Assembly.

Paxton, H.C.

1981-05-01

10

Research reactors - an overview  

SciTech Connect

A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs.

West, C.D.

1997-03-01

11

Bioconversion reactor  

DOEpatents

A bioconversion reactor for the anaerobic fermentation of organic material. The bioconversion reactor comprises a shell enclosing a predetermined volume, an inlet port through which a liquid stream containing organic materials enters the shell, and an outlet port through which the stream exits the shell. A series of vertical and spaced-apart baffles are positioned within the shell to force the stream to flow under and over them as it passes from the inlet to the outlet port. The baffles present a barrier to the microorganisms within the shell causing them to rise and fall within the reactor but to move horizontally at a very slow rate. Treatment detention times of one day or less are possible.

McCarty, Perry L. (Stanford, CA); Bachmann, Andre (Palo Alto, CA)

1992-01-01

12

(Reactor dosimetry)  

SciTech Connect

The lead in most aspects of research reactor design and use passed from the USA about 15 years ago, soon after the construction of the HFIR and HFBR. The Europeans have consistently upgraded and improved their existing facilities and have built new ones including the HFR at Grenoble and ORPHEE at Saclay. They studied ultra-high flux concepts ({approximately}10{sup 20}/m{sup {minus}2}{center dot}s{sup {minus}1}) about 10 years ago, and are in the design phase of a new, highly efficient medium flux reactor to be built at Garching, near Munich in Germany. A visit was made to Interatom, the firm -- the equivalent of the Architect/Engineer for the ANS project -- responsible, under contract to the Technical University of Munich, for the new Munich reactor design. There are many similarities to the ANS design, and we reviewed and discussed technical and safety aspects of the two reactors. A request was made for some new, hitherto proprietary, experimental data on reactor thermal hydraulics and cooling that will be very valuable to the ANS project. I presented a seminar on the ANS project. A visit was made to Kernforschungszentrum Karlsruhe and knowledge was gained from Dr. Kuchle, a true pioneer of ultra-high flux reactor concepts, of their work. Dr. Kuchle kindly reviewed the ANS reference core and cooling system design (with favorable conclusions). I then talked with researchers working on materials irradiation damage and activation of structural materials by neutron irradiation, both key issues for the ANS. I was shown some new techniques they have developed for testing materials irradiation effects at high fluences, in a short time, using accelerated particle beams.

West, C.D.

1990-09-13

13

Fast reactor programme  

Microsoft Academic Search

India's fast reactor programme is described in detail. A 15 MW(e) Fast ; Breeder Test Reactor (FBTR) under construction in the Reactor Research Centre at ; Kalpakkam, will provide experience in construction and operation of a sodium ; cooled fast reactor. Fuel and material testing is an essential aspect of fast ; reactor development. For this purpose, FBTR will serve

Srinivasan

1973-01-01

14

Nuclear Reactors. Revised.  

ERIC Educational Resources Information Center

This publication is one of a series of information booklets for the general public published by the United States Atomic Energy Commission. Among the topics discussed are: How Reactors Work; Reactor Design; Research, Teaching, and Materials Testing; Reactors (Research, Teaching and Materials); Production Reactors; Reactors for Electric Power…

Hogerton, John F.

15

Photocatalytic reactor  

DOEpatents

A photocatalytic reactor is described for processing selected reactants from a fluid medium comprising at least one permeable photocatalytic membrane having a photocatalytic material. The material forms an area of chemically active sites when illuminated by light at selected wavelengths. When the fluid medium is passed through the illuminated membrane, the reactants are processed at these sites separating the processed fluid from the unprocessed fluid. A light source is provided and a light transmitting means, including an optical fiber, for transmitting light from the light source to the membrane. 4 figs.

Bischoff, B.L.; Fain, D.E.; Stockdale, J.A.D.

1999-01-19

16

Photocatalytic reactor  

DOEpatents

A photocatalytic reactor for processing selected reactants from a fluid medium comprising at least one permeable photocatalytic membrane having a photocatalytic material. The material forms an area of chemically active sites when illuminated by light at selected wavelengths. When the fluid medium is passed through the illuminated membrane, the reactants are processed at these sites separating the processed fluid from the unprocessed fluid. A light source is provided and a light transmitting means, including an optical fiber, for transmitting light from the light source to the membrane.

Bischoff, Brian L. (Knoxville, TN); Fain, Douglas E. (Oak Ridge, TN); Stockdale, John A. D. (Knoxville, TN)

1999-01-01

17

Hybrid adsorptive membrane reactor  

NASA Technical Reports Server (NTRS)

A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

Tsotsis, Theodore T. (Inventor); Sahimi, Muhammad (Inventor); Fayyaz-Najafi, Babak (Inventor); Harale, Aadesh (Inventor); Park, Byoung-Gi (Inventor); Liu, Paul K. T. (Inventor)

2011-01-01

18

Hybrid adsorptive membrane reactor  

DOEpatents

A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

Tsotsis, Theodore T. (Huntington Beach, CA); Sahimi, Muhammad (Altadena, CA); Fayyaz-Najafi, Babak (Richmond, CA); Harale, Aadesh (Los Angeles, CA); Park, Byoung-Gi (Yeosu, KR); Liu, Paul K. T. (Lafayette Hill, PA)

2011-03-01

19

Evaluation of LLNL's Nuclear Accident Dosimeters at the CALIBAN Reactor September 2010  

SciTech Connect

The Lawrence Livermore National Laboratory uses neutron activation elements in a Panasonic TLD holder as a personnel nuclear accident dosimeter (PNAD). The LLNL PNAD has periodically been tested using a Cf-252 neutron source, however until 2009, it was more than 25 years since the PNAD has been tested against a source of neutrons that arise from a reactor generated neutron spectrum that simulates a criticality. In October 2009, LLNL participated in an intercomparison of nuclear accident dosimeters at the CEA Valduc Silene reactor (Hickman, et.al. 2010). In September 2010, LLNL participated in a second intercomparison of nuclear accident dosimeters at CEA Valduc. The reactor generated neutron irradiations for the 2010 exercise were performed at the Caliban reactor. The Caliban results are described in this report. The procedure for measuring the nuclear accident dosimeters in the event of an accident has a solid foundation based on many experimental results and comparisons. The entire process, from receiving the activated NADs to collecting and storing them after counting was executed successfully in a field based operation. Under normal conditions at LLNL, detectors are ready and available 24/7 to perform the necessary measurement of nuclear accident components. Likewise LLNL maintains processing laboratories that are separated from the areas where measurements occur, but contained within the same facility for easy movement from processing area to measurement area. In the event of a loss of LLNL permanent facilities, the Caliban and previous Silene exercises have demonstrated that LLNL can establish field operations that will very good nuclear accident dosimetry results. There are still several aspects of LLNL's nuclear accident dosimetry program that have not been tested or confirmed. For instance, LLNL's method for using of biological samples (blood and hair) has not been verified since the method was first developed in the 1980's. Because LLNL and the other DOE participants were limited in what they were allowed to do at the Caliban and Silene exercises and testing of various elements of the nuclear accident dosimetry programs cannot always be performed as guests at other sites, it has become evident that DOE needs its own capability to test nuclear accident dosimeters. Angular dependence determination and correction factors for NADs desperately need testing as well as more evaluation regarding the correct determination of gamma doses. It will be critical to properly design any testing facility so that the necessary experiments can be performed by DOE laboratories as well as guest laboratories. Alternate methods of dose assessment such as using various metals commonly found in pockets and clothing have yet to be evaluated. The DOE is planning to utilize the Godiva or Flattop reactor for testing nuclear accident dosimeters. LLNL has been assigned the primary operational authority for such testing. Proper testing of nuclear accident dosimeters will require highly specific characterization of the pulse fields. Just as important as the characterization of the pulsed fields will be the design of facilities used to process the NADs. Appropriate facilities will be needed to allow for early access to dosimeters to test and develop quick sorting techniques. These facilities will need appropriate laboratory preparation space and an area for measurements. Finally, such a facility will allow greater numbers of LLNL and DOE laboratory personnel to train on the processing and interpretation of nuclear accident dosimeters and results. Until this facility is fully operational for test purposes, DOE laboratories may need to continue periodic testing as guests of other reactor facilities such as Silene and Caliban.

Hickman, D P; Wysong, A R; Heinrichs, D P; Wong, C T; Merritt, M J; Topper, J D; Gressmann, F A; Madden, D J

2011-06-21

20

Reactor safety method  

DOEpatents

This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature.

Vachon, Lawrence J. (Clairton, PA)

1980-03-11

21

Nuclear reactor  

DOEpatents

A nuclear reactor comprising a cylindrical pressure vessel, an elongated annular core centrally disposed within and spaced from the pressure vessel, and a plurality of ducts disposed longitudinally of the pressure vessel about the periphery thereof, said core comprising an annular active portion, an annular reflector just inside the active portion, and an annular reflector just outside the active a portion, said annular active portion comprising rectangular slab, porous fuel elements radially disposed around the inner reflector and extending the length of the active portion, wedge-shaped, porous moderator elements disposed adjacent one face of each fuel element and extending the length of the fuel element, the fuel and moderator elements being oriented so that the fuel elements face each other and the moderator elements do likewise, adjacent moderator elements being spaced to provide air inlet channels, and adjacent fuel elements being spaced to provide air outlet channels which communicate with the interior of the peripheral ducts, and means for introducing air into the air inlet channels which passes through the porous moderator elements and porous fuel elements to the outlet channel.

Thomson, Wallace B. (Severna Park, MD)

2004-03-16

22

Attrition reactor system  

DOEpatents

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

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

1993-09-28

23

Attrition reactor system  

DOEpatents

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

Scott, Charles D. (Oak Ridge, TN); Davison, Brian H. (Knoxvile, TN)

1993-01-01

24

The Chemical Batch Reactor  

Microsoft Academic Search

\\u000a This chapter faces the description of the main object of the book, namely, the chemical batch reactor. First, the batch reactor\\u000a is compared to the continuous reactors in the light of ideal physical models, the main ideas of chemical kinetics are reviewed,\\u000a and the relevant modeling of the isothermal batch reactor is developed. Then, the heat balance introduces elements of

Fabrizio Caccavale; Mario Iamarino; Francesco Pierri; Vincenzo Tufano

25

Advanced Test Reactor Tour  

SciTech Connect

The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored. For more information about INL research, visit http://www.facebook.com/idahonationallaboratory.

Miley, Don

2011-01-01

26

Advanced Test Reactor Tour  

ScienceCinema

The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored. For more information about INL research, visit http://www.facebook.com/idahonationallaboratory.

Miley, Don

2013-05-28

27

Structure of processes in flow reactor and closed reactor: Flow reactor  

E-print Network

Structure of processes in flow reactor and closed reactor: Flow reactor Closed reactor Active ZoneActive Zone Structure of processes in space Structure of processes in time Elementary reactions e A1 A1 A2 Ak

Greifswald, Ernst-Moritz-Arndt-Universität

28

Reactor water cleanup system  

DOEpatents

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

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

1994-12-20

29

Fission reactors and materials  

SciTech Connect

The American-designed boiling water reactor and pressurized water reactor dominate the designs currently in use and under construction worldwide. As in all energy systems, materials problems have appeared during service; these include stress-corrosion of stainless steel pipes and heat exchangers and questions regarding crack behavior in pressure vessels. To obtain the maximum potential energy from our limited uranium supplies is is essential to develop the fast breeder reactor. The materials in these reactors are subjected to higher temperatures and neutron fluxes but lower pressures than in the water reactors. The performance required of the fuel elements is more arduous in the breeder than in water reactors. Extensive materials programs are in progress in test reactors and in large test rigs to ensure that materials will be available to meet these conditions.

Frost, B.R.T.

1981-12-01

30

Reactor water cleanup system  

DOEpatents

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

Gluntz, Douglas M. (San Jose, CA); Taft, William E. (Los Gatos, CA)

1994-01-01

31

Nuclear Reactor Physics  

NASA Astrophysics Data System (ADS)

An authoritative textbook and up-to-date professional's guide to basic and advanced principles and practices Nuclear reactors now account for a significant portion of the electrical power generated worldwide. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. Nuclear reactor physics is the core discipline of nuclear engineering, and as the first comprehensive textbook and reference on basic and advanced nuclear reactor physics to appear in a quarter century, this book fills a large gap in the professional literature. Nuclear Reactor Physics is a textbook for students new to the subject, for others who need a basic understanding of how nuclear reactors work, as well as for those who are, or wish to become, specialists in nuclear reactor physics and reactor physics computations. It is also a valuable resource for engineers responsible for the operation of nuclear reactors. Dr. Weston Stacey begins with clear presentations of the basic physical principles, nuclear data, and computational methodology needed to understand both the static and dynamic behaviors of nuclear reactors. This is followed by in-depth discussions of advanced concepts, including extensive treatment of neutron transport computational methods. As an aid to comprehension and quick mastery of computational skills, he provides numerous examples illustrating step-by-step procedures for performing the calculations described and chapter-end problems. Nuclear Reactor Physics is a useful textbook and working reference. It is an excellent self-teaching guide for research scientists, engineers, and technicians involved in industrial, research, and military applications of nuclear reactors, as well as government regulators who wish to increase their understanding of nuclear reactors.

Stacey, Weston M.

2001-02-01

32

High Flux Isotope Reactor  

NSDL National Science Digital Library

The High Flux Reactor Isotope Facility (HFIR) produces transuranium isotopes for research, industrial, and medical applications. It is also used for a variety of neutron flux experiments. The HFIR Website gives an informative overview of the history, science, and engineering behind the reactor. Other features of the site include illustrated sections on horizontal beam poles and neutron scattering as well as daily operation status of the reactor.

2003-11-10

33

Tokamak reactor systems code  

SciTech Connect

A tokamak reactor systems code has been developed by combining a previously developed plasma engineering code with an existing reactor systems code and adding calculations for thermal hydraulics, stress analysis, physical sputtering, and neutron activation dose rate. Calculations from the thermal hydraulics and neutron activation dose rate modules are compared with results from more complex codes. The effects on reactor performance of unpredictable properties such as plasma profiles and confinement times are demonstrated.

Baxter, D.C.; Dabiri, A.E.

1983-09-01

34

Open-Plan Schools: Time for a Peek at Lady Godiva  

ERIC Educational Resources Information Center

Schools systems that employ flexible and new methods of teaching their students must take adequate care to insure that suitable goals are established. Only in this way can the freedom given to students be used to advantage. (CK)

Anderson, D. Carl

1970-01-01

35

SPRAY CALCINATION REACTOR  

Microsoft Academic Search

A spray calcination reactor for calcining reprocessin- g waste solutions ; is described. Coaxial within the outer shell of the reactor is a shorter inner ; shell having heated walls and with open regions above and below. When the ; solution is sprayed into the irner shell droplets are entrained by a current of ; gas that moves downwardly within

Johnson

1963-01-01

36

Tory reactor temperature measurements  

Microsoft Academic Search

Declassified 26 Nov 1973. The basic problem to be dealt with was how to ; continuously measure temperatures in the active core of a reactor that was to be ; operated as a high-temperature heat exchanger with compressed air used as the ; cooling medium. Some 345 to 375 measurements were made on the Tory IIA and IIC ; reactors.

1965-01-01

37

Tokamak reactor first wall  

NASA Astrophysics Data System (ADS)

A first wall construction for a tokamak reactor is disclosed. The wall comprises a series of hollow lobes, each with a pair of side wall portions with a curved end wall portion which extends between it. This invention relates to an improved first wall construction for a tokamak fusion reactor vessel, or other vessels subjected to similar pressure and thermal stresses.

Creedon, R. L.; Levine, H. E.; Wong, C.; Battaglia, J.

1984-11-01

38

MERCHANT MARINE SHIP REACTOR  

Microsoft Academic Search

A nuclear reactor for use in a merchant marine ship is described. The ; reactor is of pressurized, light water cooled and moderated design in which three ; passes of the water through the core in successive regions of low, intermediate, ; and high heat generation and downflow in a fuel region are made. The design ; makes a compact

M. F. Sankovich; J. F. Mumm; D. C. Jr. North; H. R. Rock; D. K. Gestson

1961-01-01

39

MERCHANT MARINE SHIP REACTOR  

Microsoft Academic Search

A nuclear reactor is described for use in a merchant marine ship. The ; reactor is of pressurized light water cooled and moderated design in which three ; passes of the water through the core in successive regions of low, intermediate, ; and high heat generation and downflow in a fuel region are made. The foregoing ; design makes a

J. F. Mumm; D. C. Jr. North; H. R. Rock; D. K. Geston

1961-01-01

40

GAS COOLED ATOMIC REACTOR  

Microsoft Academic Search

A device for holding the gas collection and distribution vessel bottoms ; and covers in spaced relation to the reactor gas tube outlets and to each other ; consists of discs with spaced holes bolted to the vessels. A ring formed from ; slit, flanged tubing provides tightening between the reactor tank and the ; vessels, and the bottom of

Liljeblad

1962-01-01

41

The secure heating reactor  

SciTech Connect

The SECURE heating reactor was designed by ASEA-ATOM as a realistic alternative for district heating in urban areas and for supplying heat to process industries. SECURE has unique safety characteristics, that are based on fundamental laws of physics. The safety does not depend on active components or operator intervention for shutdown and cooling of the reactor. The inherent safety characteristics of the plant cannot be affected by operator errors. Due to its very low environment impact, it can be sited close to heat consumers. The SECURE heating reactor has been shown to be competitive in comparison with other alternatives for heating Helsinki and Seoul. The SECURE heating reactor forms a basis for the power-producing SECURE-P reactor known as PIUS (Process Inherent Ultimate Safety), which is based on the same inherent safety principles. The thermohydraulic function and transient response have been demonstrated in a large electrically heated loop at the ASEA-ATOM laboratories.

Pind, C.

1987-11-01

42

Reactor Safety Research Programs  

SciTech Connect

This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from January 1 through March 31, 1981, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipeto- pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-ofcoolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

Edler, S. K.

1981-07-01

43

Slurry reactor design studies  

SciTech Connect

The objective of these studies was to perform a realistic evaluation of the relative costs of tublar-fixed-bed and slurry reactors for methanol, mixed alcohols and Fischer-Tropsch syntheses under conditions where they would realistically be expected to operate. The slurry Fischer-Tropsch reactor was, therefore, operated at low H{sub 2}/CO ratio on gas directly from a Shell gasifier. The fixed-bed reactor was operated on 2.0 H{sub 2}/CO ratio gas after adjustment by shift and CO{sub 2} removal. Every attempt was made to give each reactor the benefit of its optimum design condition and correlations were developed to extend the models beyond the range of the experimental pilot plant data. For the methanol design, comparisons were made for a recycle plant with high methanol yield, this being the standard design condition. It is recognized that this is not necessarily the optimum application for the slurry reactor, which is being proposed for a once-through operation, coproducing methanol and power. Consideration is also given to the applicability of the slurry reactor to mixed alcohols, based on conditions provided by Lurgi for an Octamix{trademark} plant using their standard tubular-fixed reactor technology. 7 figs., 26 tabs.

Fox, J.M.; Degen, B.D.; Cady, G.; Deslate, F.D.; Summers, R.L. (Bechtel Group, Inc., San Francisco, CA (USA)); Akgerman, A. (Texas A and M Univ., College Station, TX (USA)); Smith, J.M. (California Univ., Davis, CA (USA))

1990-06-01

44

REACTOR BASE, SOUTHEAST CORNER. INTERIOR WILL CONTAIN REACTOR TANK, COOLING ...  

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

REACTOR BASE, SOUTHEAST CORNER. INTERIOR WILL CONTAIN REACTOR TANK, COOLING WATER PIPES, COOLING AIR DUCTS, AND SHIELDING. INL NEGATIVE NO. 776. Unknown Photographer, 10/1950 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

45

Spherical torus fusion reactor  

DOEpatents

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.

Martin Peng, Y.K.M.

1985-10-03

46

Methanogenesis in thermophilic biogas reactors  

Microsoft Academic Search

Methanogenesis in thermophilic biogas reactors fed with different wastes is examined. The specific methanogenic activity with acetate or hydrogen as substrate reflected the organic loading of the specific reactor examined. Increasing the loading of thermophilic reactors stabilized the process as indicated by a lower concentration of volatile fatty acids in the effluent from the reactors. The specific methanogenic activity in

Birgitte Kiær Ahring

1995-01-01

47

Packed Bed Reactor Experiment  

NASA Video Gallery

The purpose of the Packed Bed Reactor Experiment in low gravity is to determine how a mixture of gas and liquid flows through a packed bed in reduced gravity. A packed bed consists of a metal pipe ...

48

Advanced spheromak fusion reactor  

SciTech Connect

The spheromak has no toroidal magnetic field coils or other structure along its geometric axis, and is thus more attractive than the leading magnetic fusion reactor concept, the tokamak. As a consequence of this and other attributes, the spheromak reactor may be compact and produce a power density sufficiently high to warrant consideration of a liquid `blanket` that breeds tritium, converts neutron kinetic energy to heat, and protects the reactor vessel from severe neutron damage. However, the physics is more complex, so that considerable research is required to learn how to achieve the reactor potential. Critical physics problems and possible ways of solving them are described. The opportunities and issues associated with a possible liquid wall are considered to direct future research.

Fowler, T.K. [California Univ., Berkeley, CA (United States). Dept. of Nuclear Engineering; Hooper, E.B. [Lawrence Livermore National Lab., CA (United States)

1996-06-19

49

Molten metal reactors  

DOEpatents

A molten metal reactor for converting a carbon material and steam into a gas comprising hydrogen, carbon monoxide, and carbon dioxide is disclosed. The reactor includes an interior crucible having a portion contained within an exterior crucible. The interior crucible includes an inlet and an outlet; the outlet leads to the exterior crucible and may comprise a diffuser. The exterior crucible may contain a molten alkaline metal compound. Contained between the exterior crucible and the interior crucible is at least one baffle.

Bingham, Dennis N; Klingler, Kerry M; Turner, Terry D; Wilding, Bruce M

2013-11-05

50

K-Reactor readiness  

SciTech Connect

This document describes some of the more significant accomplishments in the reactor restart program and details the magnitude and extent of the work completed to bring K-Reactor to a state of restart readiness. The discussion of restart achievements is organized into the three major categories of personnel, programs, and plant. Also presented is information on the scope and extent of internal and external oversight of the efforts, as well as some details on the startup plan.

Rice, P.D.

1991-12-04

51

Nuclear power technology. Volume 1. Reactor technology  

Microsoft Academic Search

This book presents papers on the nuclear industry. Topics considered include the basics of reactor operation, reactor physics, gas cooled reactors, light water reactors, fast reactors, the UKAEA interest in heavy water reactors, novel reactor concepts, prospects for fusion, fuel recycling, uranium and thorium raw materials, uranium supply, uranium enrichment, fuel design and fabrication, nuclear fuel reprocessing in the UK,

1983-01-01

52

REACTOR GROUT THERMAL PROPERTIES  

SciTech Connect

Savannah River Site has five dormant nuclear production reactors. Long term disposition will require filling some reactor buildings with grout up to ground level. Portland cement based grout will be used to fill the buildings with the exception of some reactor tanks. Some reactor tanks contain significant quantities of aluminum which could react with Portland cement based grout to form hydrogen. Hydrogen production is a safety concern and gas generation could also compromise the structural integrity of the grout pour. Therefore, it was necessary to develop a non-Portland cement grout to fill reactors that contain significant quantities of aluminum. Grouts generate heat when they set, so the potential exists for large temperature increases in a large pour, which could compromise the integrity of the pour. The primary purpose of the testing reported here was to measure heat of hydration, specific heat, thermal conductivity and density of various reactor grouts under consideration so that these properties could be used to model transient heat transfer for different pouring strategies. A secondary purpose was to make qualitative judgments of grout pourability and hardened strength. Some reactor grout formulations were unacceptable because they generated too much heat, or started setting too fast, or required too long to harden or were too weak. The formulation called 102H had the best combination of characteristics. It is a Calcium Alumino-Sulfate grout that contains Ciment Fondu (calcium aluminate cement), Plaster of Paris (calcium sulfate hemihydrate), sand, Class F fly ash, boric acid and small quantities of additives. This composition afforded about ten hours of working time. Heat release began at 12 hours and was complete by 24 hours. The adiabatic temperature rise was 54 C which was within specification. The final product was hard and displayed no visible segregation. The density and maximum particle size were within specification.

Steimke, J.; Qureshi, Z.; Restivo, M.; Guerrero, H.

2011-01-28

53

Reactor antineutrino anomaly  

SciTech Connect

Recently, new reactor antineutrino spectra have been provided for {sup 235}U, {sup 239}Pu, {sup 241}Pu, and {sup 238}U, increasing the mean flux by about 3%. To a good approximation, this reevaluation applies to all reactor neutrino experiments. The synthesis of published experiments at reactor-detector distances <100 m leads to a ratio of observed event rate to predicted rate of 0.976{+-}0.024. With our new flux evaluation, this ratio shifts to 0.943{+-}0.023, leading to a deviation from unity at 98.6% C.L. which we call the reactor antineutrino anomaly. The compatibility of our results with the existence of a fourth nonstandard neutrino state driving neutrino oscillations at short distances is discussed. The combined analysis of reactor data, gallium solar neutrino calibration experiments, and MiniBooNE-{nu} data disfavors the no-oscillation hypothesis at 99.8% C.L. The oscillation parameters are such that |{Delta}m{sub new}{sup 2}|>1.5 eV{sup 2} (95%) and sin{sup 2}(2{theta}{sub new})=0.14{+-}0.08 (95%). Constraints on the {theta}{sub 13} neutrino mixing angle are revised.

Mention, G.; Fechner, M. [CEA, Irfu, SPP, Centre de Saclay, F-91191 Gif-sur-Yvette (France); Lasserre, Th.; Cribier, M. [CEA, Irfu, SPP, Centre de Saclay, F-91191 Gif-sur-Yvette (France); Astroparticule et Cosmologie APC, 10 rue Alice Domon et Leonie Duquet, 75205 Paris cedex 13 (France); Mueller, Th. A.; Lhuillier, D.; Letourneau, A. [CEA, Irfu, SPhN, Centre de Saclay, F-91191 Gif-sur-Yvette (France)

2011-04-01

54

Heat dissipating nuclear reactor  

DOEpatents

Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extend from the metal base plate downwardly and outwardly into the earth.

Hunsbedt, A.; Lazarus, J.D.

1985-11-21

55

Reactor for exothermic reactions  

DOEpatents

A liquid phase process is described for oligomerization of C[sub 4] and C[sub 5] isoolefins or the etherification thereof with C[sub 1] to C[sub 6] alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120 to 300 F. Wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled.

Smith, L.A. Jr.; Hearn, D.; Jones, E.M. Jr.

1993-03-02

56

Dynamic bed reactor  

SciTech Connect

A dynamic bed reactor is disclosed in which a compressible open cell foam matrix is periodically compressed and expanded to move a liquid or fluid through the matrix. In preferred embodiments, the matrix contains an active material such as an enzyme, biological cell, chelating agent, oligonucleotide, adsorbent or other material that acts upon the liquid or fluid passing through the matrix. The active material may be physically immobilized in the matrix, or attached by covalent or ionic bonds. Microbeads, substantially all of which have diameters less than 50 microns, can be used to immobilize the active material in the matrix and further improve reactor efficiency. A particularly preferred matrix is made of open cell polyurethane foam, which adsorbs pollutants such as polychlorophenol or o-nitrophenol. The reactors of the present invention allow unidirectional non-laminar flow through the matrix, and promote intimate exposure of liquid reactants to active agents such as microorganisms immobilized in the matrix. 27 figs.

Stormo, K.E.

1996-07-02

57

Thermionic Reactor Design Studies  

SciTech Connect

Paper presented at the 29th IECEC in Monterey, CA in August 1994. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their (thermionic reactor) performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling.

Schock, Alfred

1994-08-01

58

Looking Southwest at Reactor Box Furnaces With Reactor Boxes and ...  

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

Looking Southwest at Reactor Box Furnaces With Reactor Boxes and Repossessed Uranium in Recycle Recovery Building - Hematite Fuel Fabrication Facility, Recycle Recovery Building, 3300 State Road P, Festus, Jefferson County, MO

59

Fusion reactor pumped laser  

DOEpatents

A nuclear pumped laser capable of producing long pulses of very high power laser radiation is provided. A toroidal fusion reactor provides energetic neutrons which are slowed down by a moderator. The moderated neutrons are converted to energetic particles capable of pumping a lasing medium. The lasing medium is housed in an annular cell surrounding the reactor. The cell includes an annular reflecting mirror at the bottom and an annular output window at the top. A neutron reflector is disposed around the cell to reflect escaping neutrons back into the cell. The laser radiation from the annular window is focused onto a beam compactor which generates a single coherent output laser beam.

Jassby, Daniel L. (Princeton, NJ)

1988-01-01

60

Fast quench reactor method  

DOEpatents

A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a means of rapidly expanding a reactant stream, such as a restrictive convergent-divergent nozzle at its outlet end. Metal halide reactants are injected into the reactor chamber. Reducing gas is added at different stages in the process to form a desired end product and prevent back reactions. The resulting heated gaseous stream is then rapidly cooled by expansion of the gaseous stream.

Detering, Brent A. (Idaho Falls, ID); Donaldson, Alan D. (Idaho Falls, ID); Fincke, James R. (Idaho Falls, ID); Kong, Peter C. (Idaho Falls, ID); Berry, Ray A. (Idaho Falls, ID)

1999-01-01

61

Leny: the reactor builder  

SciTech Connect

An interview with Jean-Claude Leny of Framatome, which has built over 40 pressurized water reactors and has projects underway that will raise the total to over 70 explores the contract problems involving plant construction in China, the reasons why Framatome did not bid on contract in Turkey, license agreements with Westinghouse, and expansion of Framatome's business beyond construction to servicing plants. Leny reviews the problems of industrial reorganization of recent years and Framatome's structural relationship with customers and shareholders. While acknowledging the rapid development of fast reactors, Leny emphasizes that the company will be ready to proceed whenever governments decide to pursue the technology.

Not Available

1987-01-01

62

Diagnostics for hybrid reactors  

SciTech Connect

The Hybrid Reactor(HR) can be considered an attractive actinide-burner or a fusion assisted transmutation for destruction of transuranic(TRU) nuclear waste. The hybrid reactor has two important subsystems: the tokamak neutron source and the blanket which includes a fuel zone where the TRU are placed and a tritium breeding zone. The diagnostic system for a HR must be as simple and robust as possible to monitor and control the plasma scenario, guarantee the protection of the machine and monitor the transmutation.

Orsitto, Francesco Paolo [ENEA Unita' Tecnica Fusione , Associazione ENEA-EURATOM sulla Fusione C R Frascati v E Fermi 45 00044 Frascati (Italy)

2012-06-19

63

Particle bed reactor modeling  

NASA Technical Reports Server (NTRS)

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

Sapyta, Joe; Reid, Hank; Walton, Lew

1993-01-01

64

Reactor Neutrino Experiments  

E-print Network

Precisely measuring $\\theta_{13}$ is one of the highest priority in neutrino oscillation study. Reactor experiments can cleanly determine $\\theta_{13}$. Past reactor neutrino experiments are reviewed and status of next precision $\\theta_{13}$ experiments are presented. Daya Bay is designed to measure $\\sin^22\\theta_{13}$ to better than 0.01 and Double Chooz and RENO are designed to measure it to 0.02-0.03. All are heading to full operation in 2010. Recent improvements in neutrino moment measurement are also briefed.

Jun Cao

2007-12-06

65

Perspectives on reactor safety  

SciTech Connect

The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor, safety concepts. The course consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

Haskin, F.E. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering; Camp, A.L. [Sandia National Labs., Albuquerque, NM (United States)

1994-03-01

66

Reactor operation environmental information document  

SciTech Connect

The Savannah River Site (SRS) produces nuclear materials, primarily plutonium and tritium, to meet the requirements of the Department of Defense. These products have been formed in nuclear reactors that were built during 1950--1955 at the SRS. K, L, and P reactors are three of five reactors that have been used in the past to produce the nuclear materials. All three of these reactors discontinued operation in 1988. Currently, intense efforts are being extended to prepare these three reactors for restart in a manner that protects human health and the environment. To document that restarting the reactors will have minimal impacts to human health and the environment, a three-volume Reactor Operations Environmental Impact Document has been prepared. The document focuses on the impacts of restarting the K, L, and P reactors on both the SRS and surrounding areas. This volume discusses the geology, seismology, and subsurface hydrology. 195 refs., 101 figs., 16 tabs.

Haselow, J.S.; Price, V.; Stephenson, D.E.; Bledsoe, H.W.; Looney, B.B.

1989-12-01

67

Innovative design of uranium startup fast reactors  

E-print Network

Sodium Fast Reactors are one of the three candidates of GEN-IV fast reactors. Fast reactors play an important role in saving uranium resources and reducing nuclear wastes. Conventional fast reactors rely on transuranic ...

Fei, Tingzhou

2012-01-01

68

Nuclear reactor operation control process  

Microsoft Academic Search

A method is described for controlling the operation of a nuclear reactor to increase the reactor power in a range in which pellet-clad mechanical-interaction occurs. The method includes the steps of increasing the reactor power from a power level in which pellet clad-mechanical-interaction begins to take place up to a predetermined power level for the nuclear reactor and controlling the

Hayashi

1981-01-01

69

Control of nuclear reactors  

SciTech Connect

In a coupling system for a nuclear reactor control rod and control rod drive, said coupling system being selectively uncouplable by an uncoupling rod, a system of passages for exercising a countervailing force on the uncoupling rod to prevent inadvertent uncoupling during scram.

Fischer, L. E.; Bean, J. E.

1985-06-04

70

NETL - Chemical Looping Reactor  

SciTech Connect

NETL's Chemical Looping Reactor unit is a high-temperature integrated CLC process with extensive instrumentation to improve computational simulations. A non-reacting test unit is also used to study solids flow at ambient temperature. The CLR unit circulates approximately 1,000 pounds per hour at temperatures around 1,800 degrees Fahrenheit.

None

2013-07-24

71

Nuclear reactor building  

DOEpatents

A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed there above. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define there between an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin. 4 figures.

Gou, P.F.; Townsend, H.E.; Barbanti, G.

1994-04-05

72

Fossil fuel furnace reactor  

DOEpatents

A fossil fuel furnace reactor is provided for simulating a continuous processing plant with a batch reactor. An internal reaction vessel contains a batch of shale oil, with the vessel having a relatively thin wall thickness for a heat transfer rate effective to simulate a process temperature history in the selected continuous processing plant. A heater jacket is disposed about the reactor vessel and defines a number of independent controllable temperature zones axially spaced along the reaction vessel. Each temperature zone can be energized to simulate a time-temperature history of process material through the continuous plant. A pressure vessel contains both the heater jacket and the reaction vessel at an operating pressure functionally selected to simulate the continuous processing plant. The process yield from the oil shale may be used as feedback information to software simulating operation of the continuous plant to provide operating parameters, i.e., temperature profiles, ambient atmosphere, operating pressure, material feed rates, etc., for simulation in the batch reactor.

Parkinson, William J. (Los Alamos, NM)

1987-01-01

73

Cermet fuel reactors  

SciTech Connect

Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. Key features of the cermet fueled reactor design are (1) the ability to achieve very high coolant exit temperatures, and (2) thermal shock resistance during rapid power changes, and (3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, thre is a potential for achieving a long operating life because of (1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and (2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core. In addition, the neutronic properties of the refractory materials assure that the reactor remains substantially subcritical under conditions of water immersion. It is concluded that cermet fueled reactors can be utilized to meet the power requirements for a broad range of advanced space applications. 4 refs., 4 figs., 3 tabs.

Cowan, C.L.; Palmer, R.S.; Van Hoomissen, J.E.; Bhattacharyya, S.K.; Barner, J.O.

1987-09-01

74

Future of nuclear reactors  

Microsoft Academic Search

Nuclear technology makes materials available which may be used in nuclear explosives. This problem cannot be solved either in a purely technical manner nor by a purely political approach. Nuclear reactors may be most useful in developed countries, and they would drive down oil prices. Various technical ways of modifying nuclear technology to impede the production of weapons are described:

1976-01-01

75

NETL - Chemical Looping Reactor  

ScienceCinema

NETL's Chemical Looping Reactor unit is a high-temperature integrated CLC process with extensive instrumentation to improve computational simulations. A non-reacting test unit is also used to study solids flow at ambient temperature. The CLR unit circulates approximately 1,000 pounds per hour at temperatures around 1,800 degrees Fahrenheit.

None

2014-06-26

76

Progressing Batch Hydrolysis Reactor  

SciTech Connect

In all dilute acid hydrolysis processes for glucose production, conditions severe enough to hydrolyze crystalline cellulose to glucose are also severe enough to degrade the glucose into undesirable compounds such as hydroxy-methylfurfural (HMF), levulinic acid, and formic acid. One way to minimize the sugar degradation is to remove the sugars from the reaction zone before substantial degradation occurs. Sugars are most efficiently removed by a reactor system that uses countercurrent flow of liquids and solids, which allows simultaneous achievement of high yields and high sugar concentrations. The progressing batch hydrolysis process, invented and now under development at SERI, uses several percolation reactors in series to simulate countercurrent flow of liquids and solids. In this way, the advantages of countercurrent flow are achieved, and the mechanical and operational simplicity of the percolation reactor is retained. This paper describes the theory and operation of the progressing batch hydrolysis reactor and presents the results of our mathematical modeling of the system. 25 refs., 7 figs.

Wright, J.D.; Bergeron, P.W.; Werdene, P.J.

1985-09-01

77

Thermal Reactor Safety  

SciTech Connect

Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods.

Not Available

1980-06-01

78

Reactor Monitoring with Neutrinos  

E-print Network

The fundamental knowledge on neutrinos acquired in the recent years open the possibility of applied neutrino physics. Among it the automatic and non intrusive monitoring of nuclear reactor by its antineutrino signal could be very valuable to IAEA in charge of the control of nuclear power plants. Several efforts worldwide have already started.

M. Cribier

2007-04-06

79

Reactor control rod timing system  

Microsoft Academic Search

A fluid driven jet-edge whistle timing system for control rods of a nuclear reactor for producing real-time detection of the timing of each control rod in its scram operation. An important parameter in reactor safety, particularly for liquid metal fast breeder reactors (LMFBR), is the time deviation between the time the control rod is released and the time the rod

Peter T. K

1982-01-01

80

Nuclear reactor with control rods  

Microsoft Academic Search

A liquid-cooled nuclear reactor including fuel assemblies mounted within a reactor vessel having linearly movable control rods passing through control rod guide tubes into respective aligned fuel assemblies is described. Reactor coolant circulates through the assemblies. Guide tubes and other vessel internals structures located above the assemblies and is discharged through an outlet nozzle positioned above the elevation of primary

F. D. Obermeyer; R. T. Berringer

1979-01-01

81

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

NASA Technical Reports Server (NTRS)

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

Roth, R. J.

1976-01-01

82

Reactor vessel support system. [LMFBR  

DOEpatents

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

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

1980-05-09

83

Integrated Microfluidic Reactors  

PubMed Central

Summary Microfluidic reactors exhibit intrinsic advantages of reduced chemical consumption, safety, high surface-area-to-volume ratios, and improved control over mass and heat transfer superior to the macroscopic reaction setting. In contract to a continuous-flow microfluidic system composed of only a microchannel network, an integrated microfluidic system represents a scalable integration of a microchannel network with functional microfluidic modules, thus enabling the execution and automation of complicated chemical reactions in a single device. In this review, we summarize recent progresses on the development of integrated microfluidics-based chemical reactors for (i) parallel screening of in situ click chemistry libraries, (ii) multistep synthesis of radiolabeled imaging probes for positron emission tomography (PET), (iii) sequential preparation of individually addressable conducting polymer nanowire (CPNW), and (iv) solid-phase synthesis of DNA oligonucleotides. These proof-of-principle demonstrations validate the feasibility and set a solid foundation for exploring a broad application of the integrated microfluidic system. PMID:20209065

Lin, Wei-Yu; Wang, Yanju; Wang, Shutao; Tseng, Hsian-Rong

2009-01-01

84

Fusion reactor pumped laser  

DOEpatents

A nuclear pumped laser capable of producing long pulses of very high power laser radiation is provided. A toroidal fusion reactor provides energetic neutrons which are slowed down by a moderator. The moderated neutrons are converted to energetic particles capable of pumping a lasing medium. The lasing medium is housed in an annular cell surrounding the reactor. The cell includes an annular reflecting mirror at the bottom and an annular output window at the top. A neutron reflector is disposed around the cell to reflect escaping neutrons back into the cell. The laser radiation from the annular window is focused onto a beam compactor which generates a single coherent output laser beam. 10 figs.

Jassby, D.L.

1987-09-04

85

Nuclear reactor shutdown system  

DOEpatents

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

Bhate, Suresh K. (Niskayuna, NY); Cooper, Martin H. (Monroeville, PA); Riffe, Delmar R. (Murrysville, PA); Kinney, Calvin L. (Penn Hills, PA)

1981-01-01

86

Gaseous fuel reactor research  

NASA Technical Reports Server (NTRS)

The paper reviews studies dealing with the concept of a gaseous fuel reactor and describes the structure and plans of the current NASA research program of experiments on uranium hexafluoride systems and uranium plasma systems. Results of research into the basic properties of uranium plasmas and fissioning gases are reported. The nuclear pumped laser is described, and the main results of experiments with these devices are summarized.

Thom, K.; Schneider, R. T.

1977-01-01

87

Fissioning Plasma Core Reactor  

NASA Technical Reports Server (NTRS)

Institute for Scientific Research, Inc. (ISR) research program consist of: 1.Study core physics by adapting existing codes: MCNP4C - Monte Carlo code; COMBINE/VENTURE - diffusion theory; SCALE4 - Monte Carlo, with many utility codes. 2. Determine feasibility and study major design parameters: fuel selection, temperature and reflector sizing. 3. Study reactor kinetics: develop QCALC1 to model point kinetics; study dynamic behavior of the power release.

Albright, Dennis; Butler, Carey; West, Nicole; Cole, John W. (Technical Monitor)

2002-01-01

88

Polarized advanced fuel reactors  

SciTech Connect

The d-/sup 3/He reaction has the same spin dependence as the d-t reaction. It produces no neutrons, so that if the d-d reactivity could be reduced, it would lead to a neutron-lean reactor. The current understanding of the possible suppression of the d-d reactivity by spin polarization is discussed. The question as to whether a suppression is possible is still unresolved. Other advanced fuel reactions are briefly discussed. 11 refs.

Kulsrud, R.M.

1987-07-01

89

Reactor vessel sectioning demonstration  

SciTech Connect

A technical demonstration was successfully completed of simulated reactor vessel sectioning using the combined techniques of air arc gouging and flame cutting. A 4-ft x 3-ft x 9-in. thick sample was fabricated of A36 carbon steel to simulate a reactor vessel wall. A 1/4-in. layer of stainless steel (SS) was tungsten inert gas (TIG)-welded to the carbon steel. Several techniques were considered to section the simulated reactor vessel; air arc gouging was selected to penetrate the stainless steel, and flame cutting was selected to sever the carbon steel. Three sectioning operations were demonstrated. For all three, the operating parameters were the same; but the position of the sample was varied. For the first cut, the sample was placed in a horizontal position, and it was successfully severed from the SS side. For the second cut, the sample was turned over and cut from the carbon steel side. Cutting from the carbon steel side has the advantages of cost reduction

Lundgren, R.A.

1981-09-01

90

Modeling Chemical Reactors I: Quiescent Reactors  

E-print Network

We introduce a fully generalized quiescent chemical reactor system in arbitrary space $\\vdim =1,2$ or 3, with $n\\in\\mathbb{N}$ chemical constituents $\\alpha_{i}$, where the character of the numerical solution is strongly determined by the relative scaling between the local reactivity of species $\\alpha_{i}$ and the local functional diffusivity $\\mathscr{D}_{ij}(\\alpha)$ of the reaction mixture. We develop an operator time-splitting predictor multi-corrector RK--LDG scheme, and utilize $hp$-adaptivity relying only on the entropy $\\mathscr{S}_{\\mathfrak{R}}$ of the reactive system $\\mathfrak{R}$. This condition preserves these bounded nonlinear entropy functionals as a necessarily enforced stability condition on the coupled system. We apply this scheme to a number of application problems in chemical kinetics; including a difficult classical problem arising in nonequilibrium thermodynamics known as the Belousov-Zhabotinskii reaction where we utilize a concentration-dependent diffusivity tensor $\\mathscr{D}_{ij}(\\alpha)$, in addition to solving a simple equilibrium problem in order to evaluate the numerical error behavior.

C. E. Michoski; J. A. Evans; P. G. Schmitz

2010-12-28

91

Thermionic Reactor Design Studies  

SciTech Connect

During the 1960's and early 70's the author performed extensive design studies, analyses, and tests aimed at thermionic reactor concepts that differed significantly from those pursued by other investigators. Those studies, like most others under Atomic Energy Commission (AEC and DOE) and the National Aeronautics and Space Administration (NASA) sponsorship, were terminated in the early 1970's. Some of this work was previously published, but much of it was never made available in the open literature. U.S. interest in thermionic reactors resumed in the early 80's, and was greatly intensified by reports about Soviet ground and flight tests in the late 80's. This recent interest resulted in renewed U.S. thermionic reactor development programs, primarily under Department of Defense (DOD) and Department of Energy (DOE) sponsorship. Since most current investigators have not had an opportunity to study all of the author's previous work, a review of the highlights of that work may be of value to them. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling. Where the author's concepts differed from the later Topaz-2 design was in the relative location of the emitter and the collector. Placing the fueled emitter on the outside of the cylindrical diodes permits much higher axial conductances to reduce ohmic losses in the electrodes of full-core-height diodes. Moreover, placing the fuel on the outside of the diode makes possible reactors with much higher fuel volume fractions, which enable power-flattened fast reactors scalable to very low power levels without the need for life-limiting hydride moderators or the use of efficiency-limiting driver fuel. In addition, with the fuel on the outside its swelling does not increase the emitter diameter or reduce the interelectrode gap. This should permit long lifetimes even with closer spacings, which can significantly improve the system efficiences. This was confirmed by coupled neutronic, thermal, thermionic, and electrical system analyses - some of which are presented in this paper - and by subsequent experiments. A companion paper presented next describes the fabrication and testing of full-scale converter elements, both fueled and unfueled, and summarizes the test results obtained. There is a duplicate copy in the file.

Schock, Alfred

1994-06-01

92

Reactor Monitoring with Neutrino Detectors  

NASA Astrophysics Data System (ADS)

The study of the use of neutrino detectors to monitor nuclear reactors is currently a very active field of research. While neutrino detectors located close to reactors have been used to provide information about the global performance of the reactors, a general improvement of the technique is needed in order to use it in a practical way to monitor the fissile contents of the fuel of the nuclear reactors or the thermal power delivered. I describe the current status of the Angra Neutrino Project, aimed to building a low-mass neutrino detector to monitor the Angra II reactor of the Brazilian nuclear power plant Almirante Alvaro Ramos in order to explore new approaches to reactor monitoring with neutrino detectors.

Casimiro Linares, Edgar

2011-09-01

93

Reactor steam isolation cooling system  

SciTech Connect

This patent describes a reactor steam isolation cooling system. It comprises: a containment building having a containment wall; a reactor pressure vessel disposed inside the containment building and including a nuclear reactor core therein operable for generating reactor steam; an isolation pool disposed outside the containment building and adjacent to the containment wall and containing pool water; an isolation condenser including: a plurality of parallel heat pipes; a tube sheet disposed between the hot and cold tubes and through which the heat pipes are disposed in sealing contact therewith; and means for selectively channeling the reactor steam from the pressure vessel between the hot tubes of the evaporator assembly for removing heat therefrom to form reactor condensate.

Dillmann, C.W.

1992-10-27

94

Fast quench reactor and method  

DOEpatents

A fast quench reactor includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This ``freezes`` the desired end product(s) in the heated equilibrium reaction stage. 7 figs.

Detering, B.A.; Donaldson, A.D.; Fincke, J.R.; Kong, P.C.

1998-05-12

95

Prospects for spheromak fusion reactors  

SciTech Connect

The reactor study of Hagenson and Krakowski demonstrated the attractiveness of the spheromak as a compact fusion reactor, based on physics principles confirmed in CTX experiments in many respects. Most uncertain was the energy confinement time and the role of magnetic turbulence inherent in the concept. In this paper, a one-dimensional model of heat confinement, calibrated by CTX, predicts negligible heat loss by magnetic turbulence at reactor scale.

Fowler, T.K.; Hua, D.D. [Univ. of California, Berkeley, CA (United States)

1995-06-01

96

Transport Reactor Facility  

SciTech Connect

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

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

1996-12-31

97

Method of operating a nuclear reactor  

Microsoft Academic Search

In raising the power of a nuclear reactor, before the linear heat generating rate of nuclear fuel elements arranged in the core of the nuclear reactor reaches 240 W\\/cm, the power rise of the reactor is suspended at least once and the reactor power is held at the fixed level. The raise of the power of the nuclear reactor before

K. Shinbo; T. Hosokawa

1979-01-01

98

Mathematical simulation of fixed bed reactors  

Microsoft Academic Search

Anaerobic digestion of agricultural wastes using conventional stirred-reactor technology has been unsuccessful in treatment of dilute waste streams. The development of ''fixed bed'' anaerobic reactors has provided an effective method of utilizing dilute waste for methane production. Fixed bed reactors, or anaerobic filters, are unique among anaerobic reactor designs in that a fixed support medium is placed inside the reactor

J. P. Bolte; R. A. Nordstedt; M. V. Thomas

1983-01-01

99

Robust controller design for batch polymerization reactors  

Microsoft Academic Search

This paper discusses several robust controller design methods for the batch polymerization reactors. The dynamic characteristics of batch polymerization reactors considerably change according to the progress of reaction because the reactors have strong nonlinearity. The reactor models have parameter uncertainty. On the other hand, both reactor temperature control and polymer quality control are important specifications for batch polymerization. In this

T. Yumotot; T. Ohtani; H. Ohmori; A. Sano

1994-01-01

100

Sequencing Batch Reactors  

Microsoft Academic Search

\\u000a A sequencing batch reactor (SBR) can be either a biological SBR (BIO-SBR) or a physicochemical SBR (PC-SBR). BIO-SBR includes\\u000a traditional sedimentation biological SBR, innovative flotation biological SBR (BIO-DAF-SBR), innovative membrane biological\\u000a SBR (MBR-SBR), aerobic digestion SBR (AD-SBR), etc. All PC-SBR are innovative processes including at least sedimentation PC-SBR\\u000a (PC-SED-SBR), flotation PC-SBR (PC-DAF-SBR), membrane PC-SBR (PC-membrance-SBR), granular activated carbon PC-SBR (PC-GAC-SBR),

Lawrence K. Wang; Yan Li

101

Fast Reactor Technology Preservation  

SciTech Connect

There is renewed worldwide interest in developing and implementing a new generation of advanced fast reactors. International cooperative efforts are underway such as the Global Nuclear Energy Partnership (GNEP). Advanced computer modeling and simulation efforts are a key part of these programs. A recognized and validated set of Benchmark Cases are an essential component of such modeling efforts. Testing documentation developed during the operation of the Fast Flux Test Facility (FFTF) provide the information necessary to develop a very useful set of Benchmark Cases.

Wootan, David W.; Omberg, Ronald P.

2008-01-11

102

Reactor refueling containment system  

DOEpatents

A method of refueling a nuclear reactor whereby the drive mechanism is disengaged and removed by activating a jacking mechanism that raises the closure head. The area between the barrier plate and closure head is exhausted through the closure head penetrations. The closure head, upper drive mechanism, and bellows seal are lifted away and transported to a safe area. The barrier plate acts as the primary boundary and each drive and control rod penetration has an elastomer seal preventing excessive tritium gases from escaping. The individual instrumentation plugs are disengaged allowing the corresponding fuel assembly to be sealed and replaced.

Gillett, James E. (Greensburg, PA); Meuschke, Robert E. (Pittsburgh, PA)

1995-01-01

103

Fast neutron nuclear reactor  

SciTech Connect

The invention relates to a fast neutron nuclear reactor of the integrated type comprising a cylindrical inner vessel. The inner vessel comprises two concentric ferrules and the connection between the hot collector defined within this vessel and the inlet port of the exchangers is brought about by a hot structure forming a heat baffle and supported by the inner ferrule and by a cold structure surrounding the hot structure, supported by the outer ferrule and sealingly connected to the exchanger. Application to the generation of electric power in nuclear power stations.

Cabrillat, M. Th.; Lions, N.

1985-01-08

104

Biparticle fluidized bed reactor  

DOEpatents

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

Scott, C.D.

1993-12-14

105

Reactor refueling containment system  

DOEpatents

A method of refueling a nuclear reactor is disclosed whereby the drive mechanism is disengaged and removed by activating a jacking mechanism that raises the closure head. The area between the barrier plate and closure head is exhausted through the closure head penetrations. The closure head, upper drive mechanism, and bellows seal are lifted away and transported to a safe area. The barrier plate acts as the primary boundary and each drive and control rod penetration has an elastomer seal preventing excessive tritium gases from escaping. The individual instrumentation plugs are disengaged allowing the corresponding fuel assembly to be sealed and replaced. 2 figs.

Gillett, J.E.; Meuschke, R.E.

1995-05-02

106

High flux reactor  

DOEpatents

A high flux reactor is comprised of a core which is divided into two symetric segments housed in a pressure vessel. The core segments include at least one radial fuel plate. The spacing between the plates functions as a coolant flow channel. The core segments are spaced axially apart such that a coolant mixing plenum is formed between them. A channel is provided such that a portion of the coolant bypasses the first core section and goes directly into the mixing plenum. The outlet coolant from the first core segment is mixed with the bypass coolant resulting in a lower inlet temperature to the lower core segment.

Lake, James A. (Idaho Falls, ID); Heath, Russell L. (Idaho Falls, ID); Liebenthal, John L. (Idaho Falls, ID); DeBoisblanc, Deslonde R. (Summit, NJ); Leyse, Carl F. (Idaho Falls, ID); Parsons, Kent (Idaho Falls, ID); Ryskamp, John M. (Idaho Falls, ID); Wadkins, Robert P. (Idaho Falls, ID); Harker, Yale D. (Idaho Falls, ID); Fillmore, Gary N. (Idaho Falls, ID); Oh, Chang H. (Idaho Falls, ID)

1988-01-01

107

Biparticle fluidized bed reactor  

DOEpatents

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

Scott, Charles D. (Oak Ridge, TN)

1993-01-01

108

Biparticle fluidized bed reactor  

DOEpatents

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

Scott, Charles D. (Oak Ridge, TN); Marasco, Joseph A. (Kingston, TN)

1995-01-01

109

Biparticle fluidized bed reactor  

DOEpatents

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

Scott, Charles D. (Oak Ridge, TN); Marasco, Joseph A. (Kingston, TN)

1996-01-01

110

Biparticle fluidized bed reactor  

DOEpatents

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

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

1995-04-25

111

Licensed reactor nuclear safety criteria applicable to DOE reactors  

SciTech Connect

The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC (Nuclear Regulatory Commission) licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor.

Not Available

1991-04-01

112

POWER FROM NA COOLED REACTOR REACTOR DESIGN AND FEASIBILITY PROBLEM  

Microsoft Academic Search

Work done on the study of a Na cooled, graphite moderated nuclear ; reactor for power production is summarized. The preliminary design calculations ; and results are described for a heterogeneous, thermal reactor using rod type ; fuel elements of slightly enriched U. The total heat capacity is 300,000 kw with ; a nominal electric power producing capacity of 75,000

D. R. Bennion; R. L. Hickmott; N. A. Krohn; W. J. McCarthy; R. A. Moore; H. J. Stumpf; L. A. Waldman

1952-01-01

113

Fast Reactor Fuel Type and Reactor Safety Performance  

SciTech Connect

Fast Reactor Fuel Type and Reactor Safety Performance R. Wigeland , Idaho National Laboratory J. Cahalan, Argonne National Laboratory The sodium-cooled fast neutron reactor is currently being evaluated for the efficient transmutation of the highly-hazardous, long-lived, transuranic elements that are present in spent nuclear fuel. One of the fundamental choices that will be made is the selection of the fuel type for the fast reactor, whether oxide, metal, carbide, nitride, etc. It is likely that a decision on the fuel type will need to be made before many of the related technologies and facilities can be selected, from fuel fabrication to spent fuel reprocessing. A decision on fuel type should consider all impacts on the fast reactor system, including safety. Past work has demonstrated that the choice of fuel type may have a significant impact on the severity of consequences arising from accidents, especially for severe accidents of low probability. In this paper, the response of sodium-cooled fast reactors is discussed for both oxide and metal fuel types, highlighting the similarities and differences in reactor response and accident consequences. Any fast reactor facility must be designed to be able to successfully prevent, mitigate, or accommodate all consequences of potential events, including accidents. This is typically accomplished by using multiple barriers to the release of radiation, including the cladding on the fuel, the intact primary cooling system, and most visibly the reactor containment building. More recently, this has also included the use of ‘inherent safety’ concepts to reduce or eliminate the potential for serious damage in some cases. Past experience with oxide and metal fuel has demonstrated that both fuel types are suitable for use as fuel in a sodium-cooled fast reactor. However, safety analyses for these two fuel types have also shown that there can be substantial differences in accident consequences due to the neutronic and thermophysical properties of the fuel and their compatibility with the reactor coolant, with corresponding differences in the challenges presented to the reactor developers. Accident phenomena are discussed for the sodium-cooled fast reactor based on the mechanistic progression of conditions from accident initiation to accident termination, whether a benign state is achieved or more severe consequences are expected. General principles connecting accident phenomena and fuel properties are developed from the oxide and metal fuel safety analyses, providing guidelines that can be used as part of the evaluation for selection of fuel type for the sodium-cooled fast reactor.

R. Wigeland; J. Cahalan

2009-09-01

114

Fluidized bed nuclear fission reactor  

Microsoft Academic Search

For the further development of nuclear fission reactors, the feasibility of a new concept is evaluated. It concerns a fluidized bed reactor in which carbon particles with a uranium core are fluidized and cooled by a high velocity pressurized helium flow. Nuclear reaction takes place if the bed is in fluidized conditions at a void fraction above 80% and it

T. H. J. J. Van Der Hagen; H. Van Dam

1996-01-01

115

Fast mixed spectrum reactor concept  

Microsoft Academic Search

The Fast Mixed Spectrum Reactor is a highly promising concept for a fast reactor with improved features of proliferation resistance, and excellent utilization of uranium resources. In technology, it can be considered to be a branch of fast breeder development, though its operation and implications are different from those of FBR'S in important respects. Successful development programs are required in

H. J. C. Kouts; G. J. Fischer; R. J. Cerbone

1979-01-01

116

Proton Collimators for Fusion Reactors  

NASA Technical Reports Server (NTRS)

Proton collimators have been proposed for incorporation into inertial-electrostatic-confinement (IEC) fusion reactors. Such reactors have been envisioned as thrusters and sources of electric power for spacecraft and as sources of energetic protons in commercial ion-beam applications.

Miley, George H.; Momota, Hiromu

2003-01-01

117

KINETIC BEHAVIOUR OF NUCLEAR REACTORS  

Microsoft Academic Search

Nuclear reactors are described in terms of their physical behavior in ; various regimes of complexity, corresponding to increasing power. The persisting ; distribution is given special attention since it is the consequence of the ; tendency of reactors to behave as a single entity. It is pointed out that in the ; most general case there is no unique

Lewins

1962-01-01

118

Neutrino Oscillations with Reactor Neutrinos  

E-print Network

Prospect measurements of neutrino oscillations with reactor neutrinos are reviewed in this document. The following items are described: neutrinos oscillations status, reactor neutrino experimental strategy, impact of uncertainties on the neutrino oscillation sensitivity and, finally, the experiments in the field. This is the synthesis of the talk delivered during the NOW2006 conference at Otranto (Italy) during September 2006.

Anatael Cabrera

2007-01-11

119

CONTROL RODS FOR NUCLEAR REACTORS  

Microsoft Academic Search

A means for controlling the control rod in emergency, when it is desired ; to shutdown the reactor with the shortest possible delay, is described. When the ; emergency occurs the control rod is allowed to drop freely under gravity from the ; control rod support tube into the bore in the reactor core. A normal shutdown is ; reached

Bell

1963-01-01

120

Radioactive wastes from fusion reactors.  

PubMed

Calculation of the amount of tritium released from a hypothetical fusion reactor shows that it is 2 x 10(5) the amount released by generation of an equivalent amount of electricity by a fission reactor. Release of the tritium generated by a power economy, if the nuclear power were all fusion, would result in unacceptable worldwide dosages by the year 2000. PMID:17737473

Parkfr, F L

1968-01-01

121

Radiation Shielding for Fusion Reactors  

SciTech Connect

Radiation shielding requirements for fusion reactors present different problems than those for fission reactors and accelerators. Fusion devices, particularly tokamak reactors, are complicated by geometry constraints that complicate disposition of fully effective shielding. This paper reviews some of these shielding issues and suggested solutions for optimizing the machine and biological shielding. Radiation transport calculations are essential for predicting and confirming the nuclear performance of the reactor and, as such, must be an essential part of the reactor design process. Development and optimization of reactor components from the first wall and primary shielding to the penetrations and containment shielding must be carried out in a sensible progression. Initial results from one-dimensional transport calculations are used for scoping studies and are followed by detailed two- and three-dimensional analyses to effectively characterize the overall radiation environment. These detail model calculations are essential for accounting for the radiation leakage through ports and other penetrations in the bulk shield. Careful analysis of component activation and radiation damage is cardinal for defining remote handling requirements, in-situ replacement of components, and personnel access at specific locations inside the reactor containment vessel. Radiation shielding requirements for fusion reactors present different problems than those for fission reactors and accelerators. Fusion devices, particularly tokamak reactors, are complicated by geometry constraints that complicate disposition of fully effective shielding. This paper reviews some of these shielding issues and suggested solutions for optimizing the machine and biological shielding. Radiation transport calculations are essential for predicting and confirming the nuclear performance of the reactor and, as such, must be an essential part of the reactor design process. Development and optimization of reactor components from the first wall and primary shielding to the penetrations and containment shielding must be carried out in a sensible progression. Initial results from one-dimensional transport calculations are used for scoping studies and are followed by detailed two- and three-dimensional analyses to effectively characterize the overall radiation environment. These detail model calculations are essential for accounting for the radiation leakage through ports and other penetrations in the bulk shield. Careful analysis of component activation and radiation damage is cardinal for defining remote handling requirements, in-situ replacement of components, and personnel access at specific locations inside the reactor containment vessel.

Santoro, R.T.

1999-10-01

122

Solvent refined coal reactor quench system  

DOEpatents

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

Thorogood, Robert M. (Macungie, PA)

1983-01-01

123

Solvent refined coal reactor quench system  

DOEpatents

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

Thorogood, R.M.

1983-11-08

124

Reactor for making uniform capsules  

NASA Technical Reports Server (NTRS)

The present invention provides a novel reactor for making capsules with uniform membrane. The reactor includes a source for providing a continuous flow of a first liquid through the reactor; a source for delivering a steady stream of drops of a second liquid to the entrance of the reactor; a main tube portion having at least one loop, and an exit opening, where the exit opening is at a height substantially equal to the entrance. In addition, a method for using the novel reactor is provided. This method involves providing a continuous stream of a first liquid; introducing uniformly-sized drops of the second liquid into the stream of the first liquid; allowing the drops to react in the stream for a pre-determined period of time; and collecting the capsules.

Wang, Taylor G. (Inventor); Anikumar, Amrutur V. (Inventor); Lacik, Igor (Inventor)

1999-01-01

125

Fast reactors and nuclear nonproliferation  

SciTech Connect

Problems are discussed with regard to nuclear fuel cycle resistance in fast reactors to nuclear proliferation risk due to the potential for use in military programs of the knowledge, technologies and materials gained from peaceful nuclear power applications. Advantages are addressed for fast reactors in the creation of a more reliable mode of nonproliferation in the closed nuclear fuel cycle in comparison with the existing fully open and partially closed fuel cycles of thermal reactors. Advantages and shortcomings are also discussed from the point of view of nonproliferation from the start with fast reactors using plutonium of thermal reactor spent fuel and enriched uranium fuel to the gradual transition using their own plutonium as fuel. (authors)

Avrorin, E.N. [Russian Federal Nuclear Center - Zababakhin Institute of Applied Physics, Snezhinsk (Russian Federation); Rachkov, V.I.; Chebeskov, A.N. [State Scientific Center of the Russian Federation - Institute for Physics and Power Engineering, Bondarenko Square, 1, Obninsk, Kaluga region, 249033 (Russian Federation)

2013-07-01

126

Advanced Reactors Around the World  

SciTech Connect

At the end of 2002, 441 nuclear power plants were operating around the globe and providing 17% of the world's electricity. Although the rate of population growth has slowed, recent United Nations data suggest that two billion more people will be added to the world by 2050. A special report commissioned by the Intergovernmental Panel on Climate Change estimated that electricity demand would grow almost eight-fold from 2000 to 2050 in a high economic grown scenario and more than double in a low-growth scenario. There is also a global aspiration to keep the environment pristine. Because of these reasons, it is expected that a large number of new nuclear reactors may be operating by 2050. Realization of this has created an impetus for the development of a new generation of reactors in several countries. The goal is to make nuclear power cost-competitive with other resources and to enhance safety to a level that no evacuation outside a plant site would be necessary. It should also generate less waste, prevent materials diversion for weapons production, and be sustainable. This article discusses the status of next-generation reactors under development around the world. Specifically highlighted are efforts related to the Generation IV International Forum (GIF) and its six reactor concepts for research and development: Very High Temperature Reactor (VHTR); Gas-Cooled Fast Reactor (GFR); Supercritical Water-Cooled Reactor (SCWR); Sodium-Cooled Fast Reactor (SFR); Lead-Cooled Fast Reactor (LFR); and Molten Salt Reactor (MSR). Also highlighted are nuclear activities specific to Russia and India.

Majumdar, Debu

2003-09-01

127

Nanoliter Reactors Improve Multiple Displacement Amplification  

E-print Network

Nanoliter Reactors Improve Multiple Displacement Amplification of Genomes from Single Cells Yann) Nanoliter reactors improve multiple displacement amplification of genomes from single cells. PLoS Genet 3

Quake, Stephen R.

128

Nuclear reactor control  

DOEpatents

1. In a nuclear reactor incorporating a plurality of columns of tubular fuel elements disposed in horizontal tubes in a mass of graphite wherein water flows through the tubes to cool the fuel elements, the improvement comprising at least one control column disposed in a horizontal tube including fewer fuel elements than in a normal column of fuel elements and tubular control elements disposed at both ends of said control column, and means for varying the horizontal displacement of the control column comprising a winch at the upstream end of the control column and a cable extending through the fuel and control elements and attached to the element at the downstream end of the column.

Cawley, William E. (Phoenix, AZ); Warnick, Robert F. (Pasco, WA)

1982-01-01

129

University Reactor Sharing Program  

SciTech Connect

Research projects supported by the program include items such as dating geological material and producing high current super conducting magnets. The funding continues to give small colleges and universities the valuable opportunity to use the NSC for teaching courses in nuclear processes; specifically neutron activation analysis and gamma spectroscopy. The Reactor Sharing Program has supported the construction of a Fast Neutron Flux Irradiator for users at New Mexico Institute of Mining and Technology and the University of Houston. This device has been characterized and has been found to have near optimum neutron fluxes for A39/Ar 40 dating. Institution final reports and publications resulting from the use of these funds are on file at the Nuclear Science Center.

W.D. Reese

2004-02-24

130

Licensed reactor nuclear safety criteria applicable to DOE reactors  

SciTech Connect

This document is a compilation and source list of nuclear safety criteria that the Nuclear Regulatory Commission (NRC) applies to licensed reactors; it can be used by DOE and DOE contractors to identify NRC criteria to be evaluated for application to the DOE reactors under their cognizance. The criteria listed are those that are applied to the areas of nuclear safety addressed in the safety analysis report of a licensed reactor. They are derived from federal regulations, USNRC regulatory guides, Standard Review Plan (SRP) branch technical positions and appendices, and industry codes and standards.

Not Available

1993-11-01

131

Advances in reactor physics education: Visualization of reactor parameters  

SciTech Connect

Modern computer codes allow detailed neutron transport calculations. In combination with advanced 3D visualization software capable of treating large amounts of data in real time they form a powerful tool that can be used as a convenient modern educational tool for reactor operators, nuclear engineers, students and specialists involved in reactor operation and design. Visualization is applicable not only in education and training, but also as a tool for fuel management, core analysis and irradiation planning. The paper treats the visualization of neutron transport in different moderators, neutron flux and power distributions in two nuclear reactors (TRIGA type research reactor and a typical PWR). The distributions are calculated with MCNP and CORD-2 computer codes and presented using Amira software. (authors)

Snoj, L.; Kromar, M.; Zerovnik, G. [Josef Stefan Inst., Jamova cesta 39, SI-1000 Ljubljana (Slovenia)

2012-07-01

132

Chemistry in water reactors. Reserapport. (Chemistry in water reactors).  

National Technical Information Service (NTIS)

The international conference Chemistry in Water Reactors was arranged in Nice 24-27/04/1994 by the French Nuclear Energy Society. Examples of technical program areas were primary chemistry, operational experience, fundamental studies and new technology. F...

H. P. Hermansson, K. Norring

1994-01-01

133

Reactor physics design of supercritical CO?-cooled fast reactors  

E-print Network

Gas-Cooled Fast Reactors (GFRs) are among the GEN-IV designs proposed for future deployment. Driven by anticipated plant cost reduction, the use of supercritical CO? (S-CO?) as a Brayton cycle working fluid in a direct ...

Pope, Michael A. (Michael Alexander)

2004-01-01

134

Reactor protection system design alternatives for sodium fast reactors  

E-print Network

Historically, unprotected transients have been viewed as design basis events that can significantly challenge sodium-cooled fast reactors. The perceived potential consequences of a severe unprotected transient in a ...

DeWitte, Jacob D. (Jacob Dominic)

2011-01-01

135

Directions for improved fusion reactors  

SciTech Connect

Conceptual fusion reactor studies over the past 10 to 15 years have projected systems that may be too large, complex, and costly to be of commercial interest. One main direction for improved fusion reactors points towards smaller, higher-power-density approaches. First-order economic issues (i.e., unit direct cost and cost of electricity) are used to support the need for more compact fusion reactors. A generic fusion physics/engineering/costing model is used to provide a quantiative basis for these arguments for specific fusion concepts.

Krakowski, R.A.; Miller, R.L.; Delene, J.G.

1986-01-01

136

LOCA Analysis of Super Fast Reactor  

Microsoft Academic Search

This paper describes loss of coolant accident (LOCA) analyses of the Supercritical-pressure Water-Cooled Fast Reactor (Super Fast Reactor). The features of the Super Fast Reactor are high power density and downward flow cooled fuel channels for the improvement of the economic potential of the Super Fast Reactor with high outlet steam temperature. The LOCA induces large pressure and coolant density

Satoshi IKEJIRI; Chi Young HAN; Yuki ISHIWATARI; Yoshiaki OKA

2011-01-01

137

REACTOR SAFETY, HAZARDS EVALUATION AND INSPECTION  

Microsoft Academic Search

In the prevention of release and dispersal of radioactive fission ; products, the major factors are technical knowledge of reactor characteristics, ; inherent safety features or lack thereof in reactor design, and supervision and ; control of reactor operation. In recognition of the present incomplete state of ; technical knowledge and of the serious potential hazard from a reactor accident,

C. K. Beck; M. M. Mann; P. A. Morris

1958-01-01

138

Jet-Stirred Reactors Olivier Herbinet1  

E-print Network

1 Chapter 8 Jet-Stirred Reactors Olivier Herbinet1 , Guillaume Dayma2 Abstract The jet-stirred reactor is a type of ideal continuously stirred-tank reactor which is well suited for gas phase kinetic-stirred reactor to observe new types of species and to gain accuracy in the identification and the quantification

Paris-Sud XI, Université de

139

Unique features of space reactors  

NASA Astrophysics Data System (ADS)

Space reactors are designed to meet a unique set of requirements; they must be sufficiently compact to be launched in a rocket to their operational location, operate for many years without maintenance and servicing, operate in extreme environments, and reject heat by radiation to space. To meet these restrictions, operating temperatures are much greater than in terrestrial power plants, and the reactors tend to have a fast neutron spectrum. Currently, a new generation of space reactor power plants is being developed. The major effort is in the SP-100 program, where the power plant is being designed for seven years of full power, and no maintenance operation at a reactor outlet operating temperature of 1350 K.

Buden, David

140

Advanced Catalytic Hydrogenation Retrofit Reactor  

SciTech Connect

Industrial hydrogenation is often performed using a slurry catalyst in large stirred-tank reactors. These systems are inherently problematic in a number of areas, including industrial hygiene, process safety, environmental contamination, waste production, process operability and productivity. This program proposed the development of a practical replacement for the slurry catalysts using a novel fixed-bed monolith catalyst reactor, which could be retrofitted onto an existing stirred-tank reactor and would mitigate many of the minitations and problems associated with slurry catalysts. The full retrofit monolith system, consisting of a recirculation pump, gas/liquid ejector and monolith catalyst, is described as a monolith loop reactor or MLR. The MLR technology can reduce waste and increase raw material efficiency, which reduces the overall energy required to produce specialty and fine chemicals.

Reinaldo M. Machado

2002-08-15

141

Combustion synthesis continuous flow reactor  

DOEpatents

The present invention is a reactor for combustion synthesis of inorganic powders. The reactor includes a reaction vessel having a length and a first end and a second end. The reaction vessel further has a solution inlet and a carrier gas inlet. The reactor further has a heater for heating both the solution and the carrier gas. In a preferred embodiment, the reaction vessel is heated and the solution is in contact with the heated reaction vessel. It is further preferred that the reaction vessel be cylindrical and that the carrier gas is introduced tangentially into the reaction vessel so that the solution flows helically along the interior wall of the reaction vessel. As the solution evaporates and combustion produces inorganic material powder, the carrier gas entrains the powder and carries it out of the reactor.

Maupin, Gary D. (Richland, WA); Chick, Lawrence A. (West Richland, WA); Kurosky, Randal P. (Maple Valley, WA)

1998-01-01

142

Reactor operation environmental information document  

SciTech Connect

This volume is a reactor operation environmental information document for the Savannah River Plant. Topics include meteorology, surface hydrology, transport, environmental impacts, and radiation effects. 48 figs., 56 tabs. (KD)

Bauer, L.R.; Hayes, D.W.; Hunter, C.H.; Marter, W.L.; Moyer, R.A.

1989-12-01

143

Interfacial effects in fast reactors  

E-print Network

The problem of increased resonance capture rates near zone interfaces in fast reactor media has been examined both theoretically and experimentally. An interface traversing assembly was designed, constructed and employed ...

Saidi, Mohammad Said

1979-01-01

144

Reactor physics project final report  

E-print Network

This is the final report in an experimental and theoretical program to develop and apply single- and few-element methods for the determination of reactor lattice parameters. The period covered by the report is January 1, ...

Driscoll, Michael J.

1970-01-01

145

Overview of fusion reactor safety  

NASA Astrophysics Data System (ADS)

Use of deuterium-tritium fusion reactors requires examination of several major safety and environmental issues: (1) tritium inventory control; (2) neutron activation of structural materials, fluid streams and reactor hall environment; (3) release of radioactivity from energy sources including lithium spill reactions, superconducting magnet stored energy release, and plasma disruptions; (4) high magnetic and electromagnetic fields associated with fusion reactor superconducting magnets and radio frequency heating devices; and (5) handling and disposal of radioactive waste. Early recognition of potential safety problems with fusion reactors provides the opportunity for improvement in design and materials to eliminate or greatly reduce these problems. With an early start in this endeavor, fusion should be among the lower risk technologies for generation of commercial electrical power.

Cohen, S.; Crocker, J. G.

146

Reactor core isolation cooling system  

DOEpatents

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

Cooke, F.E.

1992-12-08

147

Nuclear Reactors and Technology; (USA)  

SciTech Connect

Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database (EDB) during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency's Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on EDB and Nuclear Science Abstracts (NSA) database. Current information, added daily to EDB, is available to DOE and its contractors through the DOE integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user's needs.

Cason, D.L.; Hicks, S.C. (eds.)

1991-01-01

148

Solid State Reactor Final Report  

SciTech Connect

The Solid State Reactor (SSR) is an advanced reactor concept designed to take advantage of Oak Ridge National Laboratory's (ORNL's) recently developed graphite foam that has enhanced heat transfer characteristics and excellent high-temperature mechanical properties, to provide an inherently safe, self-regulated, source of heat for power and other potential applications. This work was funded by the U.S. Department of Energy's Nuclear Energy Research Initiative (NERI) program (Project No. 99-064) from August 1999 through September 30, 2002. The initial concept of utilizing the graphite foam as a basis for developing an advanced reactor concept envisioned that a suite of reactor configurations and power levels could be developed for several different applications. The initial focus was looking at the reactor as a heat source that was scalable, independent of any heat removal/power conversion process. These applications might include conventional power generation, isotope production and destruction (actinides), and hydrogen production. Having conducted the initial research on the graphite foam and having performed the scoping parametric analyses from neutronics and thermal-hydraulic perspectives, it was necessary to focus on a particular application that would (1) demonstrate the viability of the overall concept and (2) require a reasonably structured design analysis process that would synthesize those important parameters that influence the concept the most as part of a feasible, working reactor system. Thus, the application targeted for this concept was supplying power for remote/harsh environments and a design that was easily deployable, simplistic from an operational standpoint, and utilized the new graphite foam. Specifically, a 500-kW(t) reactor concept was pursued that is naturally load following, inherently safe, optimized via neutronic studies to achieve near-zero reactivity change with burnup, and proliferation resistant. These four major areas of research were undertaken: (1) establishing the design and safety-related basis via neutronic and reactor control assessments with the graphite foam as heat transfer medium; (2) evaluating the thermal performance of the graphite foam for heat removal, reactor stability, reactor operations, and overall core thermal characteristics; (3) characterizing the physical properties of the graphite foam under normal and irradiated conditions to determine any effects on structure, dimensional stability, thermal conductivity, and thermal expansion; and (4) developing a power conversion system design to match the reactor operating parameters.

Mays, G.T.

2004-03-10

149

Reactor shroud joint  

DOEpatents

A shroud for a nuclear reactor is described. In one embodiment, the shroud includes first and second shroud sections, and each shroud section includes a substantially cylindrical main body having a first end and a second end. With respect to each shroud section, a flange is located at the main body first end, and the flange has a plurality of bolt openings therein and a plurality of scalloped regions. The first shroud section is welded to the second shroud section, and at least some of the bolt openings in the first shroud section flange align with respective bolt openings in the second shroud section flange. In the event that the onset of inter-granular stress corrosion cracking is ever detected in the weld between the shroud section, bolts are inserted through bolt openings in the first shroud section flange and through aligned bolt openings the second shroud section flange. Each bolt, in one embodiment, has a shank section and first and second threaded end sections. Nuts are threadedly engaged to the threaded end sections and tightened against the respective flanges. 4 figs.

Ballas, G.J.; Fife, A.B.; Ganz, I.

1998-04-07

150

Reactor shroud joint  

DOEpatents

A shroud for a nuclear reactor is described. In one embodiment, the shroud includes first and second shroud sections, and each shroud section includes a substantially cylindrical main body having a first end and a second end. With respect to each shroud section, a flange is located at the main body first end, and the flange has a plurality of bolt openings therein and a plurality of scalloped regions. The first shroud section is welded to the second shroud section, and at least some of the bolt openings in the first shroud section flange align with respective bolt openings in the second shroud section flange. In the event that the onset of inter-granular stress corrosion cracking is ever detected in the weld between the shroud section, bolts are inserted through bolt openings in the first shroud section flange and through aligned bolt openings the second shroud section flange. Each bolt, in one embodiment, has a shank section and first and second threaded end sections. Nuts are threadedly engaged to the threaded end sections and tightened against the respective flanges.

Ballas, Gary J. (San Jose, CA); Fife, Alex Blair (San Jose, CA); Ganz, Israel (San Jose, CA)

1998-01-01

151

Reactor Simulator Testing  

NASA Technical Reports Server (NTRS)

As part of the Nuclear Systems Office Fission Surface Power Technology Demonstration Unit (TDU) project, a reactor simulator test loop (RxSim) was design & built to perform integrated testing of the TDU components. In particular, the objectives of RxSim testing was to verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. In addition, it was decided to include a thermal test of a cold trap purification design and a pump performance test at pump voltages up to 150 V since the targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V. This paper summarizes RxSim testing. The gas and vacuum ground support test equipment performed effectively in NaK fill, loop pressurization, and NaK drain operations. The instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. The cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained which was lower than the predicted 750 K but 156 K higher than the cold temperature indicating the design provided some heat regeneration. The annular linear induction pump (ALIP) tested was able to produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

Schoenfeld, Michael P.; Webster, Kenny L.; Pearson, Boise J.

2013-01-01

152

Reactor pressure vessel nozzle  

DOEpatents

A nozzle for joining a pool of water to a nuclear reactor pressure vessel includes a tubular body having a proximal end joinable to the pressure vessel and a distal end joinable in flow communication with the pool. The body includes a flow passage therethrough having in serial flow communication a first port at the distal end, a throat spaced axially from the first port, a conical channel extending axially from the throat, and a second port at the proximal end which is joinable in flow communication with the pressure vessel. The inner diameter of the flow passage decreases from the first port to the throat and then increases along the conical channel to the second port. In this way, the conical channel acts as a diverging channel or diffuser in the forward flow direction from the first port to the second port for recovering pressure due to the flow restriction provided by the throat. In the backflow direction from the second port to the first port, the conical channel is a converging channel and with the abrupt increase in flow area from the throat to the first port collectively increase resistance to flow therethrough. 2 figs.

Challberg, R.C.; Upton, H.A.

1994-10-04

153

Reactor pressure vessel nozzle  

DOEpatents

A nozzle for joining a pool of water to a nuclear reactor pressure vessel includes a tubular body having a proximal end joinable to the pressure vessel and a distal end joinable in flow communication with the pool. The body includes a flow passage therethrough having in serial flow communication a first port at the distal end, a throat spaced axially from the first port, a conical channel extending axially from the throat, and a second port at the proximal end which is joinable in flow communication with the pressure vessel. The inner diameter of the flow passage decreases from the first port to the throat and then increases along the conical channel to the second port. In this way, the conical channel acts as a diverging channel or diffuser in the forward flow direction from the first port to the second port for recovering pressure due to the flow restriction provided by the throat. In the backflow direction from the second port to the first port, the conical channel is a converging channel and with the abrupt increase in flow area from the throat to the first port collectively increase resistance to flow therethrough.

Challberg, Roy C. (Livermore, CA); Upton, Hubert A. (Morgan Hill, CA)

1994-01-01

154

Solar solids reactor  

DOEpatents

A solar powered kiln is provided, that is of relatively simple design and which efficiently uses solar energy. The kiln or solids reactor includes a stationary chamber with a rearward end which receives solid material to be reacted and a forward end through which reacted material is disposed of, and a screw conveyor extending along the bottom of the chamber for slowly advancing the material between the chamber ends. Concentrated solar energy is directed to an aperture at the forward end of the chamber to heat the solid material moving along the bottom of the chamber. The solar energy can be reflected from a mirror facing at an upward incline, through the aperture and against a heat-absorbing material near the top of the chamber, which moves towards the rear of the chamber to distribute heat throughout the chamber. Pumps at the forward and rearward ends of the chamber pump heated sweep gas through the length of the chamber, while minimizing the flow of gas through an open aperture through which concentrated sunlight is received.

Yudow, Bernard D. (Chicago, IL)

1987-01-01

155

Reactor Simulator Testing Overview  

NASA Technical Reports Server (NTRS)

Test Objectives Summary: a) Verify operation of the core simulator, the instrumentation & control system, and the ground support gas and vacuum test equipment. b) Examine cooling & heat regeneration performance of the cold trap purification. c) Test the ALIP pump at voltages beyond 120V to see if the targeted mass flow rate of 1.75 kg/s can be obtained in the RxSim. Testing Highlights: a) Gas and vacuum ground support test equipment performed effectively for operations (NaK fill, loop pressurization, and NaK drain). b) Instrumentation & Control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings and ramped within prescribed constraints. It effectively interacted with reactor simulator control model and defaulted back to temperature control mode if the transient fluctuations didn't dampen. c) Cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained which was lower than the predicted 750 K but 156 K higher than the minimum temperature indicating the design provided some heat regeneration. d) ALIP produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

Schoenfeld, Michael P.

2013-01-01

156

Novel Catalytic Membrane Reactors  

SciTech Connect

There are many industrial catalytic organic reversible reactions with amines or alcohols that have water as one of the products. Many of these reactions are homogeneously catalyzed. In all cases removal of water facilitates the reaction and produces more of the desired chemical product. By shifting the reaction to right we produce more chemical product with little or no additional capital investment. Many of these reactions can also relate to bioprocesses. Given the large number of water-organic compound separations achievable and the ability of the Compact Membrane Systems, Inc. (CMS) perfluoro membranes to withstand these harsh operating conditions, this is an ideal demonstration system for the water-of-reaction removal using a membrane reactor. Enhanced reaction synthesis is consistent with the DOE objective to lower the energy intensity of U.S. industry 25% by 2017 in accord with the Energy Policy Act of 2005 and to improve the United States manufacturing competitiveness. The objective of this program is to develop the platform technology for enhancing homogeneous catalytic chemical syntheses.

Stuart Nemser, PhD

2010-10-01

157

How far is a Fusion Power Reactor from an Experimental Reactor?  

E-print Network

1 How far is a Fusion Power Reactor from an Experimental Reactor? R. Toschi(1) , P. Barabaschi(2 September 2000 Madrid, Spain #12;2 How far is a fusion power reactor from an experimental reactor? R. Toschi. Recommendations to make a fusion power reactor an attractive source of electrical energy To inject some

158

Microchannel Reactors for ISRU Applications  

NASA Astrophysics Data System (ADS)

Affordable planning and execution of prolonged manned space missions depend upon the utilization of local resources and the waste products which are formed in manned spacecraft and surface bases. Successful in-situ resources utilization (ISRU) will require component technologies which provide optimal size, weight, volume, and power efficiency. Microchannel reactors enable the efficient chemical processing of in situ resources. The reactors can be designed for the processes that generate the most benefit for each mission. For instance, propellants (methane) can be produced from carbon dioxide from the Mars atmosphere using the Sabatier reaction and ethylene can be produced from the partial oxidation of methane. A system that synthesizes ethylene could be the precursor for systems to synthesize ethanol and polyethylene. Ethanol can be used as a nutrient for Astrobiology experiments, as well as the production of nutrients for human crew (e.g. sugars). Polyethylene can be used in the construction of habitats, tools, and replacement parts. This paper will present recent developments in miniature chemical reactors using advanced Micro Electro Mechanical Systems (MEMS) and microchannel technology to support ISRU of Mars and lunar missions. Among other applications, the technology has been demonstrated for the Sabatier process and for the partial oxidation of methane. Microchannel reactors were developed based on ceramic substrates as well as metal substrates. In both types of reactors, multiple layers coated with catalytic material are bonded, forming a monolithic structure. Such reactors are readily scalable with the incorporation of extra layers. In addition, this reactor structure minimizes pressure drop and catalyst settling, which are common problems in conventional packed bed reactors.

Carranza, Susana; Makel, Darby B.; Blizman, Brandon; Ward, Benjamin J.

2005-02-01

159

Automatic safety rod for reactors  

DOEpatents

An automatic safety rod for a nuclear reactor containing neutron absorbing material and designed to be inserted into a reactor core after a loss-of-core flow. Actuation is based upon either a sudden decrease in core pressure drop or the pressure drop decreases below a predetermined minimum value. The automatic control rod includes a pressure regulating device whereby a controlled decrease in operating pressure due to reduced coolant flow does not cause the rod to drop into the core.

Germer, John H. (San Jose, CA)

1988-01-01

160

Fast quench reactor and method  

DOEpatents

A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This "freezes" the desired end product(s) in the heated equilibrium reaction stage.

Detering, Brent A. (Idaho Falls, ID); Donaldson, Alan D. (Idaho Falls, ID); Fincke, James R. (Idaho Falls, ID); Kong, Peter C. (Idaho Falls, ID)

1998-01-01

161

Fast quench reactor and method  

DOEpatents

A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This "freezes" the desired end product(s) in the heated equilibrium reaction stage.

Detering, Brent A. (Idaho Falls, ID); Donaldson, Alan D. (Idaho Falls, ID); Fincke, James R. (Idaho Falls, ID); Kong, Peter C. (Idaho Falls, ID)

2002-01-01

162

Reactor vessel material surveillance program  

Microsoft Academic Search

B&W conducts surveillance programs for its 850-MW(e) class of reactors ; to determine the effects of neutron irradiation on the full Charpy V-notch curves ; of the reactor vessel materials. Charpy and tensile specimens, which are ; machined from as-fabricated metal, heat-affected-zone metal (HAZ), and weld ; metal, are placed in six surveillance capsules containing dosimeters and ; temperature monitors.

G. J. Snyder; G. S. Carter

1973-01-01

163

MARS: Mirror Advanced Reactor Study  

SciTech Connect

A recently completed two-year study of a commercial tandem mirror reactor design (Mirror Advanced Reactor Study (MARS)) is briefly reviewed. The end plugs are designed for trapped particle stability, MHD ballooning, balanced geodesic curvature, and small radial electric fields in the central cell. New technologies such as lithium-lead blankets, 24T hybrid coils, gridless direct converters and plasma halo vacuum pumps are highlighted.

Logan, B.G.

1984-09-10

164

Alternate-fuel reactor studies  

SciTech Connect

A number of studies related to improvements and/or greater understanding of alternate-fueled reactors is presented. These studies cover the areas of non-Maxwellian distributions, materials and lifetime analysis, a /sup 3/He-breeding blanket, tritium-rich startup effects, high field magnet support, and reactor operation spanning the range from full D-T operation to operation with no tritium breeding.

Evans, K. Jr.; Ehst, D.A.; Gohar, Y.; Jung, J.; Mattas, R.F.; Turner, L.R.

1983-02-01

165

When Do Commercial Reactors Permanently Shut Down?  

EIA Publications

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

2011-01-01

166

The nuclear reactor strategy between fast breeder reactors and advanced pressurized water reactors  

SciTech Connect

A nuclear reactor strategy between fast breeder reactors (FBRs) and advanced pressurized water reactors (APWRs) is being studied. The principal idea of this strategy is that the discharged plutonium from light water reactors (LWRs) provides the inventories of the FBRs and the high-converter APWRs, whereby the LWRs are installed according to the derivative of a logistical S curve. Special emphasis is given to the dynamics of reaching an asymptotic symbiosis between FBRs and APWRs. The main conclusion is that if a symbiotic APWR-FBR family with an asymptotic total power level in the terawatt range is to exist in about half a century from now, we need a large number of FBRs already in an early phase.

Seifritz, W.

1983-11-01

167

Reactor Simulator Testing  

NASA Technical Reports Server (NTRS)

As part of the Nuclear Systems Office Fission Surface Power Technology Demonstration Unit (TDU) project, a reactor simulator test loop (RxSim) was design & built to perform integrated testing of the TDU components. In particular, the objectives of RxSim testing was to verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. In addition, it was decided to include a thermal test of a cold trap purification design and a pump performance test at pump voltages up to 150 V since the targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V. This paper summarizes RxSim testing. The gas and vacuum ground support test equipment performed effectively in NaK fill, loop pressurization, and NaK drain operations. The instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. The cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained which was lower than the predicted 750 K but 156 K higher than the cold temperature indicating the design provided some heat regeneration. The annular linear induction pump (ALIP) tested was able to produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz. Keywords: fission, space power, nuclear, liquid metal, NaK.

Schoenfeld, Michael P.; Webster, Kenny L.; Pearson, Boise Jon

2013-01-01

168

(Liquid metal reactor/fast breeder reactor research and development)  

SciTech Connect

The second meeting of the UJCC was held in Japan on June 6--8, 1990. The first day was devoted to presentations of the status of the US and Japanese Fast Breeder Reactor (FBR) programs and the status of specific areas of cooperative work. Briefly, the Japanese are following the FBR development program which has been in place since the 1970s. This program includes an FBR test reactor (JOYO), a pilot-scale reactor (MONJU), a demonstration-scale plant, and commercial-scale plants by about 2020. The US program has been redirected toward an actinide recycle mission using metal fuel and pyroprocessing of spent fuel to recovery both Pu and the higher actinides for return to the Liquid Metal Reactor (LMR). The second day was spent traveling from Tokyo to Tsuruga for a tour of the MONJU reactor. The tour was especially interesting. The third day was spent writing the minutes of the meeting and the return trip to Tokyo.

Homan, F.J.

1990-06-20

169

56. ARAII. View inside reactor building looking at SL1 reactor ...  

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

56. ARA-II. View inside reactor building looking at SL-1 reactor vessel. November 19, 1957. Ineel photo no. 57-5864. Photographer: Jack L. Anderson. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

170

98. ARAIII. ML1 reactor pressure vessel is lowered into reactor ...  

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

98. ARA-III. ML-1 reactor pressure vessel is lowered into reactor pit by hoist. July 13, 1963. Ineel photo no. 63-4049. Photographer: Lowin. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

171

UCLA program in reactor studies: The ARIES tokamak reactor study  

SciTech Connect

The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Four ARIES visions are currently planned for the ARIES program. The ARIES-1 design is a DT-burning reactor based on modest'' extrapolations from the present tokamak physics database and relies on either existing technology or technology for which trends are already in place, often in programs outside fusion. ARIES-2 and ARIES-4 are DT-burning reactors which will employ potential advances in physics. The ARIES-2 and ARIES-4 designs employ the same plasma core but have two distinct fusion power core designs; ARIES-2 utilize the lithium as the coolant and breeder and vanadium alloys as the structural material while ARIES-4 utilizes helium is the coolant, solid tritium breeders, and SiC composite as the structural material. Lastly, the ARIES-3 is a conceptual D-{sup 3}He reactor. During the period Dec. 1, 1990 to Nov. 31, 1991, most of the ARIES activity has been directed toward completing the technical work for the ARIES-3 design and documenting the results and findings. We have also completed the documentation for the ARIES-1 design and presented the results in various meetings and conferences. During the last quarter, we have initiated the scoping phase for ARIES-2 and ARIES-4 designs.

Not Available

1991-01-01

172

Rapid starting methanol reactor system  

DOEpatents

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

Chludzinski, Paul J. (38 Berkshire St., Swampscott, MA 01907); Dantowitz, Philip (39 Nancy Ave., Peabody, MA 01960); McElroy, James F. (12 Old Cart Rd., Hamilton, MA 01936)

1984-01-01

173

International Research Reactor Decommissioning Project  

SciTech Connect

Many research reactors have been or will be shut down and are candidates for decommissioning. Most of the respective countries neither have a decommissioning policy nor the required expertise and funds to effectively implement a decommissioning project. The IAEA established the Research Reactor Decommissioning Demonstration Project (R{sup 2}D{sup 2}P) to help answer this need. It was agreed to involve the Philippine Research Reactor (PRR-1) as model reactor to demonstrate 'hands-on' experience as it is just starting the decommissioning process. Other facilities may be included in the project as they fit into the scope of R{sup 2}D{sup 2}P and complement to the PRR-1 decommissioning activities. The key outcome of the R{sup 2}D{sup 2}P will be the decommissioning of the PRR-1 reactor. On the way to this final goal the preparation of safety related documents (i.e., decommissioning plan, environmental impact assessment, safety analysis report, health and safety plan, cost estimate, etc.) and the licensing process as well as the actual dismantling activities could provide a model to other countries involved in the project. It is expected that the R{sup 2}D{sup 2}P would initiate activities related to planning and funding of decommissioning activities in the participating countries if that has not yet been done.

Leopando, Leonardo [Philippine Nuclear Research Institute, Quezon City (Philippines); Warnecke, Ernst [International Atomic Energy Agency, Vienna (Austria)

2008-01-15

174

Imaging Fukushima Daiichi reactors with muons  

NASA Astrophysics Data System (ADS)

A study of imaging the Fukushima Daiichi reactors with cosmic-ray muons to assess the damage to the reactors is presented. Muon scattering imaging has high sensitivity for detecting uranium fuel and debris even through thick concrete walls and a reactor pressure vessel. Technical demonstrations using a reactor mockup, detector radiation test at Fukushima Daiichi, and simulation studies have been carried out. These studies establish feasibility for the reactor imaging. A few months of measurement will reveal the spatial distribution of the reactor fuel. The muon scattering technique would be the best and probably the only way for Fukushima Daiichi to make this determination in the near future.

Miyadera, Haruo; Borozdin, Konstantin N.; Greene, Steve J.; Luki?, Zarija; Masuda, Koji; Milner, Edward C.; Morris, Christopher L.; Perry, John O.

2013-05-01

175

The Very High Temperature Reactor  

SciTech Connect

The High Temperature Reactor (HTR) and Very High Temperature Reactor (VHTR) are types of nuclear power plants that, as the names imply, operate at temperatures above those of the conventional nuclear power plants that currently generate electricity in the US and other countries. Like existing nuclear plants, heat generated from the fission of uranium or plutonium atoms is carried off by a working fluid and can be used generate electricity. The very hot working fluid also enables the VHTR to drive other industrial processes that require high temperatures not achievable by conventional nuclear plants (Figure 1). For this reason, the VHTR is being considered for non-electrical energy applications. The reactor and power conversion system are constructed using special materials that make a core meltdown virtually impossible.

Hans D. Gougar; David A. Petti

2011-06-01

176

Nuclear Reactor Engineering Analysis Laboratory  

SciTech Connect

The Nuclear Reactor Engineering Analysis Laboratory (NREAL) is a sophisticated computer system with state-of-the-art analytical tools and technology for analysis of light water reactors. Multiple application software tools can be activated to carry out different analyses and studies such as nuclear fuel reload evaluation, safety operation margin measurement, transient and severe accident analysis, nuclear reactor instability, operator training, normal and emergency procedures optimization, and human factors engineering studies. An advanced graphic interface, driven through touch-sensitive screens, provides the means to interact with specialized software and nuclear codes. The interface allows the visualization and control of all observable variables in a nuclear power plant (NPP), as well as a selected set of nonobservable or not directly controllable variables from conventional control panels.

Carlos Chavez-Mercado; Jaime B. Morales-Sandoval; Benjamin E. Zayas-Perez

1998-12-31

177

Light water reactor health physics.  

PubMed

In this article an overview of the historical development of light water reactor health physics programs is presented. Operational health physics programs have developed and matured as experience in operating and maintaining light water reactors has been gained. Initial programs grew quickly in both size and complexity with the number and size of nuclear units under construction and in operation. Operational health physics programs evolved to face various challenges confronted by the nuclear industry, increasing the effectiveness of radiological safety measures. Industry improvements in radiological safety performance have resulted in significant decreases in annual collective exposures from a high value of 790 person-rem in 1980 to 117 person-rem per reactor in 2002. Though significant gains have been made, the continued viability of the nuclear power industry is confronted with an aging workforce, as well as the challenges posed by deregulation and the need to maintain operational excellence. PMID:15891460

Prince, Robert J; Bradley, Scott E

2005-06-01

178

Light water reactor health physics.  

PubMed

In this article an overview of the historical development of light water reactor health physics programs is presented. Operational health physics programs have developed and matured as experience in operating and maintaining light water reactors has been gained. Initial programs grew quickly in both size and complexity with the number and size of nuclear units under construction and in operation. Operational health physics programs evolved to face various challenges confronted by the nuclear industry, increasing the effectiveness of radiological safety measures. Industry improvements in radiological safety performance have resulted in significant decreases in annual collective exposures from a high value of 790 person-rem in 1980 to 117 person-rem per reactor in 2002. Though significant gains have been made, the continued viability of the nuclear power industry is confronted with an aging workforce, as well as the challenges posed by deregulation and the need to maintain operational excellence. PMID:15551785

Prince, Robert J; Bradley, Scott E

2004-11-01

179

Reactor control rod timing system  

DOEpatents

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

Wu, Peter T. K. (Clifton Park, NY)

1982-01-01

180

Plasma reactor waste management systems  

NASA Technical Reports Server (NTRS)

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

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

1992-01-01

181

Space reactor preliminary mechanical design  

SciTech Connect

An analysis was performed on the SABRE reactor space power system to determine the effect of the number and size of heat pipes on the design parameters of the nuclear subsystem. Small numbers of thin walled heat pipes were found to give a lower subsystem mass, but excessive fuel swelling resulted. The SP-100 preliminary design uses 120 heat pipes because of acceptable fuel swelling and a minimum nuclear subsystem mass of 1875 kg. Salient features of the reactor preliminary design are: individual fuel modules, ZrO/sub 2/ block core mounts, bolted collar fuel module restraints, and a BeO central plug.

Meier, K.L.

1983-01-01

182

SP-100 space reactor safety  

SciTech Connect

The SP-100 space reactor power system is being developed to meet the large electrical power requirements of civilian and military missions planned for the 1990's and beyond. It will remove the restrictions on electrical power generation that have tended to limit missions and will enable the fuller exploration and utilization of space. This booklet describes the SP-100 space reactor power system and its development. Particular emphasis is given to safety. The design aand operational features as well as the design and safety review process that will assure that the SP-100 can be launched nd operated safely are described.

Not Available

1987-05-01

183

(International Thermonuclear Experimental Reactor support)  

SciTech Connect

This report summarizes the activities under LLNL Purchase Order B089367, the purpose of which is to support the University/Lawrence Livermore National Laboratory Magnetic Fusion Program by evaluating the status of research relative to other national and international programs and assist in long-range plans and development strategies for magnetic fusion in general and for ITER in particular.'' Two specific subtasks are included: to review the LLNL Magnet Technology Development Program in the context of the International Thermonuclear Experimental Reactor Design Study'' and to assist LLNL to organize and prepare materials for an International Thermonuclear Experimental Reactor Design Study information meeting.''

Dean, S.O.

1990-10-15

184

From CANDLE reactor to pebble-bed reactor  

SciTech Connect

This paper attempts to reveal theoretically, by studying a diffusion-burn-up coupled neutronic model, that a so-called CANDLE reactor and a pebble-bed type reactor have a common burn-up feature. As already known, a solitary burn-up wave that can develop in the common U-Pu and Th-U conversion processes is the basic mechanism of the CANDLE reactor. In this paper it is demonstrated that a family of burn-up wave solution exists in the boundary value problem characterizing a pebble bed reactor, in which the fuel is loaded from above into the core and unloaded from bottom. Among this solution family there is a particular case, namely, a partial solitary wave solution, which begins from the fuel entrance side and extends into infinity on the exit side, and has a maximal bum-up rate in this family. An example dealing with the {sup 232}Th-{sup 233}U conversion chain is studied and the solutions are presented in order to show the mechanism of the burn-up wave. (authors)

Chen, X. N.; Maschek, W. [Inst. for Nuclear and Energy Technologies, Forschungszentrum Karlsruhe, P.O.B. 3640, D-76021 Karlsruhe (Germany)

2006-07-01

185

FLUIDIZED BED REACTOR STUDY. Reactor Design and Feasibility Study  

Microsoft Academic Search

A feasibility study of a fluidized bed reactor system was undertaken. ; Initial studies were conducted on a fast, sodium cooled system, a gas cooled ; system, and organic and water moderated thermal systems. The organic moderated ; and light water moderated systems showed promise. The light water system was ; chosen for a detailed study to enable comparison with

C. L. Teeter; T. Ciarlariello; B. L. Hoffman; D. H. Jorgensen; F. D. Judge; L. J. King; D. A. McCune; N. R. Scheve; H. E. Zellnik

1957-01-01

186

(Safety related reactor physics calculation for HTGR type reactors)  

SciTech Connect

To address certain needs for validation data for gas-cooled reactors, a series of criticals are being planned at the PROTEUS facility at the Paul Scherrer Institute (PSI) in Wuerenlingen, Switzerland. The International Atomic Energy Agency is establishing a Coordinated Research Program (CRP) to provide a means for interested member countries to participate in the PROTEUS review.

Cleveland, J.C.

1989-10-21

187

Summary of advanced LMR (Liquid Metal Reactor) evaluations: PRISM (Power Reactor Inherently Safe Module) and SAFR (Sodium Advanced Fast Reactor)  

Microsoft Academic Search

In support of the US Nuclear Regulatory Commission (NRC), Brookhaven National Laboratory (BNL) has performed independent analyses of two advanced Liquid Metal Reactor (LMR) concepts. The designs, sponsored by the US Department of Energy (DOE), the Power Reactor Inherently Safe Module (PRISM) (Berglund, 1987) and the Sodium Advanced Fast Reactor (SAFR) (Baumeister, 1987), were developed primarily by General Electric (GE)

G. J. Van Tuyle; G. C. Slovik; B. C. Chan; R. J. Kennett; H. S. Cheng; P. G. Kroeger

1989-01-01

188

Fuel Systems for Compact Fast Space Reactors.  

National Technical Information Service (NTIS)

About 200 refractory metal clad ceramic fuel pins have been irradiated in thermal reactors under the 1200 K to 1550 K cladding temperature conditions of primary relevance to space reactors. This paper reviews performance with respect to fissile atom densi...

C. M. Cox, D. S. Dutt, R. A. Karnesky

1983-01-01

189

Development of high temperature catalytic membrane reactors.  

National Technical Information Service (NTIS)

Significant progress was made in 1991 on the development of ceramic membranes as catalytic reactors. Efforts were focused on the design, construction and startup of a reactor system capable of duplicating relevant commercial operating conditions. With thi...

1992-01-01

190

Stability analysis of supercritical water cooled reactors  

E-print Network

The Supercritical Water-Cooled Reactor (SCWR) is a concept for an advanced reactor that will operate at high pressure (25MPa) and high temperature (500°C average core exit). The high coolant temperature as it leaves the ...

Zhao, Jiyun, Ph. D. Massachusetts Institute of Technology

2005-01-01

191

RACEWAY REACTOR FOR MICROALGAL BIODIESEL PRODUCTION  

EPA Science Inventory

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

192

Microfluidic reactors for the synthesis of nanocrystals  

E-print Network

Several microfluidic reactors were designed and applied to the synthesis of colloidal semiconductor nanocrystals (NCs). Initially, a simple single-phase capillary reactor was used for the synthesis of CdSe NCs. Precursors ...

Yen, Brian K. H

2007-01-01

193

International Forum for Reactor Aging Management (IFRAM)  

SciTech Connect

The Nuclear Regulatory Commission has undertaken a program to lay the groundwork for defining proactive actions to manage degradation of materials in light water reactors (LWRs). This article discusses the international forum for reactor aging management.

Bond, Leonard J.

2010-11-01

194

Request for Naval Reactors Comment on Proposed Prometheus Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to JPL  

SciTech Connect

The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophy on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory.

D. Kokkinos

2005-04-28

195

Designing the Cascade inertial confinement fusion reactor  

SciTech Connect

The primary goal in designing inertial confinement fusion (ICF) reactors is to produce electrical power as inexpensively as possible, with minimum activation and without compromising safety. This paper discusses a method for designing the Cascade rotating ceramic-granule-blanket reactor (Pitts, 1985) and its associated power plant (Pitts and Maya, 1985). Although focus is on the cascade reactor, the design method and issues presented are applicable to most other ICF reactors.

Pitts, J.H.

1987-02-09

196

Digital computer operation of a nuclear reactor  

DOEpatents

A method is described for the safe operation of a complex system such as a nuclear reactor using a digital computer. The computer is supplied with a data base containing a list of the safe state of the reactor and a list of operating instructions for achieving a safe state when the actual state of the reactor does not correspond to a listed safe state, the computer selects operating instructions to return the reactor to a safe state.

Colley, R.W.

1982-06-29

197

How far is a fusion power reactor from an experimental reactor  

Microsoft Academic Search

To support a request of very substantial resources to build and operate an experimental reactor such as ITER, it is necessary to show that such a device is well positioned on the route towards a reactor and not too far from the reactor in parameter space. For the reactor definition, we choose to start from ITER design and to identify,

R Toschi; P Barabaschi; D Campbell; F Elio; D Maisonnier; D Ward

2001-01-01

198

Discontinuous Galerkin hp-adaptive methods for multiscale chemical reactors: quiescent reactors  

E-print Network

Discontinuous Galerkin hp-adaptive methods for multiscale chemical reactors: quiescent reactors C?, Toronto, ON, M5P 2X8 Abstract We present a class of chemical reactor systems, modeled numerically using-dimensional nonlinear Lotka-Volterra chemical systems. Keywords: Chemical reactors, reaction-diffusion equations, SSPRK

Evans, John A.

199

MTR BUILDING INTERIOR, TRA603, REACTOR FLOOR. DETAIL OF REACTOR TEST ...  

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

MTR BUILDING INTERIOR, TRA-603, REACTOR FLOOR. DETAIL OF REACTOR TEST HOLE OPENING IN WEST FACE. CAMERA FACING NORTHEAST. INL NEGATIVE NO. HD46-2-1. Mike Crane, Photographer, 2/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

200

MTR BUILDING INTERIOR, TRA603, REACTOR FLOOR. DETAIL OF REACTOR'S SOUTH ...  

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

MTR BUILDING INTERIOR, TRA-603, REACTOR FLOOR. DETAIL OF REACTOR'S SOUTH FACE. CAMERA FACING NORTHWESTERLY. INL NEGATIVE NO. HD46-1-1. Mike Crane, Photographer, 2/2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

201

Synthetic fuels from fusion reactors  

Microsoft Academic Search

The technical, environmental, and economic features of a synthetic fuels economy based on fusion reactors are evaluated. Analyses of alternate possible U.S. energy systems for 2020 AD indicate that CTR's can deliver synthetic fuels based on electrolytic hydrogen (H2 gas, H2 liquid, and methanol) at costs competitive with natural fossil fuels and synthetic fuels derived from coal. With less conservative

J. R. Powell; F. J. Salzano; W. Sevian; P. Bezler; G. R. Hopkins; B. Yalof

1974-01-01

202

Nozzle for electric dispersion reactor  

DOEpatents

A nozzle for an electric dispersion reactor includes two coaxial cylindrical bodies, the inner one of the two delivering disperse phase fluid into a continuous phase fluid. A potential difference generated by a voltage source creates a dispersing electric field at the end of the inner electrode. 5 figs.

Sisson, W.G.; Harris, M.T.; Scott, T.C.; Basaran, O.A.

1998-06-02

203

Nozzle for electric dispersion reactor  

DOEpatents

A nozzle for an electric dispersion reactor includes two concentric electrodes, the inner one of the two delivering disperse phase fluid into a continuous phase fluid. A potential difference generated by a voltage source creates a dispersing electric field at the end of the inner electrode. 4 figs.

Sisson, W.G.; Basaran, O.A.; Harris, M.T.

1998-04-14

204

Nozzle for electric dispersion reactor  

DOEpatents

A nozzle for an electric dispersion reactor includes two coaxial cylindrical bodies, the inner one of the two delivering disperse phase fluid into a continuous phase fluid. A potential difference generated by a voltage source creates a dispersing electric field at the end of the inner electrode. 5 figs.

Sisson, W.G.; Harris, M.T.; Scott, T.C.; Basaran, O.A.

1996-04-02

205

Nozzle for electric dispersion reactor  

DOEpatents

A nozzle for an electric dispersion reactor includes two concentric electrodes, the inner one of the two delivering disperse phase fluid into a continuous phase fluid. A potential difference generated by a voltage source creates a dispersing electric field at the end of the inner electrode. 4 figs.

Sisson, W.G.; Basaran, O.A.; Harris, M.T.

1995-11-07

206

Argonne National Laboratory's Reactor Performance  

E-print Network

" the physical processes that occur in a nuclear reactor core, including neutron transport, thermal hydraulics transitional tools to aid industry's migration to future commercially viable petascale computing platforms supported by industrial collaborators and related Department of Energy research programs. 1. Analysis

Kemner, Ken

207

RADIATION EMBRITTLEMENT OF REACTOR VESSELS  

Microsoft Academic Search

A review is presented on the effects of irradiation on pressure-vessel ; steels, on the results of a survey of 19 reactor vessels with respect to their ; anticipated neutron doses, and on the research programs in this field. (D.L.C.);

J. J. DiNunno; A. B. Holt

1963-01-01

208

Physics of nuclear reactor safety  

Microsoft Academic Search

Provides a concise review of the physical aspects of safety of nuclear fission reactors. It covers the developments of roughly the last decade. The introductory chapter contains an analysis of the changes in safety philosophy that are characteristic of the last decade and that have given rise to an increased importance of physical aspects because of the emphasis on passive

H. van Dam

1992-01-01

209

Turbulence may sink titanic reactor  

SciTech Connect

The $10 billion International Thermonuclear Experimental Reactor project is meant to show that fusion is a practical energy source, but a new set of calculations says ITER will fizzle. This article describes the expectations and the projections about its future, as well as the challenges recently mounted using new calculations. 3 figs.

Glanz, J.

1996-12-06

210

Distribution ICategory: General Reactor Technology  

E-print Network

that is not widely shared. The misunderstanding, often deliberate, of the scientific bases for technology, can lead--- Distribution ICategory: General Reactor Technology (UC-520) ANl-92/23 AR(;ONNE NATIONAL LABORATORY 9700 South (:ass Avenue, Argonne, Illinois 60439 IS THiERE A lARGE RISK OF RADIATION? A CRITlt

Shlyakhter, Ilya

211

PCCF flow analysis -- DR Reactor  

Microsoft Academic Search

This report contains an analysis of PCCF tube flow and Panellit pressure relations at DR reactor. Supply curves are presented at front header pressures from 480 to 600 psig using cold water and the standard 0.236 inch orifice with taper down stream and the pigtail valve (plug or ball) open. Demand curves are presented for slug column lengths of 200

Calkin

1961-01-01

212

Fluidized-Bed Reactor System  

NASA Technical Reports Server (NTRS)

Gas pyrolysis in hot fluidized beds minimized by use of selectively filtered radiation and parabolic cavity. Reactor is parabolic cavity of two or more axes in which light emanating from one axis bounces off walls of cavity and passes through object axis to heat sample.

Morrison, A. D.

1985-01-01

213

A Simple Tubular Reactor Experiment.  

ERIC Educational Resources Information Center

Using the hydrolysis of crystal violet dye by sodium hydroxide as an example, the theory, apparatus, and procedure for a laboratory demonstration of tubular reactor behavior are described. The reaction presented can occur at room temperature and features a color change to reinforce measured results. (WB)

Hudgins, Robert R.; Cayrol, Bertrand

1981-01-01

214

Search for other natural fission reactors  

Microsoft Academic Search

Precambrian uranium ores have been surveyed for evidence of other natural fission reactors. The requirements for formation of a natural reactor direct investigations to uranium deposits with large, high-grade ore zones. Massive zones with volumes approximately greater than 1 m³ and concentrations approximately greater than 20 percent uranium are likely places for a fossil reactor if they are approximately greater

K. E. Apt; J. P. Balagna; E. A. Bryant; G. A. Cowan; W. R. Daniels; R. J. Vidale

1977-01-01

215

Bifurcation Analysis of Chemical Reactors with Energy  

E-print Network

Points #12;Abstract This thesis is a study of packed-bed reactors with integrated heat exchangers with focus on bifurcation analysis. A mathematical model of a packed-bed reactor is derived using the mo- lar lowers the amount of bifurcations. #12;Contents Preface iv 1 Introduction 1 2 The Packed-Bed Reactor

216

Nuclear reactor accident at Chernobyl, USSR  

SciTech Connect

The sequence of events and the consequences of the nuclear reactor accident at Chernobyl, USSR in April 1986 are reviewed. The background material in nuclear reactor and nuclear explosion physics required for understanding these events is extensively explained, and the differences between U.S. and Chernobyl-type reactors are pointed out.

Cohen, B.L.

1987-12-01

217

Tritiated water processing for fusion reactors  

Microsoft Academic Search

Tritiated water represents a source of occupational exposure and environmental emissions for fusion and fission reactors. Fusion reactors must operate within stringent radionuclide emission limits. A range of tritiated water concentrations can be generated in fusion reactors, mostly in the form of tritiated light water. In contrast, tritium removal plants have been built in Canada and France to remove tritium

S. K. Sood; K. M. Kalyanam

1995-01-01

218

THE TORY II-A REACTOR PROGRAM  

Microsoft Academic Search

The design requirements, the findamental design problems, and the test ; facility for the Tory II-A reactor are described. Tory II-A was an engineering ; test reactor, built and tested to gain information on the materials, engineering, ; and physics problems. The available air allowed fullpower full-temperature ; operation of the reactor for a period of less thand two minutes.

1963-01-01

219

Physics design of the upgraded TREAT reactor  

Microsoft Academic Search

With the deferral of the Safety Test Facility (STF), the TREAT Upgrade (TU) reactor has assumed a lead role in the US LMFBR safety test program for the foreseeable future. The functional requirements on TU require a significant enhancement of the capability of the current TREAT reactor. A design of the TU reactor has been developed that modifies the central

S. K. Bhattacharyya; R. M. Lell; J. R. Liaw; A. J. Ulrich; D. C. Wade; S. T. Yang

1980-01-01

220

CONFERENCES AND SYMPOSIA FUSION REACTOR DESIGN IV  

E-print Network

CONFERENCES AND SYMPOSIA FUSION REACTOR DESIGN IV Report on the Fourth IAEA Technical Committee; 7.3. Recommendations; 8. Hybrid Fusion-Fission Reactors; 8.1. Status; 8.2. Progress of fusion power reactor development is to bring to the world a new source of unlimited energy. While

Abdou, Mohamed

221

Reactor control rod timing system. [LMFBR  

Microsoft Academic Search

A fluid driven jet-edge whistle timing system is described for control rods of a nuclear reactor for producing real-time detection of the timing of each control rod in its scram operation. An important parameter in reactor safety, particularly for liquid metal fast breeder reactors (LMFBR), is the time deviation between the time the control rod is released and the time

P. T. K

1980-01-01

222

Methanosaeta fibers in anaerobic migrating blanket reactors  

E-print Network

Methanosaeta fibers in anaerobic migrating blanket reactors L.T. Angenent,* D. Zheng,* S. Sung An anaerobic migrating blanket reactor (AMBR) was seeded with flocculent biomass from a digester and fed of operation, a mature granular blanket developed in the reactor. Moreover, fibers of approximately 1 cm long

Angenent, Lars T.

223

REACTOR SAFETY KEYWORDS: best estimate plus  

E-print Network

REACTOR SAFETY KEYWORDS: best estimate plus uncertainty analysis, epistemic error and aleatory phe- nomena that underlie the safety analyses. The use of BE codes within the reactor technology in advance and that result from a variety of operating conditions or states. These arise because the reactor

Hoppe, Fred M.

224

Interdisciplinary Institute for Innovation Nuclear reactors' construction  

E-print Network

Interdisciplinary Institute for Innovation Nuclear reactors' construction costs: The role of lead@mines-paristech.fr hal-00956292,version1-6Mar2014 #12;hal-00956292,version1-6Mar2014 #12;Nuclear reactors' construction reactor construction costs in France and the United States. Studying the cost of nuclear power has often

Paris-Sud XI, Université de

225

Pebble Flow Experiments For Pebble Bed Reactors  

E-print Network

Pebble Flow Experiments For Pebble Bed Reactors Andrew C. Kadak1 Department of Nuclear Engineering of Technology 2nd International Topical Meeting on High Temperature Reactor Technology Institute of Nuclear) to explore several key aspects of pebble flow in pebble-bed reactors. These experiments were done to assess

Bazant, Martin Z.

226

CASCADE OPTIMIZATION AND CONTROL OF BATCH REACTORS  

E-print Network

CASCADE OPTIMIZATION AND CONTROL OF BATCH REACTORS Xiangming Hua, Sohrab Rohani and Arthur Jutan reactors is proposed. A simple physical conservation model is used to describe the dynamics of the batch reactor. Using model reduction a cascade system is developed, which can effectively combine optimization

Jutan, Arthur

227

Savannah River Site production reactor technical specifications. K Production Reactor  

SciTech Connect

These technical specifications are explicit restrictions on the operation of the Savannah River Site K Production Reactor. They are designed to preserve the validity of the plant safety analysis by ensuring that the plant is operated within the required conditions bounded by the analysis, and with the operable equipment that is assumed to mitigate the consequences of an accident. Technical specifications preserve the primary success path relied upon to detect and respond to accidents. This report describes requirements on thermal-hydraulic limits; limiting conditions for operation and surveillance for the reactor, power distribution control, instrumentation, process water system, emergency cooling and emergency shutdown systems, confinement systems, plant systems, electrical systems, components handling, and special test exceptions; design features; and administrative controls.

NONE

1996-02-01

228

Shutdown system for a nuclear reactor  

DOEpatents

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

Groh, Edward F. (Naperville, IL); Olson, Arne P. (Western Springs, IL); Wade, David C. (Naperville, IL); Robinson, Bryan W. (Oak Lawn, IL)

1984-01-01

229

Fast-acting nuclear reactor control device  

DOEpatents

A fast-acting nuclear reactor control device for moving and positioning a fety control rod to desired positions within the core of the reactor between a run position in which the safety control rod is outside the reactor core, and a shutdown position in which the rod is fully inserted in the reactor core. The device employs a hydraulic pump/motor, an electric gear motor, and solenoid valve to drive the safety control rod into the reactor core through the entire stroke of the safety control rod. An overrunning clutch allows the safety control rod to freely travel toward a safe position in the event of a partial drive system failure.

Kotlyar, Oleg M. (Idaho Falls, ID); West, Phillip B. (Idaho Falls, ID)

1993-01-01

230

Shutdown system for a nuclear reactor  

DOEpatents

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

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

1984-06-05

231

Reactor assessments of advanced bumpy torus configurations  

SciTech Connect

Recently, several configurational approaches and concept improvement schemes were introduced for enhancing the performance of the basic ELMO Bumpy Torus (EBT) concept and for improving its reactor potential. These configurations include planar racetrack and square geometries, Andreoletti coil systems, and bumpy torus-stellarator hybrids (which include twisted racetrack and helical axis stellarator-snakey torus). Preliminary evaluations of reactor implications of each of these configurations have been carried out based on magnetics (vacuum) calculations, transport and scaling relationships, and stability properties. Results indicate favorable reactor projections with a significant reduction in reactor physical size as compared to conventional EBT reactor designs carried out in the past.

Uckan, N.A.; Owen, L.W.; Spong, D.A.; Miller, R.L.; Ard, W.B.; Pipkins, J.F.; Schmitt, R.J.

1983-01-01

232

Tandem Mirror Reactor Systems Code (Version I)  

SciTech Connect

A computer code was developed to model a Tandem Mirror Reactor. Ths is the first Tandem Mirror Reactor model to couple, in detail, the highly linked physics, magnetics, and neutronic analysis into a single code. This report describes the code architecture, provides a summary description of the modules comprising the code, and includes an example execution of the Tandem Mirror Reactor Systems Code. Results from this code for two sensitivity studies are also included. These studies are: (1) to determine the impact of center cell plasma radius, length, and ion temperature on reactor cost and performance at constant fusion power; and (2) to determine the impact of reactor power level on cost.

Reid, R.L.; Finn, P.A.; Gohar, M.Y.; Barrett, R.J.; Gorker, G.E.; Spampinaton, P.T.; Bulmer, R.H.; Dorn, D.W.; Perkins, L.J.; Ghose, S.

1985-09-01

233

Reactor pulse repeatability studies at the annular core research reactor  

SciTech Connect

The Annular Core Research Reactor (ACRR) at Sandia National Laboratories is a water-moderated pool-type reactor designed for testing many types of objects in the pulse and steady-state mode of operations. Personnel at Sandia began working to improve the repeatability of pulse operations for experimenters in the facility. The ACRR has a unique UO{sub 2}-BeO fuel that makes the task of producing repeatable pulses difficult with the current operating procedure. The ACRR produces a significant quantity of photoneutrons through the {sup 9}Be({gamma}, n){sup 8}Be reaction in the fuel elements. The photoneutrons are the result of the gammas produced during fission and in fission product decay, so their production is very much dependent on the reactor power history and changes throughout the day/week of experiments in the facility. Because the photoneutrons interfere with the delayed-critical measurements required for accurate pulse reactivity prediction, a new operating procedure was created. The photoneutron effects at delayed critical are minimized when using the modified procedure. In addition, the pulse element removal time is standardized for all pulse operations with the modified procedure, and this produces less variation in reactivity removal times. (authors)

DePriest, K.R. [Applied Nuclear Technologies, Sandia National Laboratories, Mail Stop 1146, Post Office Box 5800, Albuquerque, NM 87185-1146 (United States); Trinh, T.Q. [Nuclear Facility Operations, Sandia National Laboratories, Mail Stop 0614, Post Office Box 5800, Albuquerque, NM 87185-1146 (United States); Luker, S. M. [Applied Nuclear Technologies, Sandia National Laboratories, Mail Stop 1146, Post Office Box 5800, Albuquerque, NM 87185-1146 (United States)

2011-07-01

234

Fission fragment assisted reactor concept for space propulsion: Foil reactor  

NASA Technical Reports Server (NTRS)

The concept is to fabricate a reactor using thin films or foils of uranium, uranium oxide and then to coat them on substrates. These coatings would be made so thin as to allow the escaping fission fragments to directly heat a hydrogen propellant. The idea was studied of direct gas heating and direct gas pumping in a nuclear pumped laser program. Fission fragments were used to pump lasers. In this concept two substrates are placed opposite each other. The internal faces are coated with thin foil of uranium oxide. A few of the advantages of this technology are listed. In general, however, it is felt that if one look at all solid core nuclear thermal rockets or nuclear thermal propulsion methods, one is going to find that they all pretty much look the same. It is felt that this reactor has higher potential reliability. It has low structural operating temperatures, very short burn times, with graceful failure modes, and it has reduced potential for energetic accidents. Going to a design like this would take the NTP community part way to some of the very advanced engine designs, such as the gas core reactor, but with reduced risk because of the much lower temperatures.

Wright, Steven A.

1991-01-01

235

Research Program of a Super Fast Reactor  

SciTech Connect

Research program of a supercritical-pressure light water cooled fast reactor (Super Fast Reactor) is funded by MEXT (Ministry of Education, Culture, Sports, Science and Technology) in December 2005 as one of the research programs of Japanese NERI (Nuclear Energy Research Initiative). It consists of three programs. (1) development of Super Fast Reactor concept; (2) thermal-hydraulic experiments; (3) material developments. The purpose of the concept development is to pursue the advantage of high power density of fast reactor over thermal reactors to achieve economic competitiveness of fast reactor for its deployment without waiting for exhausting uranium resources. Design goal is not breeding, but maximizing reactor power by using plutonium from spent LWR fuel. MOX will be the fuel of the Super Fast Reactor. Thermal-hydraulic experiments will be conducted with HCFC22 (Hydro chlorofluorocarbons) heat transfer loop of Kyushu University and supercritical water loop at JAEA. Heat transfer data including effect of grid spacers will be taken. The critical flow and condensation of supercritical fluid will be studied. The materials research includes the development and testing of austenitic stainless steel cladding from the experience of PNC1520 for LMFBR. Material for thermal insulation will be tested. SCWR (Supercritical-Water Cooled Reactor) of GIF (Generation-4 International Forum) includes both thermal and fast reactors. The research of the Super Fast Reactor will enhance SCWR research and the data base. The research period will be until March 2010. (authors)

Oka, Yoshiaki; Ishiwatari, Yuki; Liu, Jie; Terai, Takayuki; Nagasaki, Shinya; Muroya, Yusa; Abe, Hiroaki [Nuclear Professional School / Department of Nuclear Engineering and Management, The University of Tokyo, Tokaimura, Naka-gun, Ibaraki, 319-1188 (Japan); Mori, Hideo [Department of Mechanical Engineering, Kyushu University (Japan); Akiba, Masato; Akimoto, Hajime; Okumura, Keisuke; Akasaka, Naoaki [Japan Atomic Energy Agency (Japan); GOTO, Shoji [Tokyo Electric Power Company (Japan)

2006-07-01

236

Safety control circuit for a neutronic reactor  

DOEpatents

A neutronic reactor comprising an active portion containing material fissionable by neutrons of thermal energy, means to control a neutronic chain reaction within the reactor comprising a safety device and a regulating device, a safety device including means defining a vertical channel extending into the reactor from an aperture in the upper surface of the reactor, a rod containing neutron-absorbing materials slidably disposed within the channel, means for maintaining the safety rod in a withdrawn position relative to the active portion of the reactor including means for releasing said rod on actuation thereof, a hopper mounted above the active portion of the reactor having a door disposed at the bottom of the hopper opening into the vertical channel, a plurality of bodies of neutron-absorbing materials disposed within the hopper, and means responsive to the failure of the safety rod on actuation thereof to enter the active portion of the reactor for opening the door in the hopper.

Ellsworth, Howard C. (Richland, WA)

2004-04-27

237

Design analysis of the upgraded TREAT reactor  

SciTech Connect

The TREAT reactor, fueled by a dilute dispersion of fully enriched UO/sub 2/ in graphite, has been a premier transient testing facility since 1959. A major Upgrade of the reactor is in progress to enhance its transient testing capability in support of the LMFBR safety program. The TREAT Upgrade (TU) reactor features a modified central zone of the core with higher fissile loadings of the same fuel, clad in Inconel to allow operation at higher temperatures. The demanding functional requirements on the reactor necessitated the use of unique features in the core design which, in turn, presented major calculational complexities in the analysis. Special design methods had to be used in many cases to treat these complexities. The addition of an improved Reactor Control System, a safety grade Plant Protection System and an enhanced Coolant/Filtration System produces a reactor that can meet the functional requirements on the reactor in a safe manner.

Bhattacharyya, S.K.

1982-01-01

238

Microwave coupling in EBT reactor  

SciTech Connect

For a typical size ELMO Bumpy Torus (EBT) reactor (approx. 1000 MWe), microwave frequencies required lie in the range of 60 to 110 GHz at power levels of 50 to 75 MW. As the frequency rises, the unloaded cavity (i.e., without plasma) quality factor Q decreases. Because of the short wavelengths of microwave heating power and the large cavity dimensions of a reactor, it is possible to apply quasi-optical principles in the efficient coupling of power to the plasma. The use of a confocal Fabry-Perot resonator with spherical mirrors is discussed; these serve to confine the microwave power to the region occupied by the plasma. The potential advantages of these resonators include high efficiency utilization of microwave power, minimal thermal burden on the cryopumping system, and significant benefit in preventing microwave leakage from the device. An estimation of the unloaded cavity quality factor Q and the design considerations of Fabry-Perot resonator are given.

Uckan, N.A.; Uckan, T.; Dandl, R.A.

1980-02-01

239

Requirements for Reactor Physics Design  

SciTech Connect

It has been recognized that there is a need for requirements and guidance for design and operation of nuclear power plants. This is becoming more important as more reactors are being proposed to be built. In parallel with activities in individual countries are norms established by international organizations. This paper discusses requirements/guidance for neutronic design and operation as promulgated by the U.S. Nuclear Regulatory Commission (NRC). As an example, details are given for one reactor physics parameter, namely, the moderator temperature reactivity coefficient. The requirements/guidance from the NRC are discussed in the context of those generated for the International Atomic Energy Agency. The requirements/guidance are not identical from the two sources although they are compatible.

Diamond,D.J.

2008-04-11

240

Modular Stellarator Fusion Reactor concept  

SciTech Connect

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

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

1981-08-01

241

Vanadium recycling for fusion reactors  

SciTech Connect

Very stringent purity specifications must be applied to low activation vanadium alloys, in order to meet recycling goals requiring low residual dose rates after 50--100 years. Methods of vanadium production and purification which might meet these limits are described. Following a suitable cooling period after their use, the vanadium alloy components can be melted in a controlled atmosphere to remove volatile radioisotopes. The aim of the melting and decontamination process will be the achievement of dose rates low enough for ``hands-on`` refabrication of new reactor components from the reclaimed metal. The processes required to permit hands-on recycling appear to be technically feasible, and demonstration experiments are recommended. Background information relevant to the use of vanadium alloys in fusion reactors, including health hazards, resources, and economics, is provided.

Dolan, T.J.; Butterworth, G.J.

1994-04-01

242

Operate Your Own Tokamak Reactor!  

NSDL National Science Digital Library

Princeton University's Plasma Physics Laboratory's Interactive Plasma Physics Education Experience Web site has updated its interactive Virtual Tokamak. The Java applet is designed to illustrate the basic principles of magnetically confined fusion, and users can now type in the three parameters that include the heating power, magnetic field, and plasma density. Although the applet doesn't work on older PCs, older browsers, and on most Macs, it's worth finding a newer PC to interactively learn about these specific types of reactors.

Stotler, Daren.

2008-03-11

243

Microbial remediation reactor and process  

US Patent & Trademark Office Database

Embodiments of a remediation reactor and mixer/contactor blade for the reactor are shown and described, the reactor (11) being for containing a liquid slurry, suspension or settled bed of solid particles containing microorganisms. Inside the vessel is a supply conduit and at least one generally horizontal stirrer blade (27) in fluid connection with the supply conduit. The stirrer blade is a mixer/contactor which has a leading side having openings through which fluid may pass. The stirrer blade rotates in the vessel, and this rotation is made easier by the hydraulic forces of fluid flowing out from the stirrer blade. The flowing fluid creates a fluidization zone in the slurry, suspension or settled bed at or near the leading edge of the stirrer blade. The fluidization zone is less dense than the rest of the sediment bed, and the stirrer blade tends to rotate into the fluidization zone. This way, controlled rotation of the stirrer blade may be created near the bottom of the vessel, enhancing mixing of the microorganisms with the slurry, suspension or settled bed in the vessel, without unnecessarily damaging the microorganisms, especially when the flowing fluid contains contaminants which are nutrients for the microorganisms.

2002-02-12

244

The ARIES tokamak reactor study  

SciTech Connect

The ARIES study is a community effort to develop several visions of tokamaks as fusion power reactors. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak which would operate at a higher beta in the 2nd MHD stability regime. It employs both potential advances in the physics and expected advances in technology and engineering. ARIES-II will examine the potential of the tokamak and the D{sup 3}He fuel cycle. This report is a collection of 14 papers on the results of the ARIES study which were presented at the IEEE 13th Symposium on Fusion Engineering (October 2-6, 1989, Knoxville, TN). This collection describes the ARIES research effort, with emphasis on the ARIES-I design, summarizing the major results, the key technical issues, and the central conclusions.

Not Available

1989-10-01

245

Launch of Russian reactor postponed  

SciTech Connect

Astronomers and weapons scientists seemed heated on a collision course a few months ago over the military's plans to send a Russian nuclear reactor into space. But an agreement reached in late January has prevented a pile-up, at least for 6 months. The astronomers, led by Donald Lamb of the University of Chicago, were objecting to plans by the Strategic Defense Initiative Office (SDIO) to launch Topaz 2, an experimental Russian nuclear reactor, arguing that rogue particles from it might ruin sensitive gamma ray experiments. The reactor is designed to propel itself in space with a jet of xenon ions. One worry was that leaking gamma rays and positrons, which can travel in the earth's magnetic field and pop up in the darndest places, might cause false signals in gamma ray monitors (Science, 18 December 1992, p. 1878). The worry has abated now that SDI officials will postpone choosing a rocket and mission altitutde for Topaz 2 for 6 months, while experts study how its emissions at various altitudes might affect instruments aboard the Gamma Ray Observatory and other satellites. In effect, the SDIO has agreed to an environmental impact study for space, following an unusual meeting organized by former Russian space official Roald Sagdeev at the University of Maryland on 19 January. There the Russian designers of Topaz 2, its new owners at the SDIO, and critics in the astronomy community achieved common ground: that more study was needed.

Not Available

1993-02-05

246

Rationale for university research reactors  

SciTech Connect

University research reactors (URRs), of which 36 are currently operating, have been declining in number, at the rate of {approximately} 2/yr, since the mid-1980s. This decrease is often attributed to the continuing malaise of the nuclear power industry. However, such reasoning is specious because, while many URRs do provide training to prospective power plant operators, few, if any, were constructed solely to support nuclear power generators. Rather, the primary mission of each URR is to serve the educational and research needs of its university. For example, URRs provide laboratory exercises for undergraduates, practical training in the radiological sciences at the master`s level, and a means for advanced studies in such fields as geology, pollution control, archaeology, nuclear medicine, control engineering, and materials as well as reactor physics and the nuclear sciences. This paper explores the rationale for university-based research reactors. The principal argument is that URRs are not tied to a given industry or technology. Rather, they provide a means to educate students and to conduct research in a variety of disciplines, and, as such, their value does not diminish with time.

Bernard, J.A. [Massachusetts Institute of Technology, Cambridge, MA (United States)

1994-12-31

247

In-reactor performance of pressure tubes in CANDU reactors  

NASA Astrophysics Data System (ADS)

The pressure tubes in CANDU reactors have been operating for times up to about 25 years. The in-reactor performance of Zr-2.5Nb pressure tubes has been evaluated by sampling and periodic inspection. This paper describes the behaviour and discusses the factors controlling the behaviour of these components in currently operating CANDU reactors. The mechanical properties (such as ultimate tensile strength, UTS, and fracture toughness), and delayed-hydride-cracking properties (crack growth rate Vc, and threshold stress intensity factor, KIH) change with irradiation; the former reach a limiting value at a fluence of <1 × 10 25 n m -2, while Vc and KIH reach a steady-state condition after a fluence of about 3 × 10 25 n m -2 and 3 × 10 24 n m -2, respectively. At saturation the UTS is raised by about 200 MPa, toughness is reduced to about 40% of its initial value, Vc increases by about a factor of ten while KIH is only slightly reduced. The role of microstructure and trace elements in these behaviours is described. Pressure tubes exhibit elongation and diametral expansion. The deformation behaviour is a function of operating conditions and material properties that vary from tube-to-tube and as a function of axial location. Semi-empirical predictive models have been developed to describe the deformation response of average tubes as a function of operating conditions. For corrosion and, more importantly deuterium pickup, semi-empirical predictive models have also been developed to represent the behaviour of an average tube. The effect of material variability on corrosion behaviour is less well defined compared with other properties. Improvements in manufacturing have increased fracture resistance by minimising trace elements, especially H and Cl, and reduced variability by tightening controls on forming parameters, especially hot-working temperatures.

Rodgers, D. K.; Coleman, C. E.; Griffiths, M.; Bickel, G. A.; Theaker, J. R.; Muir, I.; Bahurmuz, A. A.; Lawrence, S. St.; Resta Levi, M.

2008-12-01

248

The Shippingport Pressurized Water Reactor and Light Water Breeder Reactor  

SciTech Connect

This report discusses the Shippingport Atomic Power Station, located in Shippingport, Pennsylvania, which was the first large-scale nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. A program was started in 1953 at the Bettis Laboratory to confirm the practical application of nuclear power for large-scale electric power generation. It led to the development of zirconium alloy (Zircaloy) clad fuel element containing bulk actinide oxide ceramics (UO{sub 2}, ThO{sub 2}, ThO{sub 2} -- UO{sub 2}, ZrO{sub 2} -- UO{sub 2}) as nuclear reactor fuels. The program provided much of the technology being used for design and operation of the commercial, central-station nuclear power plants now in use. The Shippingport Pressurized Water Reactor (PWR) began initial power operation on December 18, 1957, and was a reliable electric power producer until February 1974. In 1965, subsequent to the successful operation of the Shippingport PWR (UO{sub 2}, ZrO{sub 2} -- UO{sub 2} fuels), the Bettis Laboratory undertook a research and development program to design and build a Light Water Breeder Reactor (LWBR) core for operation in the Shippingport Station. Thorium was the fertile fuel in the LWBR core and was the base oxide for ThO{sub 2} and ThO{sub 2} -- UO{sub 2} fuel pellets. The LWBR core was installed in the pressure vessel of the original Shippingport PWR as its last core before decommissioning. The LWBR core started operation in the Shippingport Station in the autumn of 1977 and finished routine power operation on October 1, 1982. Successful LWBR power operation to over 160% of design lifetime demonstrated the performance capability of the core for both base-load and swing-load operation. Postirradiation examinations confirmed breeding and successful performance of the fuel system.

Clayton, J.C.

1993-11-01

249

Evaluation of in-vessel corium retention through external reactor vessel cooling for integral reactor  

SciTech Connect

In-vessel corium retention through external reactor vessel cooling (IVR-ERVC) for a small integral reactor has been evaluated to determine the thermal margin for the prevention of a reactor vessel failure. A thermal load analysis from the corium pool to the outer reactor vessel wall in the lower plenum of the reactor vessel has been performed to determine the heat flux distribution. The critical heat flux (CHF) on the outer reactor vessel wall has been determined to fix the maximum heat removal rate through the external coolant between the outer reactor vessel and the insulation of the reactor vessel. Finally, the thermal margin has been evaluated by comparison of the thermal load with the maximum heat removal rate of the CHF on the outer reactor vessel wall. The maximum heat flux from the corium pool to the outer reactor vessel is estimated at approximately 0.25 MW/m{sup 2} in the metallic layer because of the focusing effect. The CHF of the outer reactor vessel is approximately 1.1 MW/m{sup 2} because of a two phase natural circulation mass flow. Since the thermal margin for the IVR-ERVC is sufficient, the reactor vessel integrity is maintained during a severe accident of a small integral reactor. (authors)

Park, R. J.; Lee, J. R.; Kim, S. B.; Jin, Y.; Kim, H. Y. [Korea Atomic Energy Research Inst., 1045 Daedeok-daero, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of)

2012-07-01

250

Thermionic reactors for space nuclear power  

NASA Technical Reports Server (NTRS)

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

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

1985-01-01

251

PID Control Effectiveness for Surface Reactor Concepts  

SciTech Connect

Control of space and surface fission reactors should be kept as simple as possible, because of the need for high reliability and the difficulty to diagnose and adapt to control system failures. Fortunately, compact, fast-spectrum, externally controlled reactors are very simple in operation. In fact, for some applications it may be possible to design low-power surface reactors without the need for any reactor control after startup; however, a simple proportional, integral, derivative (PID) controller can allow a higher performance concept and add more flexibility to system operation. This paper investigates the effectiveness of a PID control scheme for several anticipated transients that a surface reactor might experience. To perform these analyses, the surface reactor transient code FRINK was modified to simulate control drum movements based on bulk coolant temperature.

Dixon, David D. [North Carolina State University, Raleigh, NC (United States); Los Alamos National Laboratory, Los Alamos, NM (United States); Marsh, Christopher L. [United States Naval Academy, Annapolis, MD (United States); Los Alamos National Laboratory, Los Alamos, NM (United States); Poston, David I. [Los Alamos National Laboratory, Los Alamos, NM (United States)

2007-01-30

252

Profiling a reactor component using ultrasonics  

SciTech Connect

Nuclear reactors have many components within the reactor vessel. During the life of a reactor it is possible for these components to be displaced or deformed because of the thermal cycles to which they are subject. Also, these components in situ therefore becomes important for the upkeep of the reactor. However, high radiation levels make it difficult to monitor using optical methods. This paper describes an ultrasonic method which was successfully employed in profiling a deformed guide tube of a reactor. The method uses the well-known ultrasonic ranging technique. However, the specialty of the method is the use of air transducers at 40 kHz to overcome the inherent divergence problems and the difficulties associated with high temperatures inherent in a sodium cooled reactor.

Pathak, L.; Seshadri, V.R.; Kumaravadivelu, C.; Sreenivasan, G.; Raghunathan, V.S. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)] [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

1995-04-01

253

Fission Product Release from SLOWPOKE-2 Reactors  

NASA Astrophysics Data System (ADS)

Increasing radiation fields at several SLOWPOKE -2 reactors fuelled with highly enriched uranium aluminum alloy fuel have begun to interfere with the daily operation of these reactors. To investigate this phenomenon, samples of reactor container water and gas from the headspace above the reactor were obtained at four SLOWPOKE-2 reactor facilities and examined by gamma ray spectroscopy methods. These radiation fields are due to the circulation of fission products within the reactor container vessel. The most likely source of the fission product release is an area of uranium-bearing material exposed to the coolant at the end weld line which originated at the time of fuel fabrication. The results of this study are compared with observations from an underwater visual examination of one core and the metallographic examination of archived fuel elements.

Harnden, Anne M. C.

254

Antineutrino Monitoring for Heavy Water Reactors  

NASA Astrophysics Data System (ADS)

In this Letter we discuss the potential application of antineutrino monitoring to the Iranian heavy water reactor at Arak, the IR-40, as a nonproliferation measure. An above ground detector positioned right outside the IR-40 reactor building could meet IAEA verification goals for reactor plutonium inventories. While detectors with the needed spectral sensitivity have been demonstrated below ground, additional research and development is needed to demonstrate an above-ground detector with this same level of sensitivity. In addition to monitoring the reactor during operation, observing antineutrino emissions from long-lived fission products could also allow monitoring the reactor when it is shut down, provided very low detector backgrounds can be achieved. Antineutrino monitoring could also be used to distinguish different levels of fuel enrichment. Most importantly, these capabilities would not require a complete reactor operational history and could provide a means to reestablish continuity of knowledge in safeguards conclusions should this become necessary.

Christensen, Eric; Huber, Patrick; Jaffke, Patrick; Shea, Thomas E.

2014-07-01

255

Recent progress in stellarator reactor conceptual design  

SciTech Connect

The Stellarator/Torsatron/Heliotron (S/T/H) class of toroidal magnetic fusion reactor designs continues to offer a distinct and in several ways superior approach to eventual commercial competitiveness. Although no major, integrated conceptual reactor design activity is presently underway, a number of international research efforts suggest avenues for the substantial improvement of the S/T/H reactor embodiment, which derive from recent experimental and theoretical progress and are responsive to current trends in fusion-reactor projection to set the stage for a third generation of designs. Recent S/T/H reactor design activity is reviewed and the impact of the changing technical and programmatic context on the direction of future S/T/H reactor design studies is outlined.

Miller, R.L.

1985-01-01

256

Experimental development of power reactor advanced controllers  

SciTech Connect

A systematic approach for developing and verifying advanced controllers with potential application to commercial nuclear power plants is suggested. The central idea is to experimentally demonstrate an advanced control concept first on an ultra safe research reactor followed by demonstration on a passively safe experimental power reactor and then finally adopt the technique for improving safety, performance, reliability and operability at commercial facilities. Prior to completing an experimental sequence, the benefits and utility of candidate advanced controllers should be established through theoretical development and simulation testing. The applicability of a robust optimal observer-based state feedback controller design process for improving reactor temperature response for a TRIGA research reactor, Liquid Metal-cooled Reactor (LMR), and a commercial Pressurized Water Reactor (PWR) is presented to illustrate the potential of the proposed experimental development concept.

Edwards, R.M. (Pennsylvania State Univ., University Park, PA (United States). Dept. of Nuclear Engineering); Weng, C.K. (Pennsylvania State Univ., University Park, PA (United States). Dept. of Mechanical Engineering); Lindsay, R.W. (Argonne National Lab., Idaho Falls, ID (United States))

1992-01-01

257

Experimental development of power reactor advanced controllers  

SciTech Connect

A systematic approach for developing and verifying advanced controllers with potential application to commercial nuclear power plants is suggested. The central idea is to experimentally demonstrate an advanced control concept first on an ultra safe research reactor followed by demonstration on a passively safe experimental power reactor and then finally adopt the technique for improving safety, performance, reliability and operability at commercial facilities. Prior to completing an experimental sequence, the benefits and utility of candidate advanced controllers should be established through theoretical development and simulation testing. The applicability of a robust optimal observer-based state feedback controller design process for improving reactor temperature response for a TRIGA research reactor, Liquid Metal-cooled Reactor (LMR), and a commercial Pressurized Water Reactor (PWR) is presented to illustrate the potential of the proposed experimental development concept.

Edwards, R.M. [Pennsylvania State Univ., University Park, PA (United States). Dept. of Nuclear Engineering; Weng, C.K. [Pennsylvania State Univ., University Park, PA (United States). Dept. of Mechanical Engineering; Lindsay, R.W. [Argonne National Lab., Idaho Falls, ID (United States)

1992-06-01

258

Dynamics of heat-pipe reactors  

NASA Technical Reports Server (NTRS)

A split-core heat pipe reactor, fueled with either U(233)C or U(235)C in a tungsten cermet and cooled by 7-Li-W heat pipes, was examined for the effects of the heat pipes on reactor while trying to safely absorb large reactivity inputs through inherent shutdown mechanisms. Limits on ramp reactivity inputs due to fuel melting temperature and heat pipe wall heat flux were mapped for the reactor in both startup and at-power operating modes.

Niederauer, G. F.

1971-01-01

259

Development of the reactor safety film  

SciTech Connect

This paper summarizes the text and describes the processes followed to develop the first computer-generated film of LASL's Reactor Safety efforts. The 11-1/2 min film with narrative and musical background gives a brief overview of reactor components, of how LASL's Reactor Safety groups develop and verify computer codes to anticipate accidents, and of how these codes were applied to the Three Mile Island accident.

Sheheen, N.N.

1980-01-01

260

Nuclear weapons and power-reactor plutonium  

Microsoft Academic Search

1-10 that for making nuclear bombs, 'reactor-grade' plutonium produced by the normal operation of uranium-fuelled power reactors is necessarily much inferior to specially made 'weapons-grade' Pu: so infe- rior in explosive power or predictability that its potential use by amateurs is not a serious problem and that governments would instead make the higher-performance weapons-grade Pu in special production reactors. Although

Amory B. Lovins

1980-01-01

261

Small Reactor for Deep Space Exploration  

SciTech Connect

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

None

2012-11-29

262

Xanthan production in a plunging jet reactor  

Microsoft Academic Search

A plunging jet reactor (0.04–0.08 m3) was used for the production of the exopolysaccharide xanthan with Xanthomonas campestris. The microorganism was not affected by the pump shear force. Similar specific growth rates and xanthan space-time yields to those in other reactor types were achieved at much lower specific power input. The better oxygen sorption efficiency in the jet reactor overcompensated

Ahmed Zaidi; Purnendu Ghosh; Adrian Schumpe; Wolf-Dieter Deckwer

1991-01-01

263

SRS's production reactor severe-accident issues  

Microsoft Academic Search

Severe accident behavior in Al-U metal-fueled production reactors is sufficiently different from commercial oxide-fueled reactors that an ongoing research and development (R and D) effort is needed to resolve issues specific to metal-fueled systems. The severe-accident program at the Savannah River Site (SRS) is based on light water reactor (LWR) technology when appropriate. A direct application of severe-accident phenomena technology

P. G. Ellison; M. L. Hyder; P. R. Monson; M. J. Ades

1991-01-01

264

NASA Reactor Facility Hazards Summary. Volume 1  

NASA Technical Reports Server (NTRS)

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.

1959-01-01

265

Desirability of small reactors, HTGR in particular  

Microsoft Academic Search

Small reactors of about 100–300 MWe, High Temperature Gas Cooled Reactors (HTGRs) in particular, are considered desirable in future, based on the following ways of thinking;Global scale enhancement of nuclear energy is considered necessary from reduction of environment impact point of view.Small reactors are desirable, due to (a) enhanced safety in terms of fuel inventory and inherent safety, then (b)

Yasuo Tsuchie

2000-01-01

266

Neutron shielding panels for reactor pressure vessels  

DOEpatents

In a nuclear reactor neutron panels varying in thickness in the circumferential direction are disposed at spaced circumferential locations around the reactor core so that the greatest radial thickness is at the point of highest fluence with lesser thicknesses at adjacent locations where the fluence level is lower. The neutron panels are disposed between the core barrel and the interior of the reactor vessel to maintain radiation exposure to the vessel within acceptable limits.

Singleton, Norman R. (Murrysville, PA)

2011-11-22

267

Search for ?13 with reactor experiments  

NASA Astrophysics Data System (ADS)

After the discovery of neutrino oscillations, two angles of the leptonic unitary mixing matrix have been determined. However, for ?13 only upper limits exist. The best constraints have been obtained in the reactor oscillation experiment CHOOZ. By reducing statistical as well as systematical uncertainties, the sensitivity on ?13 of new reactor experiments can be improved by one order of magnitude. This article focuses on the proposed reactor experiment DOUBLE-CHOOZ.

Oberauer, L.

2006-07-01

268

Small Reactor for Deep Space Exploration  

ScienceCinema

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

None

2014-05-30

269

Monitoring Reactor Antineutrino Flux for Nonproliferation  

NASA Astrophysics Data System (ADS)

Under the Non-Proliferation Treaty, the International Atomic Energy Agency has installed nuclear safeguard systems to monitor reactors. These systems, while effective, lack certain attractive features: they cannot provide real-time monitoring of reactor activities and some of them interfere with reactor operations. Antineutrino detectors can provide a continuous, real-time, and less intrusive method for monitoring reactors. This proposed safeguards system, tested at reactors in Russia and the United States, spins off from antineutrino experiments, many of which use reactors to produce antineutrinos. Monitoring antineutrino flux can detect illicit activities in reactors, such as the diversion of plutonium. Sensitivity to changes in fissile content in a few months using only antineutrino data has been demonstrated at the level of 70 kg of plutonium with >99% confidence. As part of the monitoring technique, it is useful to have accurate predictions of the evolving antineutrino flux that results from reactor fuel burnup. Simulations predicting the evolution are being developed and tested in present antineutrino reactor experiments.

Shen, Fangfei; Jones, Christopher; Keefer, Gregory; Winslow, Lindley; Djurcic, Zelimir; Bernstein, Adam; Conrad, Janet

2011-04-01

270

Heat dissipating nuclear reactor with metal liner  

DOEpatents

A nuclear reactor containment including a reactor vessel disposed within a cavity with capability for complete inherent decay heat removal in the earth and surrounded by a cast steel containment member which surrounds the vessel is described in this disclosure. The member has a thick basemat in contact with metal pilings. The basemat rests on a bed of porous particulate material, into which water is fed to produce steam which is vented to the atmosphere. There is a gap between the reactor vessel and the steel containment member. The containment member holds any sodium or core debris escaping from the reactor vessel if the core melts and breaches the vessel.

Gluekler, E.L.; Hunsbedt, A.; Lazarus, J.D.

1985-11-21

271

Continuous steroid biotransformations in microchannel reactors.  

PubMed

The use of microchannel reactor based technologies within the scope of bioprocesses as process intensification and production platforms is gaining momentum. Such trend can be ascribed a particular set of characteristics of microchannel reactors, namely the enhanced mass and heat transfer, combined with easier handling and smaller volumes required, as compared to traditional reactors. In the present work, a continuous production process of 4-cholesten-3-one by the enzymatic oxidation of cholesterol without the formation of any by-product was assessed. The production was carried out within Y-shaped microchannel reactors in an aqueous-organic two-phase system. Substrate was delivered from the organic phase to aqueous phase containing cholesterol oxidase and the product formed partitions back to the organic phase. The aqueous phase was then forced through a plug-flow reactor, containing immobilized catalase. This step aimed at the reduction of hydrogen peroxide formed as a by-product during cholesterol oxidation, to avoid cholesterol oxidase deactivation due to said by-product. This setup was compared with traditional reactors and modes of operation. The results showed that microchannel reactor geometry outperformed traditional stirred tank and plug-flow reactors reaching similar conversion yields at reduced residence time. Coupling the plug-flow reactor containing catalase enabled aqueous phase reuse with maintenance of 30% catalytic activity of cholesterol oxidase while eliminating hydrogen peroxide. A final production of 36 m of cholestenone was reached after 300 hours of operation. PMID:22008387

Marques, Marco P C; Fernandes, Pedro; Cabral, Joaquim M S; Znidarši?-Plazl, Polona; Plazl, Igor

2012-01-15

272

Italian hybrid and fission reactors scenario analysis  

SciTech Connect

Italy is a country where a long tradition of studies both in the fission and fusion field is consolidated; nevertheless a strong public opinion concerned with the destination of the Spent Nuclear Fuel hinders the development of nuclear power. The possibility to a severe reduction of the NSF mass generated from a fleet of nuclear reactors employing an hypothetical fusionfission hybrid reactor has been investigated in the Italian framework. The possibility to produce nuclear fuel for the fission nuclear reactors with the hybrid reactor was analyzed too.

Ciotti, M.; Manzano, J.; Sepielli, M. [ENEA CR Frascati, Via Enrico Fermi, 45, 00044, Frascati, Roma (Italy); ENEA CR casaccia, Via Anguillarese, 301, 00123, Santa Maria di Galeria, Roma (Italy)

2012-06-19

273

NCSU reactor sharing program. Final technical report  

SciTech Connect

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

Perez, P.B.

1997-01-10

274

Search for sterile neutrinos at reactors  

E-print Network

The sensitivity to the sterile neutrino mixing at very short baseline reactor neutrino experiments is investigated. In the case of conventional (thermal neutron) reactors it is found that the sensitivity is lost for $\\Delta m^2 \\gtrsim$ 1 eV$^2$ due to smearing of the reactor core size. On the other hand, in the case of an experimental fast neutron reactor Joyo, because of its small size, sensitivity to $\\sin^22\\theta_{14}$ can be as good as 0.03 for $\\Delta m^2 \\sim$ several eV$^2$ with the Bugey-like detector setup.

Osamu Yasuda

2011-07-24

275

Nuclear reactor system with aligned feedwater and superheater penetrations  

Microsoft Academic Search

A nuclear reactor system is described wherein a prestressed concrete reactor vessel is provided with a main cavity for the reactor core and at least one subsidiary cavity for a steam generator. At least two feedwater penetrations are provided in the reactor vessel communicating between the exterior of the reactor vessel and the subsidiary cylindrical cavity. A superheater penetration is

E. J. Hurn; J. A. Kissinger

1981-01-01

276

Reactor vital equipment determination techniques  

SciTech Connect

The Reactor Vital Equipment Determination Techniques program at the Los Alamos National Laboratory is discussed. The purpose of the program is to provide the Nuclear Regulatory Commission (NRC) with technical support in identifying vital areas at nuclear power plants using a fault-tree technique. A reexamination of some system modeling assumptions is being performed for the Vital Area Analysis Program. A short description of the vital area analysis and supporting research on modeling assumptions is presented. Perceptions of program modifications based on the research are outlined, and the status of high-priority research topics is discussed.

Bott, T.F.; Thomas, W.S.

1983-01-01

277

Operate Your Own Tokamak Reactor!  

NSDL National Science Digital Library

Princeton University's Plasma Physics Laboratory's Interactive Plasma Physics Education Experience Web site (last mentioned in the July 26, 2002 NSDL Physical Sciences ) has updated its interactive Virtual Tokamak. The Java applet is designed to illustrate the basic principles of magnetically confined fusion, and users can now type in the three parameters that include the heating power, magnetic field, and plasma density. Although the applet doesn't work on older PCs, older browsers, and on most Macs, it's worth finding a newer PC to interactively learn about these specific types of reactors.

Stotler, Daren.

2001-01-01

278

Reactor pressure vessel vented head  

DOEpatents

A head for closing a nuclear reactor pressure vessel shell includes an arcuate dome having an integral head flange which includes a mating surface for sealingly mating with the shell upon assembly therewith. The head flange includes an internal passage extending therethrough with a first port being disposed on the head mating surface. A vent line includes a proximal end disposed in flow communication with the head internal passage, and a distal end disposed in flow communication with the inside of the dome for channeling a fluid therethrough. The vent line is fixedly joined to the dome and is carried therewith when the head is assembled to and disassembled from the shell.

Sawabe, James K. (San Jose, CA)

1994-01-11

279

Reactors for nuclear electric propulsion  

SciTech Connect

Propulsion is the key to space exploitation and power is the key to propulsion. This paper examines the role of nuclear fission reactors as the primary power source for high specific impulse electric propulsion systems for space missions of the 1980s and 1990s. Particular mission applications include transfer to and a reusable orbital transfer vehicle from low-Earth orbit to geosynchronous orbit, outer planet exploration and reconnaissance missions, and as a versatile space tug supporting lunar resource development. Nuclear electric propulsion is examined as an indispensable component in space activities of the next two decades.

Buden, D.; Angelo, J.A. Jr.

1981-01-01

280

97. ARAIII. ML1 reactor has been moved into GCRE reactor ...  

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

97. ARA-III. ML-1 reactor has been moved into GCRE reactor building (ARA-608) for examination of corrosion on its underside and repair. May 24, 1963. Ineel photo no. 63-3485. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

281

Catalytic combustion in a reactor with periodic flow reversal. Part 2. Steady-state reactor model  

Microsoft Academic Search

Analysis of the reverse-flow reactor and comparison to a conventional adiabatic fixed bed reactor with external heat exchanger has shown that both systems are closely related. The required additional system parameter is the center of gravity of energy release caused by exothermic chemical reaction, which can also be rationalized as the mean fraction of the reverse-flow reactor acting as a

H. Züfle; T. Turek

1997-01-01

282

Modelling of fixed-bed reactor: two models of industrial reactor for selective hydrogenation of acetylene  

Microsoft Academic Search

Modelling and simulation of operation conditions for a heterogeneous fixed-bed reactor are investigated on the basis of the industrial reactor for selective hydrogenation of acetylene on palladium catalyst (in ethylene production). Heterogeneous and pseudohomogeneous models of the above-mentioned reactor with their advantages and drawbacks are presented. It is shown, that the proper formulation of a pseudohomogeneous model allows for its

M. Szukiewicz; K. Kaczmarski; R. Petrus

1998-01-01

283

Reactor physics calculations for alternative fuel recycling strategies using tight pressurized water reactor lattices  

Microsoft Academic Search

Physics calculations have been performed for repeated plutonium recycling in tight pressurized water reactor lattices. These calculations made use of the transport theory code CASMO combined with a 70-group nuclear data library variant that was created recently. The calculational model, which performs well for normal thermal reactors, was tested against measured data for tight lattices from the Swiss reactor PROTEUS.

Johansson

1988-01-01

284

155. ARAIII Reactor building (ARA608) Details of reactor pit showing ...  

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

155. ARA-III Reactor building (ARA-608) Details of reactor pit showing tray supports and fuel element storage rack. Aerojet-general 880-area/GCRE-608-MS-2. Date: November 1958. Ineel index code no. 063-0608-40-013-102625. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

285

Reactor Safety Evaluation Division (RSED): K-Reactor Operational Readiness Evaluation (ORE) procedure  

Microsoft Academic Search

The purpose of this procedure is to define the scope, organization, and responsibility and to provide instructions for the performance and documentation of the Reactor Safety Evaluation Division's (RSED) independent review and oversight, referred to as the Operational Readiness Evaluation (ORE), of the K-Reactor Restart Effort. The restart effort is defined in the RSED K-Reactor Operational Readiness Evaluation Program Plan.

V. S. OBlock; D. A. Busch

1989-01-01

286

64. ARAII. Interior view of SL1 reactor building with reactor ...  

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

64. ARA-II. Interior view of SL-1 reactor building with reactor head in place in center foreground. March 21, 1958. Ineel photo no. 58-1360. Photographer: Jack L. Anderson. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

287

Opportunities for the Precision Study of Reactor Antineutrinos at Very Short Baselines at US Research Reactors  

E-print Network

Opportunities for the Precision Study of Reactor Antineutrinos at Very Short Baselines at US Research Reactors T. Allen,1, 2 A.B. Balantekin,1 H.R. Band,1 A. Bernstein,3 N. Bowden,3 C. Bryan,4 S. Hans Institute of Standards and Technology, Gaithersburg, MD 20899 Antineutrinos from nuclear reactors have

288

Thermionic conversion reactor technology assessment  

NASA Astrophysics Data System (ADS)

The in-core thermionic space nuclear power supply is the only identified reactor-power concept that meets the SP-100 size functional requirements with demonstrated state-of-the-art reactor system and space-qualified power system component temperatures. The SP-100 configuration limits provide a net 40 m(2) of primary non-deployed radiator area. If a reasonable 7 year degradation allowance of 15% to 20% is provided then the beginning of life (BOL) net power output requirement is about 120 kWe. Consequently, the SP-100 power system must produce a P/A of 2.7 kWe/m(2). This non-deployed radiator area power density performance is only reasonably achieved by the thermionic in-core converter system, the potassium Rankine turbine system and the Stirling engine system. Past and current tests and data were examined and the potential for successful development of suitable fueled-thermionic converters that will meet SP-100 and growth requirements was assessed. The basis for the assessment will be provided and the recommended key developments plant set forth.

1984-02-01

289

Simulated nuclear reactor fuel assembly  

DOEpatents

An apparatus for electrically simulating a nuclear reactor fuel assembly. It includes a heater assembly having a top end and a bottom end and a plurality of concentric heater tubes having electrical circuitry connected to a power source, and radially spaced from each other. An outer target tube and an inner target tube is concentric with the heater tubes and with each other, and the outer target tube surrounds and is radially spaced from the heater tubes. The inner target tube is surrounded by and radially spaced from the heater tubes and outer target tube. The top of the assembly is generally open to allow for the electrical power connection to the heater tubes, and the bottom of the assembly includes means for completing the electrical circuitry in the heater tubes to provide electrical resistance heating to simulate the power profile in a nuclear reactor. The embedded conductor elements in each heater tube is split into two halves for a substantial portion of its length and provided with electrical isolation such that each half of the conductor is joined at one end and is not joined at the other end.

Berta, Victor T. (Idaho Falls, ID)

1993-01-01

290

Simulated nuclear reactor fuel assembly  

DOEpatents

An apparatus for electrically simulating a nuclear reactor fuel assembly. It includes a heater assembly having a top end and a bottom end and a plurality of concentric heater tubes having electrical circuitry connected to a power source, and radially spaced from each other. An outer target tube and an inner target tube is concentric with the heater tubes and with each other, and the outer target tube surrounds and is radially spaced from the heater tubes. The inner target tube is surrounded by and radially spaced from the heater tubes and outer target tube. The top of the assembly is generally open to allow for the electrical power connection to the heater tubes, and the bottom of the assembly includes means for completing the electrical circuitry in the heater tubes to provide electrical resistance heating to simulate the power profile in a nuclear reactor. The embedded conductor elements in each heater tube is split into two halves for a substantial portion of its length and provided with electrical isolation such that each half of the conductor is joined at one end and is not joined at the other end.

Berta, V.T.

1993-04-06

291

New Production Reactors Program Plan  

SciTech Connect

Part I of this New Production Reactors (NPR) Program Plan: describes the policy basis of the NPR Program; describes the mission and objectives of the NPR Program; identifies the requirements that must be met in order to achieve the mission and objectives; and describes and assesses the technology and siting options that were considered, the Program's preferred strategy, and its rationale. The implementation strategy for the New Production Reactors Program has three functions: Linking the design, construction, operation, and maintenance of facilities to policies requirements, and the process for selecting options. The development of an implementation strategy ensures that activities and procedures are consistent with the rationale and analysis underlying the Program. Organization of the Program. The strategy establishes plans, organizational structure, procedures, a budget, and a schedule for carrying out the Program. By doing so, the strategy ensures the clear assignment of responsibility and accountability. Management and monitoring of the Program. Finally, the strategy provides a basis for monitoring the Program so that technological, cost, and scheduling issues can be addressed when they arise as the Program proceeds. Like the rest of the Program Plan, the Implementation Strategy is a living document and will be periodically revised to reflect both progress made in the Program and adjustments in plans and policies as they are made. 21 figs., 5 tabs.

Not Available

1990-12-01

292

Compound cryopump for fusion reactors  

E-print Network

We reconsider an old idea: a three-stage compound cryopump for use in fusion reactors such as DEMO. The helium "ash" is adsorbed on a 4.5 K charcoal-coated surface, while deuterium and tritium are adsorbed at 15-22 K on a second charcoal-coated surface. The helium is released by raising the first surface to ~30 K. In a separate regeneration step, deuterium and tritium are released at ~110 K. In this way, the helium can be pre-separated from other species. In the simplest design, all three stages are in the same vessel, with a single valve to close the pump off from the tokamak during regeneration. In an alternative design, the three stages are in separate vessels, connected by valves, allowing the stages to regenerate without interfering with each other. The inclusion of the intermediate stage would not affect the overall pumping speed significantly. The downstream exhaust processing system could be scaled down, as much of the deuterium and tritium could be returned directly to the reactor. This could reduce ...

Kovari, M; Shephard, T

2013-01-01

293

Coupled Reactor Kinetics and Heat Transfer Model for Heat Pipe Cooled Reactors  

SciTech Connect

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

WRIGHT,STEVEN A.; HOUTS,MICHAEL

2000-11-22

294

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

NASA Astrophysics Data System (ADS)

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

Wright, Steven A.; Houts, Michael

2001-02-01

295

Looking Northeast in Oxide Building at Reactors on Second Floor ...  

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

Looking Northeast in Oxide Building at Reactors on Second Floor Including Reactor One (Left) and Reactor Two (Right) - Hematite Fuel Fabrication Facility, Oxide Building & Oxide Loading Dock, 3300 State Road P, Festus, Jefferson County, MO

296

78 FR 64028 - Decommissioning of Nuclear Power Reactors  

Federal Register 2010, 2011, 2012, 2013

...NRC-2012-0035] Decommissioning of Nuclear Power Reactors AGENCY: Nuclear Regulatory...1.184 ``Decommissioning of Nuclear Power Reactors.'' This guide describes...the decommissioning process for nuclear power reactors. The revision...

2013-10-25

297

22.312 Engineering of Nuclear Reactors, Fall 2002  

E-print Network

Engineering principles of nuclear reactors, emphasizing power reactors. Power plant thermodynamics, reactor heat generation and removal (single-phase as well as two-phase coolant flow and heat transfer), and structural ...

Todreas, Neil E.

298

22.312 Engineering of Nuclear Reactors, Fall 2004  

E-print Network

Engineering principles of nuclear reactors, emphasizing power reactors. Power plant thermodynamics, reactor heat generation and removal (single-phase as well as two-phase coolant flow and heat transfer), and structural ...

Buongiorno, Jacopo, 1971-

299

9 CFR 78.22 - Brucellosis reactor bison.  

Code of Federal Regulations, 2013 CFR

...2013-01-01 2013-01-01 false Brucellosis reactor bison. 78.22 Section 78.22 Animals...of Brucellosis § 78.22 Brucellosis reactor bison. (a) Destination. Brucellosis reactor bison may be moved interstate only for...

2013-01-01

300

9 CFR 78.7 - Brucellosis reactor cattle.  

Code of Federal Regulations, 2011 CFR

...2011-01-01 2011-01-01 false Brucellosis reactor cattle. 78.7 Section 78.7 Animals...Because of Brucellosis § 78.7 Brucellosis reactor cattle. (a) Destination. Brucellosis reactor cattle may be moved interstate only...

2011-01-01

301

9 CFR 78.31 - Brucellosis reactor swine.  

Code of Federal Regulations, 2011 CFR

...2011-01-01 2011-01-01 false Brucellosis reactor swine. 78.31 Section 78.31 Animals...of Brucellosis § 78.31 Brucellosis reactor swine. (a) Destination. Brucellosis reactor swine may be moved interstate only for...

2011-01-01

302

9 CFR 78.7 - Brucellosis reactor cattle.  

Code of Federal Regulations, 2010 CFR

...2010-01-01 2010-01-01 false Brucellosis reactor cattle. 78.7 Section 78.7 Animals...Because of Brucellosis § 78.7 Brucellosis reactor cattle. (a) Destination. Brucellosis reactor cattle may be moved interstate only...

2010-01-01

303

REACTOR BASE. THREE VERTICAL PIPES ON LEFT SIDE OF VIEW ...  

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

REACTOR BASE. THREE VERTICAL PIPES ON LEFT SIDE OF VIEW WILL BE FOR COOLING AIR. INL NEGATIVE NO. 514. Unknown Photographer, 8/30/1950 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

304

9 CFR 78.22 - Brucellosis reactor bison.  

Code of Federal Regulations, 2010 CFR

...2010-01-01 2010-01-01 false Brucellosis reactor bison. 78.22 Section 78.22 Animals...of Brucellosis § 78.22 Brucellosis reactor bison. (a) Destination. Brucellosis reactor bison may be moved interstate only for...

2010-01-01

305

9 CFR 78.22 - Brucellosis reactor bison.  

...2014-01-01 2014-01-01 false Brucellosis reactor bison. 78.22 Section 78.22 Animals...of Brucellosis § 78.22 Brucellosis reactor bison. (a) Destination. Brucellosis reactor bison may be moved interstate only for...

2014-01-01

306

9 CFR 78.7 - Brucellosis reactor cattle.  

Code of Federal Regulations, 2013 CFR

...2013-01-01 2013-01-01 false Brucellosis reactor cattle. 78.7 Section 78.7 Animals...Because of Brucellosis § 78.7 Brucellosis reactor cattle. (a) Destination. Brucellosis reactor cattle may be moved interstate only...

2013-01-01

307

9 CFR 78.22 - Brucellosis reactor bison.  

Code of Federal Regulations, 2011 CFR

...2011-01-01 2011-01-01 false Brucellosis reactor bison. 78.22 Section 78.22 Animals...of Brucellosis § 78.22 Brucellosis reactor bison. (a) Destination. Brucellosis reactor bison may be moved interstate only for...

2011-01-01

308

9 CFR 78.31 - Brucellosis reactor swine.  

...2014-01-01 2014-01-01 false Brucellosis reactor swine. 78.31 Section 78.31 Animals...of Brucellosis § 78.31 Brucellosis reactor swine. (a) Destination. Brucellosis reactor swine may be moved interstate only for...

2014-01-01

309

9 CFR 78.31 - Brucellosis reactor swine.  

Code of Federal Regulations, 2010 CFR

...2010-01-01 2010-01-01 false Brucellosis reactor swine. 78.31 Section 78.31 Animals...of Brucellosis § 78.31 Brucellosis reactor swine. (a) Destination. Brucellosis reactor swine may be moved interstate only for...

2010-01-01

310

9 CFR 78.22 - Brucellosis reactor bison.  

Code of Federal Regulations, 2012 CFR

...2012-01-01 2012-01-01 false Brucellosis reactor bison. 78.22 Section 78.22 Animals...of Brucellosis § 78.22 Brucellosis reactor bison. (a) Destination. Brucellosis reactor bison may be moved interstate only for...

2012-01-01

311

9 CFR 78.7 - Brucellosis reactor cattle.  

Code of Federal Regulations, 2012 CFR

...2012-01-01 2012-01-01 false Brucellosis reactor cattle. 78.7 Section 78.7 Animals...Because of Brucellosis § 78.7 Brucellosis reactor cattle. (a) Destination. Brucellosis reactor cattle may be moved interstate only...

2012-01-01

312

9 CFR 78.31 - Brucellosis reactor swine.  

Code of Federal Regulations, 2013 CFR

...2013-01-01 2013-01-01 false Brucellosis reactor swine. 78.31 Section 78.31 Animals...of Brucellosis § 78.31 Brucellosis reactor swine. (a) Destination. Brucellosis reactor swine may be moved interstate only for...

2013-01-01

313

9 CFR 78.7 - Brucellosis reactor cattle.  

...2014-01-01 2014-01-01 false Brucellosis reactor cattle. 78.7 Section 78.7 Animals...Because of Brucellosis § 78.7 Brucellosis reactor cattle. (a) Destination. Brucellosis reactor cattle may be moved interstate only...

2014-01-01

314

9 CFR 78.31 - Brucellosis reactor swine.  

Code of Federal Regulations, 2012 CFR

...2012-01-01 2012-01-01 false Brucellosis reactor swine. 78.31 Section 78.31 Animals...of Brucellosis § 78.31 Brucellosis reactor swine. (a) Destination. Brucellosis reactor swine may be moved interstate only for...

2012-01-01

315

Hydrogasification reactor and method of operating same  

DOEpatents

The present invention provides a system and method for evaluating effects of process parameters on hydrogasification processes. The system includes a hydrogasification reactor, a pressurized feed system, a hopper system, a hydrogen gas source, and a carrier gas source. Pressurized carbonaceous material, such as coal, is fed to the reactor using the carrier gas and reacted with hydrogen to produce natural gas.

Hobbs, Raymond; Karner, Donald; Sun, Xiaolei; Boyle, John; Noguchi, Fuyuki

2013-09-10

316

Selective purge for hydrogenation reactor recycle loop  

DOEpatents

Processes and apparatus for providing improved contaminant removal and hydrogen recovery in hydrogenation reactors, particularly in refineries and petrochemical plants. The improved contaminant removal is achieved by selective purging, by passing gases in the hydrogenation reactor recycle loop or purge stream across membranes selective in favor of the contaminant over hydrogen.

Baker, Richard W. (Palo Alto, CA); Lokhandwala, Kaaeid A. (Union City, CA)

2001-01-01

317

Profiling a reactor component using ultrasonics  

Microsoft Academic Search

The use of ultrasonic pulse echo ranging for profiling an inaccessible component in the complex and hostile structure of a nuclear reactor is described. The authors outline the ultrasonic method successfully employed in profiling a deformed guide tube of a reactor. The uniqueness of the method is the use of air transducers at 40 kHz, overcoming the inherent divergence problems

L. Pathak; V. R. Seshadri; C. Kumaravadivelu; G. Srinivasan; V. S. Raghunathan

1993-01-01

318

Fuel elements of research reactor CM  

SciTech Connect

In 1961 the CM research reactor was commissioned at the Research Institute of Atomic Reactors (Dimitrovgrad, Russia), it was intended to carry on investigations and the production of transuranium nuclides. The reactor is of a tank type. Original fuel assembly contained plate fuels that were spaced with vanes and corrugated bands. Nickel was used as a cladding material, fuel meat was produced from UO{sub 2} + electrolytic nickel composition. Fuel plates have been replaced by self-spacing cross-shaped dispersion fuels clad in stainless steel. In 2005 the reactor was updated. The purpose of this updating was to increase the quantity of irradiation channels in the reactor core and to improve the neutron balance. The updating was implemented at the expense of 20 % reduction in the quantity of fuel elements in the core which released a space for extra channels and decreased the mass of structural materials in the core. The updated reactor is loaded with modified standard fuel elements with 20 % higher uranium masses. At the same time stainless steel in fuel assembly shrouds was substituted by zirconium alloy. Today in progress are investigations and work to promote the second stage of reactor updating that involve developments of cross-shaped fuel elements having low neutron absorption matrix materials. This article gives an historical account of the design and main technical changes that occurred for the CM reactor since its commissioning.

Kozlov, A.V.; Morozov, A.V.; Vatulin, A.V.; Ershov, S.A. [Rogova St., 5A, P.O.B. 369, Moscow(Russian Federation)

2013-07-01

319

University of Virginia Reactor Facility Decommissioning Results  

SciTech Connect

The University of Virginia Reactor Facility started accelerated decommissioning in 2002. The facility consists of two licensed reactors, the CAVALIER and the UVAR. This paper will describe the progress in 2002, remaining efforts and the unique organizational structure of the project team.

Ervin, P. F.; Lundberg, L. A.; Benneche, P. E.; Mulder, R. U.; Steva, D. P.

2003-02-24

320

Fission reactor experiments for solid breeder blankets  

Microsoft Academic Search

The testing needs for solid breeder blanket development are different from those for liquid breeder blankets. In particular, a reasonable number of moderate volume test sites in a neutron environment are needed. Existing fission reactors are shown to be able to provide this environment with reasonable simulation of many important blanket conditions. Three major additional fission reactor tests are identified

P. J. Gierszewski; M. A. Abdou; R. Puigh

1986-01-01

321

Improved Fischer-Tropsch Slurry Reactors  

Microsoft Academic Search

The conversion of synthesis gas to hydrocarbons or alcohols involves highly exothermic reactions. Temperature control is a critical issue in these reactors for a number of reasons. Runaway reactions can be a serious safety issue, even raising the possibility of an explosion. Catalyst deactivation rates tend to increase with temperature, particularly of there are hot spots in the reactor. For

Andrew Lucero

2009-01-01

322

Reactor Antineutrinos Signal all over the world  

E-print Network

We present an updated estimate of reactor antineutrino signal all over the world, with particular attention to the sites proposed for existing and future geo-neutrino experiment. In our calculation we take into account the most updated data on Thermal Power for each nuclear plant, on reactor antineutrino spectra and on three neutrino oscillation mechanism.

B. Ricci; F. Mantovani; M. Baldoncini; J. Esposito; L. Ludhova; S. Zavatarelli

2014-03-17

323

Radiation effects in reactor structural alloys  

Microsoft Academic Search

Structural materials subjected to high doses of energetic neutron radiation in advanced fission reactors and planned fusion reactors experience profound changes in mechanical and physical properties. The important phenomena are swelling, creep, and embrittlement. Extensive programs of basic and applied radiation effects research are underway in the United States and in a number of other countries. This research has led

L. K. Mansur; E. E. Bloom

1982-01-01

324

Radiation effects in reactor structural alloys  

Microsoft Academic Search

Structural materials subjected to high doses of energetic neutron radiation in advanced fission reactors and in planned fusion reactors experience profound changes in mechanical and physical properties. The important phenomena are swelling, creep, and embrittlement. Extensive programs of basic and applied radiation effects research are underway in the United States and in a number of other countries. This research has

L. K. Mansur; E. E. Bloom

1982-01-01

325

INTERNATIONAL JOURNAL OF CHEMICAL REACTOR ENGINEERING  

E-print Network

INTERNATIONAL JOURNAL OF CHEMICAL REACTOR ENGINEERING Volume 6 2008 Article A7 SFGP 2007 manuscript, published in "International Journal of Chemical Reactor Engineering 6 (2008) A5" #12;SFGP 2007-dimensional transient numerical model has been developed with heat and mass transfer, fluid mechanics and de- tailed

Paris-Sud XI, Université de

326

The First Reactor, 40th Anniversary (rev.)  

SciTech Connect

This booklet, an updated version of the original booklet describing the first nuclear reactor, was written in honor of the 40th anniversary of the first reactor or "pile". It is based on firsthand accounts told to Corbin Allardice and Edward R. Trapnell, and includes recollections of Enrico and Laura Fermi.

Allardice, Corbin; Trapnell, Edward R.; Fermi, Enrico; Fermi, Laura; Williams, Robert C.

1982-12-01

327

Standards for reference reactor physics measurements  

Microsoft Academic Search

Reactor physics analysis methods require experimental testing and confirmation over the range of practical reactor configurations and states. This range is somewhat limited by practical fuel types such as actinide oxides or carbides enclosed in metal cladding. On the other hand, this range continues to broaden because of the trend of using higher enrichment, if only slightly enriched, electric utility

D. R. Harris; D. M. Cokinos; V. Uotinen

1990-01-01

328

The Economics of Fast Breeder Reactors  

Microsoft Academic Search

The overall status of the fast breeder reactor (FBR) system is periodically reviewed in France. In 1983, a report was prepared on the status and prospects of the FBR system at the request of the then Minister of Industry. Five years later, Electricite de France (EdF) and the French Atomic Energy Commission (CEA) jointly updated this report. The FBR reactor

M. Rapin; F. J. Barclay; R. H. Allardice

1990-01-01

329

Transitional arc discharges-reactor design & applications  

Microsoft Academic Search

Summary form only given. This present work reports new reactor designs to achieve the transient GA discharge in cylindrical 'tornado' geometry with a high degree of non-equilibrium and simultaneously high energy density. Reverse Vortex Flow (RVF) or 'Tornado' provided the high velocity necessary to move GA discharge over the electrodes with long residence time inside reactor, recirculation of active species,

C. S. Kalra; Y. I. Cho; A. Gutsol; A. Fridman

2004-01-01

330

Membrane reactor for enzymatic hydrolysis of cellobiose  

Microsoft Academic Search

A pressurized, stirred vessel attached with an ultrafiltration membrane was used as a membrane reactor. Cellobiose hydrolysis by cellobiase was carried out and theoretically analyzed in terms of steady-state conversion and flow rate through the membrane. When the flow rate exceeds a critical value, a significant fraction of the enzyme inside the reactor is localized in the concentration polarization layer

J. Hong; G. T. Tsao; P. C. Wankat

1981-01-01

331

ADVANCED HIGH FLUX RESEARCH REACTOR TECHNOLOGY  

Microsoft Academic Search

The development of high-flux research reactors is discussed. The major ; factor limiting further advance, restricted power density, is described. One ; method of improvement considered is designing for specific experimental use. ; Limitations were investigated in a comparative study of HâO, DâO, and ; Na cooled reactors of annularcore type. It is shown that power densities can be ;

C. N. Kelber; B. I. Spinrad; L. J. Templin

1962-01-01

332

Method of Operating a Neutronic Reactor.  

National Technical Information Service (NTIS)

A method is disclosed for operating a reactor having an active portion of a given length between a charging end and a discharging end, a first end region of the reactor extending from the charging end for one-quarter to one-third of said given length, a s...

W. K. Woods

1976-01-01

333

Steady-state spheromak reactor studies  

SciTech Connect

After summarizing the essential elements of a gun-sustained spheromak, the potential for a steady-state reactor is explored by means of a comprehensive physics/engineering/costing model. A range of cost-optimized reactor design points is presented, and the sensitivity of cost to key physics, engineering, and operational variables is reported.

Hagenson, R.L.; Krakowski, R.A.

1985-07-01

334

Reactor accelerator coupling experiments: a feasability study  

E-print Network

of the final design MCNP geometry for UT?NETL reactor accelerated coupled experiment.......................................35 Fig. 15. keff versus control rod positions... of the final design MCNP geometry for UT?NETL reactor accelerated coupled experiment.......................................35 Fig. 15. keff versus control rod positions...

Woddi Venkat Krishna, Taraknath

2006-08-16

335

Magnetic switch for reactor control rod  

Microsoft Academic Search

This patent describes a control rod system for a nuclear reactor utilizing an electromagnetic grapple mechanism for holding and releasing a control rod, the improvement comprising a magnetic reed switch assembly having a Curie-point magnetic shunt and responsive to reactor coolant temperature for short circuiting the electromagnetic grapple mechanism causing release of the control rod when the coolant temperature reaches

Germer

1986-01-01

336

Fluidized bed biofilm reactor for wastewater treatment  

Microsoft Academic Search

The fluidized bed biofilm reactor (FBBR) represents a recent innovation in biofilm processes. Immobilization of microorganisms on the small, fluidized particles of the medium results in a high reactor biomass holdup which enables the process to be operated at significantly higher liquid throughputs with the practical absence of biomass wash-out. The process intensification (i.e., a reduction in process size while

Wen K. Shieh; John D. Keenan

337

Topaz-II reactor control unit development  

SciTech Connect

The development for a new digital reactor control unit for the Topaz-II reactor is described. The unit is expected to provide the means for automated control during a possible Topaz flight experiment. The breadboard design and development is discussed.

Wyant, F.J.; Jensen, D.; Logothetis, J.

1994-12-31

338

Reactor Design from a Stability Viewpoint.  

ERIC Educational Resources Information Center

This course uses stability as a central theme around which to organize a wide range of reactor concerns. This approach brings together the subject matter of catalyst particles with that of well-stirred vessels and tubular reactor geometry. (Author/BB)

Perlmutter, D. D.

1978-01-01

339

Explosive demolition of K East Reactor Stack  

ScienceCinema

Using $420,000 in Recovery Act funds, the Department of Energy and contractor CH2M HILL Plateau Remediation Company topped off four months of preparations when they safely demolished the exhaust stack at the K East Reactor and equipment inside the reactor building on July 23, 2010.

None

2010-09-02

340

Actinide burning in the integral fast reactor  

Microsoft Academic Search

During the past few years, Argonne National Laboratory has been developing the integral fast reactor (IFR), an advanced liquid-metal reactor concept. In the IFR, the inherent properties of liquid-metal cooling are combined with a new metallic fuel and a radically different refining process to allow breakthroughs in passive safety, fuel cycle economics, and waste management. A key feature of the

1993-01-01

341

TRIGA Mark-II, III Reactor Operation.  

National Technical Information Service (NTIS)

TRIGA Mark-II reactor has been primarily utilized as usual for the fundamental reactor experiments for university students. The annual operating time is 1,100 hours and the gross thermal output is 17,159 KWH, having consumed 0.88g of U-235. The reconstuct...

J. B. Lee, C. K. Lee, B. J. Chun, Y. J. Kim, G. Y. Han

1982-01-01

342

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

NASA Technical Reports Server (NTRS)

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.

Lantz, E.; Mayo, W.

1972-01-01

343

Cooling system for a nuclear reactor  

DOEpatents

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

Amtmann, Hans H. (Rancho Santa Fe, CA)

1982-01-01

344

Self-actuating reactor shutdown system  

DOEpatents

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

Barrus, Donald M. (San Jose, CA); Brummond, Willian A (Livermore, CA); Peterson, Leslie F. (Danville, CA)

1988-01-01

345

ICP Reactor Modeling: CF4 Discharge  

NASA Technical Reports Server (NTRS)

Inductively coupled plasma (ICP) reactors are widely used now for etching and deposition applications due to their simpler design compared to other high density sources. Plasma reactor modeling has been playing an important role since it can, in principle, reduce the number of trial and error iterations in the design process and provide valuable understanding of mechanisms. Fluorocarbon precursors have been the choice for oxide etching. We have data available on CF4 from our laboratory. These are current voltage characteristics, La.ngmuir probe data, UV-absorption, and mass spectrometry measurements in a GEC-ICP reactor. We have developed a comprehensive model for ICP reactors which couples plasma generation and transport and neutral species dynamics with the gas flow equations. The model has been verified by comparison with experimental results for a nitrogen discharge in an ICP reactor. In the present work, the model has been applied to CF4 discharge and compared to available experimental data.

Bose, Deepak; Govindan, T. R.; Meyyappan, M.

1999-01-01

346

[Fission Working Group -- Molten salt reactors  

SciTech Connect

This report provides an assessment of molten salt reactors (MSRs) which are fluid fuel reactors and, as such, have several unique features, some which are important to the burning of fissile material from dismantled weapons. This material can be added on-line during operation in either continuous or batch form. The added fuel need only be in an acceptable chemical form, but no fuel manufacturing or minimum discrete amounts for a fuel element are required. Fluid fuel reactors can have partial or full on-line fuel processing. When online fuel processing is utilized, a particular fuel component, for example the plutonium, can be burned completely, or in some sense can be converted to other kinds of fuel, for example into [sup 233]U. There is no equivalent of fuel burnup in continuous processing reactors, and no need for reprocessing in external plants and manufacturing of fuel elements, transportation, and reinsertion in the reactor.

Gat, U.; Engel, J.R.

1992-01-01

347

[Fission Working Group -- Molten salt reactors  

SciTech Connect

This report provides an assessment of molten salt reactors (MSRs) which are fluid fuel reactors and, as such, have several unique features, some which are important to the burning of fissile material from dismantled weapons. This material can be added on-line during operation in either continuous or batch form. The added fuel need only be in an acceptable chemical form, but no fuel manufacturing or minimum discrete amounts for a fuel element are required. Fluid fuel reactors can have partial or full on-line fuel processing. When online fuel processing is utilized, a particular fuel component, for example the plutonium, can be burned completely, or in some sense can be converted to other kinds of fuel, for example into {sup 233}U. There is no equivalent of fuel burnup in continuous processing reactors, and no need for reprocessing in external plants and manufacturing of fuel elements, transportation, and reinsertion in the reactor.

Gat, U.; Engel, J.R.

1992-12-31

348

TREAT Reactor Control and Protection System  

SciTech Connect

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

Lipinski, W.C.; Brookshier, W.K.; Burrows, D.R.; Lenkszus, F.R.; McDowell, W.P.

1985-01-01

349

Advanced PPA Reactor and Process Development  

NASA Technical Reports Server (NTRS)

Design and development of a second generation Plasma Pyrolysis Assembly (PPA) reactor is currently underway as part of NASA s Atmosphere Revitalization Resource Recovery effort. By recovering up to 75% of the hydrogen currently lost as methane in the Sabatier reactor effluent, the PPA helps to minimize life support resupply costs for extended duration missions. To date, second generation PPA development has demonstrated significant technology advancements over the first generation device by doubling the methane processing rate while, at the same time, more than halving the required power. One development area of particular interest to NASA system engineers is fouling of the PPA reactor with carbonaceous products. As a mitigation plan, NASA MSFC has explored the feasibility of using an oxidative plasma based upon metabolic CO2 to regenerate the reactor window and gas inlet ports. The results and implications of this testing are addressed along with the advanced PPA reactor development work.

Wheeler, Raymond; Aske, James; Abney, Morgan B.; Miller, Lee A.; Greenwood, Zachary

2012-01-01

350

Hot Gas Desulfurization Using Transport Reactors  

SciTech Connect

Sierra Pacific Power Company is building a 100 MW, IGCC power plant based on KRW fluid bed gasifier technology that utilizes transport reactors for hot gas desulfurization and sorbent regeneration. Use of a transport absorber avoids the need for pre-filtration of dust-laden gasifier effluent, while a transport regenerator allows for the use of 100% air without the need for heat exchange equipment. Selection of transport reactors for hot gas desulfurization using a proprietary sorbent, based on testing performed in a transport reactor test unit (TRTU) at the M. W. Kellogg Technology Development Center and in a fixed bed reactor at Morgantown Energy Technology Center (METC), is outlined. The results obtained in these two test facilities and reasons for selecting transport reactors for the IGCC power plant in preference to either fixed bed or fluidized bed reactors are discussed. This paper reviews the evolution of the hot gas desulfurization system designs and includes selected results on H{sub 2}S absorption and regeneration of sulfided sorbent over several absorption/regeneration cycles conducted in the TRTU and the METC fixed bed reactor. The original design for the Sierra Pacific Project was based on fixed bed reactors with zinc ferrite as the sorbent. Owing to the high steam requirements of this sorbent, zinc titanate was selected and tested in a fixed bed reactor and was found unacceptable due to loss of strength on cyclic absorption/regeneration operation. Another sorbent evaluated was Z-Sorb{reg_sign}, a proprietary sorbent developed by Phillips Petroleum Company, was found to have excellent sulfur capacity, structural strength and regenerability. Steam was found unsuitable as fixed bed regenerator diluent, this results in a requirement for a large amount of inert gas, whereas a transport regenerator requires no diluent. The final Sierra design features transport reactors for both desulfurization and regeneration steps using neat air. 3 refs., 3 figs., 2 tabs.

Moorehead, E.L. [Kellogg (M.W.) Technology Co., Houston, TX (United States)

1996-12-31

351

NAVAL REACTORS--Continued Program and Financing--Continued  

E-print Network

requirements. Due to the crucial nature of nuclear reactor work, Naval Reactors is a centrally managed personnel benefits ........................................................12.1 221Travel and transportation

352

Search for sterile neutrinos at reactors with a small core  

E-print Network

The sensitivity to the sterile neutrino mixing at very short baseline reactor neutrino experiments is investigated. If the reactor core is relatively large as in the case of commercial reactors, then the sensitivity is lost for $\\Delta m^2 \\gtrsim$ 1 eV$^2$ due to smearing of the reactor core size. If the reactor core is small as in the case of the experimental fast neutron reactor Joyo, the ILL research reactor or the Osiris reactor, on the other hand, then sensitivity to $\\sin^22\\theta_{14}$ can be as good as 0.03 for $\\Delta m^2 \\sim$ several eV$^2$ because of its small size.

Osamu Yasuda

2011-10-12

353

(Meeting on fusion reactor materials)  

SciTech Connect

During his visit to the KfK, Karlsruhe, F. W. Wiffen attended the IEA 12th Working Group Meeting on Fusion Reactor Materials. Plans were made for a low-activation materials workshop at Culham, UK, for April 1991, a data base workshop in Europe for June 1991, and a molecular dynamics workshop in the United States in 1991. At the 11th IEA Executive Committee on Fusion Materials, discussions centered on the recent FPAC and Colombo panel review in the United States and EC, respectively. The Committee also reviewed recent progress toward a neutron source in the United States (CWDD) and in Japan (ESNIT). A meeting with D. R. Harries (consultant to J. Darvas) yielded a useful overview of the EC technology program for fusion. Of particular interest to the US program is a strong effort on a conventional ferritic/martensitic steel for fist wall/blanket operation beyond NET/ITER.

Jones, R.H. (Pacific Northwest Lab., Richland, WA (USA)); Klueh, R.L.; Rowcliffe, A.F.; Wiffen, F.W. (Oak Ridge National Lab., TN (USA)); Loomis, B.A. (Argonne National Lab., IL (USA))

1990-11-01

354

Reactor operation environmental information document  

SciTech Connect

The Savannah River Site (SRS) is a large United States Department of Energy installation on the upper Atlantic Coastal Plain of South Carolina. The SRS contains diverse habitats, flora, and fauna. Habitats include upland terrestrial areas, varied wetlands including Carolina Bays, the Savannah River swamp system, and impoundment related and riparian wetlands, and the aquatic habitats of several stream systems, two large cooling reservoirs, and the Savannah River. These diverse habitats support a large variety of plants and animals including many commercially or recreational valuable species and several rare, threatened or endangered species. This volume describes the major habitats and their biota found on the SRS, and discuss the impacts of continued operation of the K, L, and P production reactors.

Wike, L.D.; Specht, W.L.; Mackey, H.E.; Paller, M.H.; Wilde, E.W.; Dicks, A.S.

1989-12-01

355

Particle transport in plasma reactors  

SciTech Connect

SEMATECH and the Department of Energy have established a Contamination Free Manufacturing Research Center (CFMRC) located at Sandia National Laboratories. One of the programs underway at the CFMRC is directed towards defect reduction in semiconductor process reactors by the application of computational modeling. The goal is to use fluid, thermal, plasma, and particle transport models to identify process conditions and tool designs that reduce the deposition rate of particles on wafers. The program is directed toward defect reduction in specific manufacturing tools, although some model development is undertaken when needed. The need to produce quantifiable improvements in tool defect performance requires the close cooperation among Sandia, universities, SEMATECH, SEMATECH member companies, and equipment manufacturers. Currently, both plasma (e.g., etch, PECVD) and nonplasma tools (e.g., LPCVD, rinse tanks) are being worked on under this program. In this paper the authors summarize their recent efforts to reduce particle deposition on wafers during plasma-based semiconductor manufacturing.

Rader, D.J.; Geller, A.S.; Choi, Seung J. [Sandia National Labs., Albuquerque, NM (United States); Kushner, M.J. [Illinois Univ., Urbana, IL (United States)

1995-01-01

356

Power reactor events and issues  

SciTech Connect

This publication reviews selected operating events that have occurred at nuclear power plants and presents the results of NRC-sponsored analysis of pertinent operating issues. This study was initiated following several instances of water hammer involving the service water system at Arkansas Nuclear One. The task was to evaluate the need to reissue previous NRC guidance about water hammer or to suggest additional measures to prevent or mitigate their occurrence. Twelve from PWR and BWR type reactors were studied. This study concluded that the frequency of reported water hammer occurrences continues to drop and no new phenomena were identified as causes of water hammer. In addition, this study supports prior NRC conclusions regarding water hammer; however, some aspects that could impact safety and were identified in this study had not been previously emphasized.

Not Available

1992-10-01

357

History of critical experiments at Pajarito Site  

SciTech Connect

This account describes critical and subcritical assemblies operated remotely at the Pajarito Canyon Site at the Los Alamos National Laboratory. Earliest assemblies, directed exclusively toward the nuclear weapons program, were for safety tests. Other weapon-related assemblies provided neutronic information to check detailed weapon calculations. Topsy, the first of these critical assemblies, was followed by Lady Godiva, Jezebel, Flattop, and ultimately Big Ten. As reactor programs came to Los Alamos, design studies and mockups were tested at Pajarito Site. For example, nearly all 16 Rover reactors intended for Nevada tests were preceded by zero-power mockups and proof tests at Pajarito Site. Expanded interest and capability led to fast-pulse assemblies, culminating in Godiva IV and Skua, and to the Kinglet and Sheba solution assemblies.

Paxton, H.C.

1983-03-01

358

Photoneutron effects on pulse reactor kinetics for the Annular Core Research Reactor (ACRR).  

SciTech Connect

The Annular Core Research Reactor (ACRR) is a swimming-pool type pulsed reactor that maintains an epithermal neutron flux and a nine-inch diameter central dry cavity. One of its uses is neutron and gamma-ray irradiation damage studies on electronic components under transient reactor power conditions. In analyzing the experimental results, careful attention must be paid to the kinetics associated with the reactor to ensure that the transient behavior of the electronic device is understood. Since the ACRR fuel maintains a substantial amount of beryllium, copious quantities of photoneutrons are produced that can significantly alter the expected behavior of the reactor power, especially following a reactor pulse. In order to understand these photoneutron effects on the reactor kinetics, the KIFLE transient reactor-analysis code was modified to include the photoneutron groups associated with the beryllium. The time-dependent behavior of the reactor power was analyzed for small and large pulses, assuming several initial conditions including following several pulses during the day, and following a long steady-state power run. The results indicate that, for these types of initial conditions, the photoneutron contribution to the reactor pulse energy can have a few to tens of percent effect.

Parma, Edward J., Jr.

2009-06-01

359

Reactor Monitoring with Antineutrinos - A Progress Report  

NASA Astrophysics Data System (ADS)

The Reactor Safeguards regime is the name given to a set of protocols and technologies used to monitor the consumption and production of fissile materials in nuclear reactors. The Safeguards regime is administered by the International Atomic Energy Agency (IAEA), and is an essential component of the global Treaty on Nuclear Nonproliferation, recently renewed by its 189 remaining signators. (The 190th, North Korea, withdrew from the Treaty in 2003). Beginning in Russia in the 1980s, a number of researchers worldwide have experimentally demonstrated the potential of cubic meter scale antineutrino detectors for non-intrusive real-time monitoring of fissile inventories and power output of reactors. The detectors built so far have operated tens of meters from a reactor core, outside of the containment dome, largely unattended and with remote data acquisition for an entire 1.5 year reactor cycle, and have achieved levels of sensitivity to fissile content of potential interest for the IAEA safeguards regime. In this article, I will describe the unique advantages of antineutrino detectors for cooperative monitoring, consider the prospects and benefits of increasing the range of detectability for small reactors, and provide a partial survey of ongoing global research aimed at improving near-field and far field monitoring and discovery of nuclear reactors.

Bernstein, Adam

2012-08-01

360

Flexible Conversion Ratio Fast Reactor Systems Evaluation  

SciTech Connect

Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores reated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcme the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better themal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor.

Neil Todreas; Pavel Hejzlar

2008-06-30

361

Simulation and parameter estimation in heterogeneous two and three phase reactor modelling.  

National Technical Information Service (NTIS)

Generalized models for catalytic packed bed reactors, trickle bed reactors, slurry reactors and gas-liquid tank reactors are presented. For the catalytic packed bed reactor the pseudo- homogeneous and heterogeneous one-dimensional models were considered. ...

J. Waernaa

1994-01-01

362

Packed fluidized bed blanket for fusion reactor  

DOEpatents

A packed fluidized bed blanket for a fusion reactor providing for efficient radiation absorption for energy recovery, efficient neutron absorption for nuclear transformations, ease of blanket removal, processing and replacement, and on-line fueling/refueling. The blanket of the reactor contains a bed of stationary particles during reactor operation, cooled by a radial flow of coolant. During fueling/refueling, an axial flow is introduced into the bed in stages at various axial locations to fluidize the bed. When desired, the fluidization flow can be used to remove particles from the blanket.

Chi, John W. H. (Mt. Lebanon, PA)

1984-01-01

363

Safety design of prototype fast breeder reactor  

SciTech Connect

The basic design and safety design of Prototype Fast Breeder Reactor (PFBR) is presented. Design aspects covered include safety classification, seismic categorization, design basis conditions, design safety limits, core physics, core monitoring, shutdown system, decay heat removal system, protection against sodium leaks and tube leaks in steam generator, plant layout, radiation protection, event analysis, beyond design basis accidents, integrity of primary containment, reactor containment building and design pressure resulting from core disruptive accident. The measures provided in the design represent a robust case of the safety of the reactor. (authors)

Bhoje, S.B.; Chetal, S.C.; Singh, Om Pal [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

2004-07-01

364

Cooldown criteria for light water reactors  

SciTech Connect

This Standard provides design criteria for systems and equipment necessary for the performance of those nuclear safety functions that are required to achieve and maintain a safe shutdown of the reactor to cold shutdown conditions from a hot standby or post accident condition. The following nuclear safety functions are addressed in this Standard: (1) reactivity control; (2) reactor coolant system heat removal; and (3) reactor coolant system integrity, i.e., pressure control, and inventory control. This Standard provides criteria for the design of systems used to accomplish these functions.

Not Available

1983-01-01

365

Review of Reactor Neutrino Oscillation Experiments  

NASA Astrophysics Data System (ADS)

In this document we will review the current status of reactor neutrino oscillation experiments and present their physics potentials for measuring the ?13 neutrino mixing angle. The neutrino mixing angle ?13 is currently a high-priority topic in the field of neutrino physics. There are currently three different reactor neutrino experiments, DOUBLE CHOOZ, DAYA BAY and RENO and a few accelerator neutrino experiments searching for neutrino oscillations induced by this angle. A description of the reactor experiments searching for a nonzero value of ?13 is given, along with a discussion of the sensitivities that these experiments can reach in the near future.

Mariani, C.

366

DEVELOPMENTS IN AREAS OF ON-LINE FISSION-YIELDAND DIRECT MAW MEASUREMENTS AT THE LOS AIJIMOSSCIENTIFIC LABOWTORYf  

Microsoft Academic Search

ABSTUCT On-1ine studies of 238 U fission yields with fission spectrum neutrons were made with the Bernas surface-ionizationtechnique at the Godiva IV burst reactor facility. The target was 300 mg of 238 U in a porous graphite mixture. Ion emission rates were measurd by Z-direction motion of the collector. The isotope collection time was varied from 0.1 to 4.0 seconds

S. J. Balestrini

367

TREATMENT OF METHANOLIC WASTEWATER BY ANAEROBIC DOWN-FLOW HANGING SPONGE (ANDHS) REACTOR AND UASB REACTOR  

NASA Astrophysics Data System (ADS)

Anaerobic down-flow hanging sponge (AnDHS) reactor and UASB reactor were operated at 30℃ for over 400 days in order to investigate the process performance and the sludge characteristics of treating methanolic wastewater (2 gCOD/L). The settings OLR of AnDHS reactor and of UASB reactor were 5.0 -10.0 kgCOD/m3/d and 5.0 kgCOD/m3/d. The average of the COD removal demonstrated by both reactors were over 90% throughout the experiment. From the results of methane producing activities and the PCR-DGGE method, most methanol was directly converted to methane in both reactors. The conversion was carried out by different methanogens: one closely related to Methanomethylovorans hollandica in the AnDHS retainted sludge and the other closely related to Methanosarcinaceae and Metanosarciales in the UASB retainted sludge.

Sumino, Haruhiko; Wada, Keiji; Syutsubo, Kazuaki; Yamaguchi, Takashi; Harada, Hideki; Ohashi, Akiyoshi

368

Passive compact molten salt reactor (PCMSR), modular thermal breeder reactor with totally passive safety system  

NASA Astrophysics Data System (ADS)

Design Study Passive Compact Molten Salt Reactor (PCMSR) with totally passive safety system has been performed. The term of Compact in the PCMSR name means that the reactor system is designed to have relatively small volume per unit power output by using modular and integral concept. In term of modular, the reactor system consists of three modules, i.e. reactor module, turbine module and fuel management module. The reactor module is an integral design that consists of reactor, primary and intermediate heat exchangers and passive post shutdown cooling system. The turbine module is an integral design of a multi heating, multi cooling, regenerative gas turbine. The fuel management module consists of all equipments related to fuel preparation, fuel reprocessing and radioactive handling. The preliminary calculations show that the PCMSR has negative temperature and void reactivity coefficient, passive shutdown characteristic related to fuel pump failure and possibility of using natural circulation for post shutdown cooling system.

Harto, Andang Widi

2012-06-01

369

Small Modular Reactors: Institutional Assessment  

SciTech Connect

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

Joseph Perkowski, Ph.D.

2012-06-01

370

Cross-flow electrochemical reactor cells, cross-flow reactors, and use of cross-flow reactors for oxidation reactions  

DOEpatents

This invention discloses cross-flow electrochemical reactor cells containing oxygen permeable materials which have both electron conductivity and oxygen ion conductivity, cross-flow reactors, and electrochemical processes using cross-flow reactor cells having oxygen permeable monolithic cores to control and facilitate transport of oxygen from an oxygen-containing gas stream to oxidation reactions of organic compounds in another gas stream. These cross-flow electrochemical reactors comprise a hollow ceramic blade positioned across a gas stream flow or a stack of crossed hollow ceramic blades containing a channel or channels for flow of gas streams. Each channel has at least one channel wall disposed between a channel and a portion of an outer surface of the ceramic blade, or a common wall with adjacent blades in a stack comprising a gas-impervious mixed metal oxide material of a perovskite structure having electron conductivity and oxygen ion conductivity. The invention includes reactors comprising first and second zones seprated by gas-impervious mixed metal oxide material material having electron conductivity and oxygen ion conductivity. Prefered gas-impervious materials comprise at least one mixed metal oxide having a perovskite structure or perovskite-like structure. The invention includes, also, oxidation processes controlled by using these electrochemical reactors, and these reactions do not require an external source of electrical potential or any external electric circuit for oxidation to proceed.

Balachandran, Uthamalingam (Hinsdale, IL); Poeppel, Roger B. (Glen Ellyn, IL); Kleefisch, Mark S. (Naperville, IL); Kobylinski, Thaddeus P. (Lisle, IL); Udovich, Carl A. (Joliet, IL)

1994-01-01

371

Design Parameters for an Efficient Accelerator Driven Thermal Nuclear Reactor  

Microsoft Academic Search

Generally, accelerator driven reactors have been designed for further using the light water reactor waste products. In this work, an accelerator driven thermal thorium reactor (ADTTR) has been designed in which thorium is used as fuel for such a reactor. Thorium is abundantly found in the crust of the Earth. An ADTTR could produce 2 to 5 times the energy

Mike A. Shubov; M. A. K. Lodhi

2003-01-01

372

Heavy water accountancy for research reactors in JAERI.  

National Technical Information Service (NTIS)

The three research reactors have been operated by the Department of Research Reactor and used about 41 tons heavy water as coolant, moderator and reflector of research reactors. The JRR-2 is a tank type research reactor of 10MW in thermal power and its is...

T. Yoshijima, S. Tanaka, D. Nemoto

1998-01-01

373

Uncertainties in the Anti-neutrino Production at Nuclear Reactors  

E-print Network

Anti-neutrino emission rates from nuclear reactors are determined from thermal power measurements and fission rate calculations. The uncertainties in these quantities for commercial power plants and their impact on the calculated interaction rates in electron anti-neutrino detectors is examined. We discuss reactor-to-reactor correlations between the leading uncertainties and their relevance to reactor anti-neutrino experiments.

Z. Djurcic; J. A. Detwiler; A. Piepke; V. R. Foster Jr.; L. Miller; G. Gratta

2008-08-06

374

99. ARAIII. Overall view of drilling area in reactor pit. ...  

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

99. ARA-III. Overall view of drilling area in reactor pit. Bridge over pit in use for operations. Shows water in pool, reactor, hoist, operators, and general view of interior of reactor pit area. August 12, 1963. Ineel photo no. 63-4454. Photographer: Benson. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

375

Completely autotrophic nitrogen removal over nitrite in one single reactor  

Microsoft Academic Search

The microbiology and the feasibility of a new, single-stage, reactor for completely autotrophic ammonia removal were investigated. The reactor was started anoxically after inoculation with biomass from a reactor performing anaerobic ammonia oxidation (Anammox). Subsequently, oxygen was supplied to the reactor and a nitrifying population developed. Oxygen was kept as the limiting factor. The development of a nitrifying population was

A. Olav Sliekers; N. Derwort; J. L. Campos Gomez; M. Strous; J. G. Kuenen; M. S. M. Jetten

2002-01-01

376

Research Reactors and Radiation Facilities for Joint Use Program  

E-print Network

Research Reactors and Radiation Facilities for Joint Use Program Kyoto University Research Reactor at the Hida Observatory The Kyoto University Research Reactor Institute (KURRI) was established in 1963 of nuclear energy and radiation application. The main facility, called the Kyoto University Research Reactor

Takada, Shoji

377

LIMITED POWER BURSTS IN DISTRIBUTED MODELS OF NUCLEAR REACTORS  

E-print Network

LIMITED POWER BURSTS IN DISTRIBUTED MODELS OF NUCLEAR REACTORS M. V. Bazhenov and E. F. Sabaev UDC employed for analyzing reactor dynamics. Equations of this type are used for analyzing the stability of the reactor power, etc. Among these problems the question of the boundedness of reactor power bursts

Bazhenov, Maxim

378

PIPING FOR COOLANT WATER IS INSTALLED INSIDE REACTOR STRUCTURE PRIOR ...  

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

PIPING FOR COOLANT WATER IS INSTALLED INSIDE REACTOR STRUCTURE PRIOR TO EMBEDMENT IN CONCRETE. HIGHER PIPE IS INLET; THE OTHER, THE OUTLET LOOP. INLET PIPE WILL CONNECT TO TOP SECTION OF REACTOR VESSEL. INL NEGATIVE NO. 1287. Unknown Photographer, 1/18/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

379

Nuclear fuel performance in boiling water reactors  

SciTech Connect

A major development program is described to improve the performance of Boiling Water Reactor fuel. This sustained program is described in four parts: 1) performance monitoring, 2) fuel design changes, 3) plant operating recommendations, and 4) advanced fuel programs.

Elkins, R.B.; Baily, W.E.; Proebstle, R.A.; Armijo, J.S.; Klepfer, H.H.

1981-08-01

380

Computational evaluation of two reactor benchmark problems  

E-print Network

A neutronic evaluation of two reactor benchmark problems was performed. The benchmark problems describe typical PWR uranium and plutonium (mixed oxide) fueled lattices. WIMSd4m, a neutron transport lattice code, was used to evaluate multigroup...

Cowan, James Anthony

2012-06-07

381

The Nuclear Regulatory Commission's Reactor Safety Study  

Microsoft Academic Search

The Reactor Safety Study (WASH-1400), originally commissioned by the United States Atomic Energy Commission and published in final form under the auspices of the Nuclear Regulatory Commission, is reviewed below. In order to be useful for policy

Joel Yellin

1976-01-01

382

Advanced Test Reactor National Scientific User Facility  

SciTech Connect

The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

Frances M. Marshall; Jeff Benson; Mary Catherine Thelen

2011-08-01

383

Validation of the reactor dynamics code TRAB  

NASA Astrophysics Data System (ADS)

The validation of the one dimensional reactor dynamics code TRAB (Transient Analysis code for BWR's) is summarized. TRAB was validated with benchmark problems, comparative calculations against independent analyses, analyses of start up experiments of nuclear power plants, and real plant transients. The initial power excursion of the Chernobyl reactor accident was calculated with TRAB. TRAB was originally designed for BWR analyses, but it can in its present version be used for various modeling purposes. The core model of TRAB can be used separately for LWR calculations. For PWR modeling the core model of TRAB was coupled to circuit model SMABRE to form the SMATRA code. The versatile modeling capabilities of TRAB were used in analyses of e.g., the heating reactor SECURE and the RBMK type reactor (Chernobyl).

Raety, Hanna; Kyrki-Rajamaeki, Riitta; Rajamaeki, Markku

1991-05-01

384

High Temperature Gas Reactors Briefing to  

E-print Network

Meltdown-Proof Advanced Reactor and Gas Turbine #12;TRISO Fuel Particle -- "Microsphere" · 0.9mm diameter · TRISO acts as a pressure vessel · Reliability ­ Defective coatings during manufacture ­ ~ 1 defect

385

Rethinking the light water reactor fuel cycle  

E-print Network

The once through nuclear fuel cycle adopted by the majority of countries with operating commercial power reactors imposes a number of concerns. The radioactive waste created in the once through nuclear fuel cycle has to ...

Shwageraus, Evgeni, 1973-

2004-01-01

386

Actinide minimization using pressurized water reactors  

E-print Network

Transuranic actinides dominate the long-term radiotoxity in spent LWR fuel. In an open fuel cycle, they impose a long-term burden on geologic repositories. Transmuting these materials in reactor systems is one way to ease ...

Visosky, Mark Michael

2006-01-01

387

Self-Sustaining Thorium Boiling Water Reactors  

E-print Network

A thorium-fueled water-cooled reactor core design approach that features a radially uniform composition of fuel rods in stationary fuel assembly and is fuel-self-sustaining is described. This core design concept is similar ...

Ganda, Francesco

388

Gaseous fuel reactors for power systems  

NASA Technical Reports Server (NTRS)

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

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

1977-01-01

389

Heat pipe reactors for space power applications  

NASA Technical Reports Server (NTRS)

A family of heat pipe reactors design concepts has been developed to provide heat to a variety of electrical conversion systems. Three power plants are described that span the power range 1-500 kWe and operate in the temperature range 1200-1700 K. The reactors are fast, compact, heat-pipe cooled, high-temperature nuclear reactors fueled with fully enriched refractory fuels, UC-ZrC or UO2. Each fuel element is cooled by an axially located molybdenum heat pipe containing either sodium or lithium vapor. Virtues of the reactor designs are the avoidance of single-point failure mechanisms, the relatively high operating temperature, and the expected long lifetimes of the fuel element components.

Koenig, D. R.; Ranken, W. A.; Salmi, E. W.

1977-01-01

390

Physics design of the upgraded TREAT reactor  

SciTech Connect

With the deferral of the Safety Test Facility (STF), the TREAT Upgrade (TU) reactor has assumed a lead role in the US LMFBR safety test program for the foreseeable future. The functional requirements on TU require a significant enhancement of the capability of the current TREAT reactor. A design of the TU reactor has been developed that modifies the central 11 x 11 fuel assembly array of the TREAT reactor such as to provide the increased source of hard spectrum neutrons necessary to meet the functional requirements. A safety consequence of the increased demands on TU is that the self limiting operation capability of TREAT has proved unattainable, and reliance on a safety grade Plant Protection System is necessary to ensure that no clad damage occurs under postulated low-probability reactivity accidents. With that constraint, the physics design of TU provides a means of meeting the functional requirements with a high degree of confidence.

Bhattacharyya, S.K.; Lell, R.M.; Liaw, J.R.; Ulrich, A.J.; Wade, D.C.; Yang, S.T.

1980-01-01

391

Heterogeneous effects in fast breeder reactors  

E-print Network

Heterogeneous effects in fast breeder reactors are examined through development of simple but accurate models for the calculation of a posteriori corrections to a volume-averaged homogeneous representation. Three distinct ...

Gregory, Michael Vladimir

1973-01-01

392

Psychosocial effects of restarting a TMI reactor  

SciTech Connect

ORNL is studying human responses to hazardous environmental phenomena. This study attempts to understand the human behavior associated with the restart of TMI-1 Reactor after a nuclear event occurred at TMI-2.

Not Available

1990-01-01

393

Radial power flattening in sodium fast reactors  

E-print Network

In order to improve a new design for a uranium startup sodium cooled fast reactor which was proposed at MIT, this thesis evaluated radial power flattening by varying the fuel volume fraction at a fixed U-235 enrichment of ...

Krentz-Wee, Rebecca (Rebecca Elizabeth)

2012-01-01

394

Producing tritium in a homogenous reactor  

DOEpatents

A method and apparatus are described for the joint production and separation of tritium. Tritium is produced in an aqueous homogenous reactor and heat from the nuclear reaction is used to distill tritium from the lower isotopes of hydrogen.

Cawley, William E. (Richland, WA)

1985-01-01

395

Reactor physics project progress report no. 2  

E-print Network

This is the second annual report in an experimental and theoretical program to develop and apply single and few element heterogeneous methods for the determination of reactor lattice parameters. During the period covered ...

Driscoll, Michael J.

1969-01-01

396

Ultra high temperature particle bed reactor design  

SciTech Connect

This study is a computer analysis of a conceptual nuclear reactor. The purpose of this work is to design a direct nuclear propulsion engine which could be used for a mission to Mars. The main features of this reactor design are high values for I{sub sp} and, secondly, very efficient cooling. This particle bed reactor consists of 37 cylindrical fuel elements embedded in a cylinder of beryllium which acts as a moderator and reflector. The fuel consists of a packed bed of spherical fissionable fuel particles. Gaseous H{sub 2} passes over the fuel bed, removes the heat and is exhausted out of the rocket. The design was found to be neutronically critical and to have tolerable heating rates. Therefore, this Particle Bed Reactor Design is suitable as a propulsion unit for this mission.

Lazareth, O.; Ludewig, H.; Perkins, K.; Powell, J.

1990-01-01

397

Nonlinear, Inelastic Fast Reactor Subassembly Interaction Analyses.  

National Technical Information Service (NTIS)

Liquid Metal Fast Breeder Reactor (LMFBR) core structural design is complicated by the trade-offs associated with keeping the subassemblies closely packed for the neutronic considerations and accommodating the volumetric changes associated with irradiatio...

W. H. Sutherland, F. E. Bard

1983-01-01

398

EMERGING TECHNOLOGY BULLETIN: SPOUTED BED REACTOR  

EPA Science Inventory

The Spouted Bed Reactor (SBR) technology utilizes the unique attributes of the "spouting " fluidization regime, which can provide heat transfer rates comparable to traditional fluid beds, while providing robust circulation of highly heterogeneous solids, concurrent with very agg...

399

Critical assessment of thorium reactor technology  

E-print Network

Thorium-based fuels for nuclear reactors are being considered for use with current and future designs in both large and small-scale energy production. Thorium-232 is as abundant on Earth as lead, far more common than all ...

Drenkhahn, Robert (Robert A.)

2012-01-01

400

Thermal Shield and Reactor Structure Temperatures  

SciTech Connect

The purpose of this report is to present reactor structure and thermal shield temperature data taken during P-3 and P-5 cycles and compare them with design calculations in order to predict temperatures at higher power levels.

Collier, A.R.

2001-07-31

401

INTERNATIONAL JOURNAL OF CHEMICAL REACTOR ENGINEERING  

E-print Network

relationships for drag, heat and mass transfer in continuum level simulations. There has been significant workINTERNATIONAL JOURNAL OF CHEMICAL REACTOR ENGINEERING Volume 6 2008 Article A28 Spatiotemporal

Deymier, Pierre

402

A fast and flexible reactor physics model for simulating neutron spectra and depletion in fast reactors  

NASA Astrophysics Data System (ADS)

Determining the time dependent concentration of isotopes within a nuclear reactor core is central to the analysis of nuclear fuel cycles. We present a fast, flexible tool for determining the time dependent neutron spectrum within fast reactors. The code (VBUDS: visualization, burnup, depletion and spectra) uses a two region, multigroup collision probability model to simulate the energy dependent neutron flux and tracks the buildup and burnout of 24 actinides, as well as fission products. While originally developed for LWR simulations, the model is shown to produce fast reactor spectra that show high degree of fidelity to available fast reactor benchmarks.

Recktenwald, Geoff; Deinert, Mark

2010-03-01

403

Reactor scram events in the updated PIUS 600 advanced reactor design  

SciTech Connect

The PIUS advanced reactor is a 640-MWe pressurized water reactor concept developed by Asea Brown Boveri. A unique feature of PIUS is the absence of mechanical control and shutdown rods. Reactivity is controlled by coolant boron concentration and the temperature of the moderator coolant. Los Alamos supported the US Nuclear Regulatory Commission`s preapplication review of the PIUS reactor. Baseline calculations of the PIUS design were performed for active and passive reactor scrams using TRAC-PF1/MOD2. Additional sensitivity studies examined flow blockage and boron dilution events to explore the robustness of the PIUS concept for low-probability combination events following active-system scrams.

Boyack, B.E.; Steiner, J.L.; Harmony, S.C.; Stumpf, H.J.; Lime, J.F.

1994-12-31

404

Developments and Tendencies in Fission Reactor Concepts  

NASA Astrophysics Data System (ADS)

This chapter describes, in two parts, new-generation nuclear energy systems that are required to be in harmony with nature and to make full use of nuclear resources. The issues of transmutation and containment of radioactive waste will also be addressed. After a short introduction to the first part, Sect. 58.1.2 will detail the requirements these systems must satisfy on the basic premise of peaceful use of nuclear energy. The expected designs themselves are described in Sect. 58.1.3. The subsequent sections discuss various types of advanced reactor systems. Section 58.1.4 deals with the light water reactor (LWR) whose performance is still expected to improve, which would extend its application in the future. The supercritical-water-cooled reactor (SCWR) will also be shortly discussed. Section 58.1.5 is mainly on the high temperature gas-cooled reactor (HTGR), which offers efficient and multipurpose use of nuclear energy. The gas-cooled fast reactor (GFR) is also included. Section 58.1.6 focuses on the sodium-cooled fast reactor (SFR) as a promising concept for advanced nuclear reactors, which may help both to achieve expansion of energy sources and environmental protection thus contributing to the sustainable development of mankind. The molten-salt reactor (MSR) is shortly described in Sect. 58.1.7. The second part of the chapter deals with reactor systems of a new generation, which are now found at the research and development (R&D) stage and in the medium term of 20-30 years can shape up as reliable, economically efficient, and environmentally friendly energy sources. They are viewed as technologies of cardinal importance, capable of resolving the problems of fuel resources, minimizing the quantities of generated radioactive waste and the environmental impacts, and strengthening the regime of nonproliferation of the materials suitable for nuclear weapons production. Particular attention has been given to naturally safe fast reactors with a closed fuel cycle (CFC) - as an advanced and promising reactor system that offers solutions to the above problems. The difference (not confrontation) between the approaches to nuclear power development based on the principles of “inherent safety” and “natural safety” is demonstrated.

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

405

Solid oxide electrochemical reactor science.  

SciTech Connect

Solid-oxide electrochemical cells are an exciting new technology. Development of solid-oxide cells (SOCs) has advanced considerable in recent years and continues to progress rapidly. This thesis studies several aspects of SOCs and contributes useful information to their continued development. This LDRD involved a collaboration between Sandia and the Colorado School of Mines (CSM) ins solid-oxide electrochemical reactors targeted at solid oxide electrolyzer cells (SOEC), which are the reverse of solid-oxide fuel cells (SOFC). SOECs complement Sandia's efforts in thermochemical production of alternative fuels. An SOEC technology would co-electrolyze carbon dioxide (CO{sub 2}) with steam at temperatures around 800 C to form synthesis gas (H{sub 2} and CO), which forms the building blocks for a petrochemical substitutes that can be used to power vehicles or in distributed energy platforms. The effort described here concentrates on research concerning catalytic chemistry, charge-transfer chemistry, and optimal cell-architecture. technical scope included computational modeling, materials development, and experimental evaluation. The project engaged the Colorado Fuel Cell Center at CSM through the support of a graduate student (Connor Moyer) at CSM and his advisors (Profs. Robert Kee and Neal Sullivan) in collaboration with Sandia.

Sullivan, Neal P. (Colorado School of Mines, Golden, CO); Stechel, Ellen Beth; Moyer, Connor J. (Colorado School of Mines, Golden, CO); Ambrosini, Andrea; Key, Robert J. (Colorado School of Mines, Golden, CO)

2010-09-01

406

The 5-kwe reactor thermoelectric system summary  

NASA Technical Reports Server (NTRS)

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

Vanosdol, J. H. (editor)

1973-01-01

407

OPERATION OF THE AIRCRAFT REACTOR EXPERIMENT  

Microsoft Academic Search

The Aircrift Reactor Experiment (ARE) was oporated successfully and ; without untoward difficulty in November 1954. The reactor became critical with a ; mass of 32.8 lb of U²³⁵, which gave a concentration of 23.9 lb of U²³⁵; per cubic foot of fluoride fuel. For operation at power, the U²³⁵ content ; of the fuel mixture was increased to 26.0

W. B. Cottrell; H. E. Hungerford; J. K. Leslie; J. L. Meem

1955-01-01

408

Natural circulation in fusion reactor blankets  

Microsoft Academic Search

The relative importance of natural circulation and heat conduction as heat transfer mechanisms in lithium, sodium and flibe is investigated for a range of magnetic field strengths of interest in fusion reactor blankets. The calculations are based on an order-of-magnitude simplification of the fluid equations, and a modified version of the fission reactor thermal-hydraulic code THERMIT. The results show that

P. J. Gierszewski; B. Mikic; N. E. Todreas

1980-01-01

409

Monitoring and Control of Anaerobic Reactors  

Microsoft Academic Search

The current status in monitoring and control of anaerobic reactors is reviewed. The influence of reactor design and waste\\u000a composition on the possible monitoring and control schemes is examined. After defining the overall control structure, and\\u000a possible control objectives, the possible process measurements are reviewed in detail. In the sequel, possible manipulated\\u000a variables, such as the hydraulic retention time, the

Peter F. Pind; Irini Angelidaki; Birgitte K. Ahring; Katerina Stamatelatou; Gerasimos Lyberatos

410

Space reactors: What is a kilogram  

SciTech Connect

The use of nuclear electric propulsion can triple payloads to GEO for a single Shuttle launch. Life orbits of 300 years can be used to allow most of the fission and activation products to decay before a reactor reenters the biosphere. Enough radioactive materials remain with very long lifetimes to make it desirable to design the reactor to disperse upon reentry and little additional risk to the biosphere is introduced by initiating NEP operations from 300 km.

Buden, D.; Angelo, J. Jr.; Ek, D.; Voss, S.

1984-01-01

411

Space reactors - past, present, and future  

SciTech Connect

In the 1990s and beyond, advanced-design nuclear reactors could represent the prime source of both space power and propulsion. Many sophisticated military and civilian space missions of the future will require first kilowatt and then megawatt levels of power. This paper reviews key technology developments that accompanied past US space nuclear power development efforts, describes on-going programs, and then explores reactor technologies that will satisfy megawatt power level needs and beyond.

Buden, D.; Angelo, J.

1983-01-01

412

Renewed interest in lead cooled fast reactors  

Microsoft Academic Search

In the last few years a number of compact designs of lead-alloy cooled systems have been promoted. Moreover, in Russia a design effort was started earlier on the pure lead-cooled BREST reactor but this effort does not appear to be strongly funded any more. But now the lead cooled and compact STAR-LM reactor is promoted in the US and in

H. U. Wider; J. Carlsson; E. Loewen

2005-01-01

413

Explosive properties of reactor?grade plutonium  

Microsoft Academic Search

The following discussion focuses on the question of whether a terrorist organization or a threshold state could make use of plutonium recovered from light?water?reactor fuel to construct a nuclear explosive device having a significantly damaging yield. Questions persist in some nonproliferation policy circles as to whether a bomb could be made from reactor?grade plutonium of high burn?up, and if so,

J. Carson Marka

1993-01-01

414

Fluidized-Bed Reactor With Zone Heating  

NASA Technical Reports Server (NTRS)

Deposition of silicon on wall suppressed. In new fluidized bed, silicon seed particles heated in uppermost zone of reactor. Hot particles gradually mix with lower particles and descend through fluidized bed. Lower wall of vessel kept relatively cool. Because silane enters at bottom and circulates through reactor pyrolized to silicon at high temperatures, silicon deposited on particles in preference wall. Design of fluidized bed for production of silicon greatly reduces tendency of silicon to deposit on wall of reaction vessel.

Iya, Sridhar K.

1989-01-01

415

Interpretation of data from a pulse reactor  

E-print Network

for the reactant: (1) the input, pulse is a triangular pulse; (2) plug flow exists within the reactor; (3) isothermal condition prevails throughout the catalyst bed; (4) no axial diffusion occurs: (5) carrier gas velocity is constant; (6) external film resistance... condition, Equation 2, implies that there is no reactant present in the catalyst bed at the begining of each pulse. The boundary condition, Equation 3, means that the input pulse at the reactor entrance is known in detail. The differential equation...

Desai, Nayneshkumar Shantilal

2012-06-07

416

Environmental Information Document: L-reactor reactivation  

SciTech Connect

Purpose of this Environmental Information Document is to provide background for assessing environmental impacts associated with the renovation, restartup, and operation of L Reactor at the Savannah River Plant (SRP). SRP is a major US Department of Energy installation for the production of nuclear materials for national defense. The purpose of the restart of L Reactor is to increase the production of nuclear weapons materials, such as plutonium and tritium, to meet projected needs in the nuclear weapons program.

Mackey, H.E. Jr. (comp.)

1982-04-01

417

Trends in fusion reactor safety research  

SciTech Connect

Fusion has the potential to be an attractive energy source. From the safety and environmental perspective, fusion must avoid concerns about catastrophic accidents and unsolvable waste disposal. In addition, fusion must achieve an acceptable level of risk from operational accidents that result in public exposure and economic loss. Finally, fusion reactors must control routine radioactive effluent, particularly tritium. Major progress in achieving this potential rests on development of low-activation materials or alternative fuels. The safety and performance of various material choices and fuels for commercial fusion reactors can be investigated relatively inexpensively through reactor design studies. These studies bring together experts in a wide range of backgrounds and force the group to either agree on a reactor design or identify areas for further study. Fusion reactors will be complex with distributed radioactive inventories. The next generation of experiments will be critical in demonstrating that acceptable levels of safe operation can be achieved. These machines will use materials which are available today and for which a large database exists (e.g. for 316 stainless steel). Researchers have developed a good understanding of the risks associated with operation of these devices. Specifically, consequences from coolant system failures, loss of vacuum events, tritium releases, and liquid metal reactions have been studied. Recent studies go beyond next step designs and investigate commercial reactor concerns including tritium release and liquid metal reactions. 18 refs.

Herring, J.S.; Holland, D.F.; Piet, S.J.

1991-01-01

418

Design options for a bunsen reactor.  

SciTech Connect

This work is being performed for Matt Channon Consulting as part of the Sandia National Laboratories New Mexico Small Business Assistance Program (NMSBA). Matt Channon Consulting has requested Sandia's assistance in the design of a chemical Bunsen reactor for the reaction of SO2, I2 and H2O to produce H2SO4 and HI with a SO2 feed rate to the reactor of 50 kg/hour. Based on this value, an assumed reactor efficiency of 33%, and kinetic data from the literature, a plug flow reactor approximately 1%E2%80%9D diameter and and 12 inches long would be needed to meet the specification of the project. Because the Bunsen reaction is exothermic, heat in the amount of approximately 128,000 kJ/hr would need to be removed using a cooling jacket placed around the tubular reactor. The available literature information on Bunsen reactor design and operation, certain support equipment needed for process operation and a design that meet the specification of Matt Channon Consulting are presented.

Moore, Robert Charles

2013-10-01

419

U.S. domestic reactor conversion program  

SciTech Connect

The RERTR U.S. Domestic Conversion program continues in its support of the Global Treat Reduction Initiative (GTRI) to convert seven U.S reactors to low enriched uranium (LEU) by 2010. These reactors are located at the University of Florida, Texas A and M University, Purdue University, Washington State University, Oregon State University, the University of Wisconsin, and the Idaho National Laboratory. The reactors located at the University of Florida and Texas A and M Nuclear Science Center were successfully converted to LEU in September of 2006 through an integrated and collaborative effort involving INL, Argonne National Laboratory (ANL), DOE (headquarters and the field office), the Nuclear Regulatory Commission (NRC), the universities, and the contractors involved in analyses, fuel design and fabrication, and spent nuclear fuel (SNF) shipping and disposition. With this work completed and in anticipation of other impending conversion projects, a meeting was established to engage the project participants in a structured discussion to capture the lessons learned. The objectives of this meeting were to document the observations, insights, issues, concerns, and ideas of those involved in the reactor conversions so that future efforts could be conducted with greater effectiveness, efficiency, and with fewer challenges. The lessons learned from completing the University of Florida and Texas A and M conversions, the Purdue reactor conversion status, and an overview of the upcoming reactor conversions will be presented at the meeting. (author)

Meyer, Dana M.; Woolstenhulme, Eric C. [Idaho National Laboratory, Idaho Falls, Idaho 83415 (United States)

2008-07-15

420

Uncertainty quantification approaches for advanced reactor analyses.  

SciTech Connect

The original approach to nuclear reactor design or safety analyses was to make very conservative modeling assumptions so as to ensure meeting the required safety margins. Traditional regulation, as established by the U. S. Nuclear Regulatory Commission required conservatisms which have subsequently been shown to be excessive. The commission has therefore moved away from excessively conservative evaluations and has determined best-estimate calculations to be an acceptable alternative to conservative models, provided the best-estimate results are accompanied by an uncertainty evaluation which can demonstrate that, when a set of analysis cases which statistically account for uncertainties of all types are generated, there is a 95% probability that at least 95% of the cases meet the safety margins. To date, nearly all published work addressing uncertainty evaluations of nuclear power plant calculations has focused on light water reactors and on large-break loss-of-coolant accident (LBLOCA) analyses. However, there is nothing in the uncertainty evaluation methodologies that is limited to a specific type of reactor or to specific types of plant scenarios. These same methodologies can be equally well applied to analyses for high-temperature gas-cooled reactors and to liquid metal reactors, and they can be applied to steady-state calculations, operational transients, or severe accident scenarios. This report reviews and compares both statistical and deterministic uncertainty evaluation approaches. Recommendations are given for selection of an uncertainty methodology and for considerations to be factored into the process of evaluating uncertainties for advanced reactor best-estimate analyses.

Briggs, L. L.; Nuclear Engineering Division

2009-03-24

421

National strategies for nuclear power reactor development  

SciTech Connect

The document assesses the potential for design innovation in three areas of nuclear power plant technology: light-water reactor systems; liquid-metal reactor systems; and high-temperature-gas reactor systems. The question of how capital costs scale with unit size in nuclear power reactor systems is addressed. Small modular high temperature gas reactor designs are reviewed, and an electric-power-system capacity-planning model that allows estimates to be made of the economic implications of construction lead time and power rating variations for utility ratepayers and shareholders is presented. The effects of industrial reorganization and structure on the economic performance of the nuclear power industry in the U.S. is explored, and options available to the Nuclear Regulatory Commission for establishing the general philosphy, technical criteria, and organizational approach to the regulation of a new generation of nuclear power plants are discussed. Light-water-reactor systems are determined to be the only plausible technical option if nuclear power is to figure in the capacity expansion plans of U.S. electric utilities in the near term.

Lester, R.K.; Driscoll, M.J.; Golay, M.W.; Lanning, D.D.; Lidsky, L.M.

1985-03-01

422

A nanoliter-scale open chemical reactor.  

PubMed

An open chemical reactor is a container that exchanges matter with the exterior. Well-mixed open chemical reactors, called continuous stirred tank reactors (CSTR), have been instrumental for investigating the dynamics of out-of-equilibrium chemical processes, such as oscillations, bistability, and chaos. Here, we introduce a microfluidic CSTR, called ?CSTR, that reduces reagent consumption by six orders of magnitude. It consists of an annular reactor with four inlets and one outlet fabricated in PDMS using multi-layer soft lithography. A monolithic peristaltic pump feeds fresh reagents into the reactor through the inlets. After each injection the content of the reactor is continuously mixed with a second peristaltic pump. The efficiency of the ?CSTR is experimentally characterized using a bromate, sulfite, ferrocyanide pH oscillator. Simulations accounting for the digital injection process are in agreement with experimental results. The low consumption of the ?CSTR will be advantageous for investigating out-of-equilibrium dynamics of chemical processes involving biomolecules. These studies have been scarce so far because a miniaturized version of a CSTR was not available. PMID:23223849

Galas, Jean-Christophe; Haghiri-Gosnet, Anne-Marie; Estévez-Torres, André

2013-02-01

423

Energy conversion systems design for fusion reactors  

SciTech Connect

Various energy conversion systems have been reviewed in order to select an efficient power cycle to be compatible with the fusion reactor requirements. The power cycles were selected for a toroidal confinement system with D-T and D-D fuel cycles and a tandem mirror reactor (TMR) with a D-/sup 3/He fuel cycle. Reversed Field Pinch Reactor (RFPR) was selected as an example of a toroidal confinement system with D-T fuel cycle since there has recently been a comprehensive design study for it. Tokamak was selected as an example of a toroidal confinement system with D-D fuel cycle. Tandem mirror reactor was chosen as an example of confinement for D-/sup 3/He fuel cycle. The steam cycle was found to be most suitable for the RFP and Tokamak reactors while a combination of direct energy conversion system and steam cycle was found to be most suitable for D-/sup 3/He tandem mirror reactor.

Dabiri, A.E.

1989-03-01

424

Micro -Thermonuclear AB-Reactors for Aerospace  

E-print Network

The author offers several innovations that he first suggested publicly early in 1983 for the AB multi-reflex engine, space propulsion, getting energy from plasma, etc. (see: A. Bolonkin, Non-Rocket Space Launch and Flight, Elsevier, London, 2006, Chapters 12, 3A). It is the micro-thermonuclear AB-Reactors. That is new micro-thermonuclear reactor with very small fuel pellet that uses plasma confinement generated by multi-reflection of laser beam or its own magnetic field. The Lawson criterion increases by hundreds of times. The author also suggests a new method of heating the power-making fuel pellet by outer electric current as well as new direct method of transformation of ion kinetic energy into harvestable electricity. These offered innovations dramatically decrease the size, weight and cost of thermonuclear reactor, installation, propulsion system and electric generator. Non-industrial countries can produce these researches and constructions. Currently, the author is researching the efficiency of these innovations for two types of the micro-thermonuclear reactors: multi-reflection reactor (ICF) and self-magnetic reactor (MCF).

Alexander Bolonkin

2007-01-08

425

Antineutrino reactor safeguards - a case study  

E-print Network

Antineutrinos have been proposed as a means of reactor safeguards for more than 30 years and there has been impressive experimental progress in neutrino detection. In this paper we conduct, for the first time, a case study of the application of antineutrino safeguards to a real-world scenario - the North Korean nuclear crisis in 1994. We derive detection limits to a partial or full core discharge in 1989 based on actual IAEA safeguards access and find that two independent methods would have yielded positive evidence for a second core with very high confidence. To generalize our results, we provide detailed estimates for the sensitivity to the plutonium content of various types of reactors, including most types of plutonium production reactors, based on detailed reactor simulations. A key finding of this study is that a wide class of reactors with a thermal power of less than 0.1-1 GWth can be safeguarded achieving IAEA goals for quantitative sensitivity and timeliness with detectors right outside the reactor building. This type of safeguards does not rely on the continuity of knowledge and provides information about core inventory and power status in real-time.

Eric Christensen; Patrick Huber; Patrick Jaffke

2013-12-06

426

The integral fast reactor fuel cycle  

SciTech Connect

The liquid-metal reactor (LMR) has the potential to extend the uranium resource by a factor of 50 to 100 over current commercial light water reactors (LWRs). In the integral fast reactor (IFR) development program, the entire reactor system - reactor, fuel cycle, and waste process - is being developed and optimized at the same time as a single integral entity. A key feature of the IFR concept is the metallic fuel. The lead irradiation tests on the new U-Pu-Zr metallic fuel in the Experimental Breeder Reactor II have surpassed 185000 MWd/t burnup, and its high burnup capability has now been fully demonstrated. The metallic fuel also allows a radically improved fuel cycle technology. Pyroprocessing, which utilizes high temperatures and molten salt and molten metal solvents, can be advantageously utilized for processing metal fuels because the product is metal suitable for fabrication into new fuel elements. Direct production of a metal product avoids expensive and cumbersome chemical conversion steps that would result from use of the conventional Purex solvent extraction process. The key step in the IFR process is electrorefining, which provides for recovery of the valuable fuel constituents, uranium and plutonium, and for removal of fission products. A notable feature of the IFR process is that the actinide elements accompany plutonium through the process. This results in a major advantage in the high-level waste management.

Chang, Y.I. (Argonne National Lab., IL (United States))

1990-01-01

427

Control of volatile organic compounds by an AC energized ferroelectric pellet reactor and a pulsed corona reactor  

Microsoft Academic Search

Two laboratory-scale plasma reactors, an alternating current (AC) energized ferroelectric (high dielectric ceramic) packed bed reactor and a nanosecond pulsed corona reactor, were constructed. This study was done to develop baseline engineering data to demonstrate the feasibility of the application of plasma reactors to the destruction of various volatile organic compounds (VOCs) at PPM levels. Complete destruction was obtained for

Toshiaki Yamamoto; Kumar Ramanathan; Phil A. Lawless; David S. Ensor; J. Randall Newsome; Norman Plaks; Geddes H. Ramsey

1992-01-01

428

Trends and developments in magnetic confinement fusion reactor concepts  

SciTech Connect

An overview is presented of recent design trends and developments in reactor concepts for magnetic confinement fusion. The paper emphasizes the engineering and technology considerations of commercial fusion reactor concepts and reactors that operate on the deuterium/tritium/ lithium fuel cycle. Recent developments in tokamak, mirror, and Elmo Bumpy Torus reactor concepts are described, as well as a survey of recent developments on a wide variety of alternate magnetic fusion reactor concepts (within the last two to three years).

Baker, C.C.; Carlson, G.A.; Krakowski, R.A.

1981-01-01

429

CHARACTERIZATION OF RADIOACTIVITY IN THE REACTOR VESSEL OF THE HEAVY WATER COMPONENT TEST REACTOR  

SciTech Connect

The Heavy Water Component Test Reactor (HWCTR) facility is a pressurized heavy water reactor that was used to test candidate fuel designs for heavy water power reactors. The reactor operated at nominal power of 50 MW{sub th}. The reactor coolant loop operated at 1200 psig and 250 C. Two isolated test loop were designed into the reactor to provide special test conditions. Fig. 1 shows a cut-away view of the reactor. The two loops are contained in four inch diameter stainless steel piping. The HWCTR was operated for only a short duration, from March 1962 to December 1964 in order to test the viability of test fuel elements and other reactor components for use in a heavy water power reactor. The reactor achieved 13,882 MWd of total power while testing 36 different fuel assemblies. In the course of operation, HWCTR experienced the cladding failures of 10 separate test fuel assemblies. In each case, the cladding was breached with some release of fuel core material into the isolated test loop, causing fission product and actinide contamination in the main coolant loop and the liquid and boiling test loops. Despite the contribution of the contamination from the failed fuel, the primary source of radioactivity in the HWCTR vessel and internals is the activation products in the thermal shields, and to a lesser degree, activation products in the reactor vessel walls and liner. A detailed facility characterization report of the HWCTR facility was completed in 1996. Many of the inputs and assumptions in the 1996 characterization report were derived from the HWCTR decommissioning plan published in 1975. The current paper provides an updated assessment of the radioisotopic characteristics of the HWCTR vessel and internals to support decommissioning activities on the facility.

Vinson, Dennis

2010-06-01

430

Proliferation Resistant Nuclear Reactor Fuel  

SciTech Connect

Global appetite for fission power is projected to grow dramatically this century, and for good reason. Despite considerable research to identify new sources of energy, fission remains the most plentiful and practical alternative to fossil fuels. The environmental challenges of fossil fuel have made the fission power option increasingly attractive, particularly as we are forced to rely on reserves in ecologically fragile or politically unstable corners of the globe. Caught between a globally eroding fossil fuel reserve as well as the uncertainty and considerable costs in the development of fusion power, most of the world will most likely come to rely on fission power for at least the remainder of the 21st century. Despite inevitable growth, fission power faces enduring challenges in sustainability and security. One of fission power's greatest hurdles to universal acceptance is the risk of potential misuse for nefarious purposes of fissionable byproducts in spent fuel, such as plutonium. With this issue in mind, we have discussed intrinsic concepts in this report that are motivated by the premise that the utility, desirability, and applicability of nuclear materials can be reduced. In a general sense, the intrinsic solutions aim to reduce or eliminate the quantity of existing weapons usable material; avoid production of new weapons-usable material through enrichment, breeding, extraction; or employ engineering solutions to make the fuel cycle less useful or more difficult for producing weapons-usable material. By their nature, these schemes require modifications to existing fuel cycles. As such, the concomitants of these modifications require engagement from the nuclear reactor and fuel-design community to fully assess their effects. Unfortunately, active pursuit of any scheme that could further complicate the spread of domestic nuclear power will probably be understandably unpopular. Nevertheless, the nonproliferation and counterterrorism issues are paramount, and we posit that the exploration, development, and implementation of intrinsic mechanisms such as discussed here are part of a balanced approach aimed at preventing the misuse of nuclear material for nuclear-energy applications.

Gray, L W; Moody, K J; Bradley, K S; Lorenzana, H E

2011-02-18

431

Summary of advanced LMR (Liquid Metal Reactor) evaluations: PRISM (Power Reactor Inherently Safe Module) and SAFR (Sodium Advanced Fast Reactor)  

SciTech Connect

In support of the US Nuclear Regulatory Commission (NRC), Brookhaven National Laboratory (BNL) has performed independent analyses of two advanced Liquid Metal Reactor (LMR) concepts. The designs, sponsored by the US Department of Energy (DOE), the Power Reactor Inherently Safe Module (PRISM) (Berglund, 1987) and the Sodium Advanced Fast Reactor (SAFR) (Baumeister, 1987), were developed primarily by General Electric (GE) and Rockwell International (RI), respectively. Technical support was provided to DOE, RI, and GE, by the Argonne National Laboratory (ANL), particularly with respect to the characteristics of the metal fuels. There are several examples in both PRISM and SAFR where inherent or passive systems provide for a safe response to off-normal conditions. This is in contrast to the engineered safety systems utilized on current US Light Water Reactor (LWR) designs. One important design inherency in the LMRs is the inherent shutdown'', which refers to the tendency of the reactor to transition to a much lower power level whenever temperatures rise significantly. This type of behavior was demonstrated in a series of unscrammed tests at EBR-II (NED, 1986). The second key design feature is the passive air cooling of the vessel to remove decay heat. These systems, designated RVACS in PRISM and RACS in SAFR, always operate and are believed to be able to prevent core damage in the event that no other means of heat removal is available. 27 refs., 78 figs., 3 tabs.

Van Tuyle, G.J.; Slovik, G.C.; Chan, B.C.; Kennett, R.J.; Cheng, H.S.; Kroeger, P.G. (Brookhaven National Lab., Upton, NY (USA))

1989-10-01

432

Three-phase packed bed reactor with an evaporating solvent—II. Modelling of the reactor  

Microsoft Academic Search

In this paper two models are presented for a three-phase catalytic packed bed reactor in which in evaporating solvent is used to absorb and remove most of the reaction heat. A plug flow model and a model comprising mass and heat dispersion in the reactor are discussed. The results of both models are compared to each other and to experimental

Gelder van K. B; P. C. Borman; R. E. Weenink; K. R. Westerterp

1990-01-01

433

Utilization of TRISO fuel with reactor grade plutonium in CANDU reactors  

Microsoft Academic Search

Large quantities of plutonium have been accumulated in the nuclear waste of civilian LWRs and CANDU reactors. Reactor grade plutonium and heavy water moderator can give a good combination with respect to neutron economy. On the other hand, TRISO type fuel can withstand very high fuel burn-up levels. The paper investigates the prospects of utilization of TRISO fuel made of

Sümer ?ahin; Hac? Mehmet ?ahin; Adem Ac?r

2010-01-01

434

Section 3. Reactor physics studies Reactor physics aspects of plutonium burning in inert matrix fuels  

Microsoft Academic Search

Burnup calculations have been performed on fuels containing either reactor grade or weapons grade plutonium mixed in an inert matrix or mixed in a thorium oxide matrix. At each branching during burnup, the fuel temperature coeÅcient, the moderator void coeÅcient and the boron reactivity worth have been calculated. From the reactor physics point of view, use of thorium oxide as

J. L. Kloosterman; P. M. G. Damen

435

Selection of a toroidal fusion reactor concept for a magnetic fusion production reactor  

Microsoft Academic Search

The basic fusion driver requirements of a toroidal materials production reactor are considered. The tokamak, stellarator, bumpy torus, and reversed-field pinch are compared with regard to their demonstrated performance, probable near-term development, and potential advantages and disadvantages if used as reactors for materials production. Of the candidate fusion drivers, the tokamak is determined to be the most viable for a

D. L. Jassby

1987-01-01

436

The Japan power demonstration reactor decommissioning program - experience of nuclear power reactor dismantling  

SciTech Connect

The Japan power demonstration reactor (JPDR) decommissioning program is in progress of developing new technology for reactor decommissioning and collecting various data on project management and machine performance and operability of new dismantling techniques. The experience and the data obtained from the JPDR decommissioning program are expected to contribute to future decommissioning of commercial nuclear power plants.

Yokota, M.; Yanagihara, S.; Miki, I.; Hoshi, T. (Japan Atomic Energy Research Institute, Ibaraki-ken (Japan))

1992-01-01

437

REACTOR PRESSURE VESSEL ISSUES FOR THE LIGHT-WATER REACTOR SUSTAINABILITY PROGRAM  

Microsoft Academic Search

The Light Water Reactor Sustainability Program Plan is a collaborative program between the U.S. Department of Energy and the private sector directed at extending the life of the present generation of nuclear power plants to enable operation to at least 80 years. The reactor pressure vessel (RPV) is one of the primary components requiring significant research to enable such long-term

Randy K Nanstad; George Robert Odette

2010-01-01

438

The Fast-Neutron Breeder Fission Reactor: Safety Issues in Reactor Design and Operation: Discussion  

Microsoft Academic Search

Today's fast breeder reactors contain mixed uranium-plutonium oxide fuel and are cooled with liquid sodium. Their normal operational behaviour, including power transients, is similar to that of thermal reactors. The fact that the sodium density coefficient is positive is of no importance at normal operating temperatures because negative coefficients like Doppler or fuel expansion coefficients have compensating effects. Dangerous effects

H.-H. Hennies; J. D. Griffith; F. J. Barclay; D. Broadley; P. Dastidar

1990-01-01

439

Reactor Safety Evaluation Division: K-Reactor Operational Readiness Evaluation program plan  

Microsoft Academic Search

The purpose of this plan is to define the process to be used by the Reactor Safety Evaluation Division (RSED) to provide an independent evaluation of the K-Reactor restart effort. RSED will evaluate the implementation of the contractor related issues in the DOE restart strategy as well as the conduct of the contractor's Operational Readiness Review. Additional information regarding the

V. S. OBlock; D. A. Busch

1989-01-01

440

Optimized Transition from the Reactors of Second and Third Generations to the Thorium Molten Salt Reactor  

Microsoft Academic Search

Molten salt reactors, in the configuration presented here and called Thorium Molten Salt Reactor (TMSR), are particularly well suited to fulfil the criteria chosen by the Generation IV forum, and may be operated in simplified and safe conditions in the Th\\/233U fuel cycle with fluoride salts. Amongst all MSR configurations in the thorium cycle, many studies have highlighted the configurations

E. Merle-Lucotte; D. Heuer; M. Allibert; V. Ghetta; C. Le Brun; L. Mathieu; R. Brissot; E. Liatard

441

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

SciTech Connect

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. Many 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.

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

2011-07-18

442

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

NASA Astrophysics Data System (ADS)

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.

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

2006-01-01

443

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

SciTech Connect

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.

Dixon, David D. [Nuclear Systems Design Group, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Hiatt, Matthew T. [Nuclear Systems Design Group, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Department of Nuclear Engineering, Texas A and M University, College Station, TX 77843 (United States); Poston, David I.; Kapernick, Richard J. [Nuclear Systems Design Group, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

2006-01-20

444

HEAVY WATER COMPONENTS TEST REACTOR DECOMMISSIONING  

SciTech Connect

The Heavy Water Components Test Reactor (HWCTR) Decommissioning Project was initiated in 2009 as a Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) Removal Action with funding from the American Recovery and Reinvestment Act (ARRA). This paper summarizes the history prior to 2009, the major D&D activities, and final end state of the facility at completion of decommissioning in June 2011. The HWCTR facility was built in 1961, operated from 1962 to 1964, and is located in the northwest quadrant of the Savannah River Site (SRS) approximately three miles from the site boundary. The HWCTR was a pressurized heavy water test reactor used to develop candidate fuel designs for heavy water power reactors. In December of 1964, operations were terminated and the facility was placed in a standby condition as a result of the decision by the U.S. Atomic Energy Commission to redirect research and development work on heavy water power reactors to reactors cooled with organic materials. For about one year, site personnel maintained the facility in a standby status, and then retired the reactor in place. In the early 1990s, DOE began planning to decommission HWCTR. Yet, in the face of new budget constraints, DOE deferred dismantlement and placed HWCTR in an extended surveillance and maintenance mode. The doors of the reactor facility were welded shut to protect workers and discourage intruders. In 2009 the $1.6 billion allocation from the ARRA to SRS for site footprint reduction at SRS reopened the doors to HWCTR - this time for final decommissioning. Alternative studies concluded that the most environmentally safe, cost effective option for final decommissioning was to remove the reactor vessel, both steam generators, and all equipment above grade including the dome. The transfer coffin, originally above grade, was to be placed in the cavity vacated by the reactor vessel and the remaining below grade spaces would be grouted. Once all above equipment including the dome was removed, a concrete cover was to be placed over the remaining footprint and the groundwater monitored for an indefinite period to ensure compliance with environmental regulations.

Austin, W.; Brinkley, D.

2011-10-13

445

Thermionic reactors for space nuclear power  

NASA Astrophysics Data System (ADS)

Compact thermionic nuclear reactor systems with satisfactory mass performance are competitive with space nuclear power systems based on the organic Rankine and closed Brayton cycles. The mass characteristics of the thermionic space nuclear power system are better than that of the solar power system for power levels beyond about 10 kWe. Longlife thermionic fuel element requirements, including their optimal dimensions, and common requirements for the in-core thermionic reactor design are formulated. Thermal and fast in-core thermionic reactors are considered and the ranges of their sensible use are discussed. Some design features of the fast in-core thermionic reactors cores (power range to 1 MWe) including a choice of coolants are discussed. Mass and dimensional performance for thermionic nuclear power reactor system are assessed. It is concluded that thermionic space nuclear power systems are promising power supplies for spacecrafts and that a single basic type of thermionic fuel element may be used for power requirements ranging to several hundred kWe.

Griaznov, Georgii M.; Zhabotinskii, Evgenii E.; Serbin, Victor I.; Zrodnikov, Anatolii V.; Pupko, Victor Ia.; Ponomarev-Stepnoi, Nikolai N.; Usov, V. A.; Nikolaev, Iu. V.

446

Enzyme reactor design under thermal inactivation.  

PubMed

Temperature is a very relevant variable for any bioprocess. Temperature optimization of bioreactor operation is a key aspect for process economics. This is especially true for enzyme-catalyzed processes, because enzymes are complex, unstable catalysts whose technological potential relies on their operational stability. Enzyme reactor design is presented with a special emphasis on the effect of thermal inactivation. Enzyme thermal inactivation is a very complex process from a mechanistic point of view. However, for the purpose of enzyme reactor design, it has been oversimplified frequently, considering one-stage first-order kinetics of inactivation and data gathered under nonreactive conditions that poorly represent the actual conditions within the reactor. More complex mechanisms are frequent, especially in the case of immobilized enzymes, and most important is the effect of catalytic modulators (substrates and products) on enzyme stability under operation conditions. This review focuses primarily on reactor design and operation under modulated thermal inactivation. It also presents a scheme for bioreactor temperature optimization, based on validated temperature-explicit functions for all the kinetic and inactivation parameters involved. More conventional enzyme reactor design is presented merely as a background for the purpose of highlighting the need for a deeper insight into enzyme inactivation for proper bioreactor design. PMID:12693444

Illanes, Andrés; Wilson, Lorena

2003-01-01

447

Lateral restraint assembly for reactor core  

DOEpatents

A restraint assembly for use in restraining lateral movement of a reactor core relative to a reactor vessel wherein a plurality of restraint assemblies are interposed between the reactor core and the reactor vessel in circumferentially spaced relation about the core. Each lateral restraint assembly includes a face plate urged against the outer periphery of the core by a plurality of compression springs which enable radial preloading of outer reflector blocks about the core and resist low-level lateral motion of the core. A fixed radial key member cooperates with each face plate in a manner enabling vertical movement of the face plate relative to the key member but restraining movement of the face plate transverse to the key member in a plane transverse to the center axis of the core. In this manner, the key members which have their axes transverse to or subtending acute angles with the direction of a high energy force tending to move the core laterally relative to the reactor vessel restrain such lateral movement.

Gorholt, Wilhelm (San Diego, CA); Luci, Raymond K. (Del Mar, CA)

1986-01-01

448

NACA Zero Power Reactor Facility Hazards Summary  

NASA Technical Reports Server (NTRS)

The Lewis Flight Propulsion Laboratory of the National Advisory Committee for Aeronautics proposes to build a zero power research reactor facility which will be located in the laboratory grounds near Clevelaurd, Ohio. The purpose of this report is to inform the Advisory Commit tee on Reactor Safeguards of the U. S. Atomic Energy Commission in re gard to the design of the reactor facility, the cha,acteristics of th e site, and the hazards of operation at this location, The purpose o f this reactor is to perform critical experiments, to measure reactiv ity effects, to serve as a neutron source, and to serve as a training tool. The reactor facility is described. This is followed by a discu ssion of the nuclear characteristics and the control system. Site cha racteristics are then discussed followed by a discussion of the exper iments which may be conducted in the facility. The potential hazards of the facility are then considered, particularly, the maximum credib le accident. Finally, the administrative procedure is discussed.

1957-01-01

449

Fatigue life evaluation for reactor pressure vessels  

SciTech Connect

In terms of plant safety, the reactor pressure vessel is one of the most critical pressure boundary components in the nuclear power plant. The primary degradation mechanisms for reactor pressure vessels are irradiation embrittlement and thermal fatigue. The effects of irradiation and fatigue damage should then be considered in determining the overall lifetime of the reactor pressure vessel. For the radiation damage, related issues have been studied. For the case of fatigue resulting from the pressure and temperature changes, however, relatively less attention has been paid. The fatigue damage is generally regarded as one of the limiting factors for the safe operation life of the reactor pressure vessel. In this paper, the simplified fatigue damage evaluation procedures are applied to the commercially operating reactor vessel which is currently subjected to plant lifetime management study. Specifically, fatigue lifetime evaluation procedures for RPV inlet/outlet nozzles, shell and studs are therefore investigated based on the design basis approaches and the evaluation results are presented based on the assumed operating transient occurrences.

Roh, H.Y.; Jin, T.E. [Korea Power Engineering Co., Inc., Seoul (Korea, Republic of). Power Engineering Research Inst.; Jeong, I.S.; Hong, S.Y. [Korea Electric Power Research Inst., Daejon (Korea, Republic of). Nuclear Div.

1996-12-31

450

Alternatives to proposed replacement production reactors  

SciTech Connect

To insure adequate supplies of plutonium and tritium for defense purposes, an independent evaluation was made by Los Alamos National Laboratory of the numerous alternatives to the proposed replacement production reactors (RPR). This effort concentrated on the defense fuel cycle operation and its technical implications in identifying the principal alternatives for the 1990s. The primary options were identified as (1) existing commercial reactors, (2) existing and planned government-owned facilities (not now used for defense materials production), and (3) other RPRs (not yet proposed) such as CANDU or CANDU-type heavy-water reactors (HWR) for both plutonium and tritium production. The evaluation considered features and differences of various options that could influence choice of RPR alternatives. Barring a change in the US approach to civilian and defense fuel cycles and precluding existing commercial reactors at government-owned sites, the most significant alternatives were identified as a CANDU-type HWR at Savannah River Plant (SRP) site or the Three Mile Island commercial reactor with reprocessing capability at Barnwell Nuclear Fuel Plant and at SRP.

Cullingford, H.S.

1981-06-01

451

Sodium Reactor Experiment decommissioning. Final report  

SciTech Connect

The Sodium Reactor Experiment (SRE) located at the Rockwell International Field Laboratories northwest of Los Angeles was developed to demonstrate a sodium-cooled, graphite-moderated reactor for civilian use. The reactor reached full power in May 1958 and provided 37 GWh to the Southern California Edison Company grid before it was shut down in 1967. Decommissioning of the SRE began in 1974 with the objective of removing all significant radioactivity from the site and releasing the facility for unrestricted use. Planning documentation was prepared to describe in detail the equipment and techniques development and the decommissioning work scope. A plasma-arc manipulator was developed for remotely dissecting the highly radioactive reactor vessels. Other important developments included techniques for using explosives to cut reactor vessel internal piping, clamps, and brackets; decontaminating porous concrete surfaces; and disposing of massive equipment and structures. The documentation defined the decommissioning in an SRE dismantling plan, in activity requirements for elements of the decommissioning work scope, and in detailed procedures for each major task.

Carroll, J.W.; Conners, C.C.; Harris, J.M.; Marzec, J.M.; Ureda, B.F.

1983-08-15

452

Simulation of a marine nuclear reactor  

SciTech Connect

A Nuclear-powered ship Engineering Simulation SYstem (NESSY) has been developed by the Japan Atomic Energy Research Institute as an advanced design tool for research and development of future marine reactors. A marine reactor must respond to changing loads and to the ship`s motions because of the ship`s maneuvering and its presence in a marine environment. The NESSY has combined programs for the reactor plant behavior calculations and the ship`s motion calculations. Thus, it can simulate reactor power fluctuations caused by changing loads and the ship`s motions. It can also simulate the behavior of water in the pressurizer and steam generators. This water sloshes in response to the ship`s motions. The performance of NESSY has been verified by comparing the simulation calculations with the measured data obtained by experiments performed using the nuclear ship Mutsu. The effects of changing loads and the ship`s motions on the reactor behavior can be accurately simulated by NESSY.

Kusunoki, Tsuyoshi; Kyouya, Masahiko; Kobayashi, Hideo; Ochiai, Masaaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Office of Nuclear Ship Research and Development

1995-02-01

453

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

Federal Register 2010, 2011, 2012, 2013

...Availability Decommissioning of Nuclear Power Reactors AGENCY: Nuclear Regulatory...DG-1271 ``Decommissioning of Nuclear Power Reactors.'' This guide describes...184, ``Decommissioning of Nuclear Power Reactors,'' dated July...

2012-02-15

454

Comparison of the technology of oxidative dehydrogenation in a fluidized-bed reactor with those of other reactors for butadiene  

SciTech Connect

This paper describes a comparison among the reactor technologies used in the process of oxidation. For dehydrogenation of butene into butadiene, three reactor types are compared: (1) a fluidized-bed reactor using multi revolving link stoppers with a group VIII variable-valence catalyst, (2) an adiabatic fixed-bed reactor, and (3) a polytube, constant-temperature fixed-bed reactor. The results of the comparison indicate that the polytube fixed-bed reactor at constant temperature is better than the fluidized-bed reactor, which in turn is better than the adiabatic reactor. Using a polytube, constant-temperature fixed-bed reactor with butene`s space velocity of 400 h{sup {minus}1}, butadiene yield reached 78.7%, butene conversion reached 86.1%, and butadiene selectivity reached 91.4%. If these results can be achieved in industry, they will be the world records.

Wu Xingan; Liu Huiqin [Hunan Univ., Changsha (China). Dept. of Chemistry and Chemical Engineering

1996-08-01

455

Weld monitor and failure detector for nuclear reactor system  

DOEpatents

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

Sutton, Jr., Harry G. (Mt. Lebanon, PA)

1987-01-01

456

Microchannel heat exchangers for chemical reactors  

SciTech Connect

A novel heat exchanger which uses microchannels to enhance heat transfer rates is investigated for chemical reactor applications. Heat exchange fluid flows in parallel through multiple channels of micron dimensions to both increase the surface area available for heat transfer and shrink the coolant boundary layer which reduces heat transfer resistance. An experimental apparatus simulates the operation of an exothermic reactor, and heat removal fluxes are measured between 6.5 and 42.5 W/cm{sup 2}. Overall heat transfer coefficients range from 5,000 to 16,000 W/m{sup 2}/K, which represents nearly an order of magnitude increase over a conventional heat exchanger of similar materials. The impact of higher heat removal fluxes is discussed for safe and stable reactor design and operation.

Tonkovich, A.L.Y.; Call, C.J.; Jimenez, D.M.; Wegeng, R.S.; Drost, M.K. [Pacific Northwest National Lab., Richland, WA (United States)

1996-12-31

457

Low power reactor for remote applications  

NASA Astrophysics Data System (ADS)

A compact, low power reactor is being designed to provide electric power for remote, unattended applications. Because of the high fuel and maintenance costs for conventional power sources such as diesel generators, a reactor power supply appears especially attractive for remote and inaccessible locations. Operating at a thermal power level of 135 kWt, the power supply achieves a gross electrical output of 25 kWe from an organic Rankine cycle (ORC) engine. By intentional selection of design features stressing inherent safety, operation in an unattended mode is possible with minimal risk to the environment. Reliability is achieved through the use of components representing existing, proven technology. Low enrichment uranium particle fuel, in graphite core blocks, cooled by heat pipes coupled to an ORC converter insures long term, virtually maintenance free, operation of this reactor for remote applications.

Meier, K. L.; Palmer, R. G.; Kirchner, W. L.

1985-05-01

458

Socioeconomic consequences of nuclear reactor accidents  

SciTech Connect

This report identifies and characterizes the off-site socioeconomic consequences that would likely result from a severe radiological accident at a nuclear power plant. The types of impacts that are addressed include economic impacts, health impacts, social/psychological impacts and institutional impacts. These impacts are identified for each of several phases of a reactor accident - from the warning phase through the post-resettlement phase. The relative importance of the impact during each accident phase and the degree to which the impact can be predicted are indicated. The report also examines the methods that are currently used for assessing nuclear reactor accidents, including development of accident scenarios and the estimating of socioeconomic accident consequences with various models. Finally, a critical evaluation is made regarding the use of impact analyses in estimating the contribution of socioeconomic consequences to nuclear accident reactor accident risk. 116 references, 7 figures, 15 tables.

Tawil, J.J.; Callaway, J.W.; Coles, B.L.; Cronin, F.J.; Currie, J.W.; Imhoff, K.L.; Lewis, P.M.; Nesse, R.J.; Strenge, D.L.

1984-06-01

459

HYLIFE-II reactor chamber design refinements  

SciTech Connect

Mechanical design features of the reactor chamber for the HYLIFE-II inertial confinement fusion power plant are presented. A combination of oscillating and steady, molten salt streams (Li{sub 2}BeF{sub 4}) are used for shielding and blast protection of the chamber walls. The system is designed for a 6 Hz repetition rate. Beam path clearing, between shots, is accomplished with the oscillating flow. The mechanism for generating the oscillating streams is described. A design configuration of the vessel wall allows adequate cooling and provides extra shielding to reduce thermal stresses to tolerable levels. The bottom portion of the reactor chamber is designed to minimize splash back of the high velocity (>12 m/s) salt streams and also recover up to half of the dynamic head. Cost estimates for a 1 GWe and 2 GWe reactor chamber are presented.

House, P.A.

1994-06-01

460

Living anionic polymerization using a microfluidic reactor  

SciTech Connect

Living anionic polymerizations were conducted within aluminum-polyimide microfluidic devices. Polymerizations of styrene in cyclohexane were carried out at various conditions, including elevated temperature (60 °C) and high monomer concentration (42%, by volume). The reactions were safely maintained at a controlled temperature at all points in the reactor. Conducting these reactions in a batch reactor results in uncontrolled heat generation with potentially dangerous rises in pressure. Moreover, the microfluidic nature of these devices allows for flexible 2D designing of the flow channel. Four flow designs were examined (straight, periodically pinched, obtuse zigzag, and acute zigzag channels). The ability to use the channel pattern to increase the level of mixing throughout the reactor was evaluated. When moderately high molecular mass polymers with increased viscosity were made, the patterned channels produced polymers with narrower PDI, indicating that passive mixing arising from the channel design is improving the reaction conditions.

Iida, Kazunori; Chastek, Thomas Q.; Beers, Kathryn L.; Cavicchi, Kevin A.; Chun, Jaehun; Fasolka, Michael J.

2009-02-01

461

Oklo reactors and implications for nuclear science  

E-print Network

We summarize the nuclear physics interests in the Oklo natural nuclear reactors, focusing particularly on developments over the past two decades. Modeling of the reactors has become increasingly sophisticated, employing Monte Carlo simulations with realistic geometries and materials that can generate both the thermal and epithermal fractions. The water content and the temperatures of the reactors have been uncertain parameters. We discuss recent work pointing to lower temperatures than earlier assumed. Nuclear cross sections are input to all Oklo modeling and we discuss a parameter, the $^{175}$Lu ground state cross section for thermal neutron capture leading to the isomer $^{176\\mathrm{m}}$ Lu, that warrants further investigation. Studies of the time dependence of dimensionless fundamental constants have been a driver for much of the recent work on Oklo. We critically review neutron resonance energy shifts and their dependence on the fine structure constant $\\alpha$ and the ratio $X_q=m_q/\\Lambda$ (where $m_...

Davis, E D; Sharapov, E I

2014-01-01

462

Ultra high temperature particle bed reactor design  

NASA Technical Reports Server (NTRS)

A direct nuclear propulsion engine which could be used for a mission to Mars is designed. The main features of this reactor design are high values for I(sub sp) and very efficient cooling. This particle bed reactor consists of 37 cylindrical fuel elements embedded in a cylinder of beryllium which acts as a moderator and reflector. The fuel consists of a packed bed of spherical fissionable fuel particles. Gaseous H2 passes over the fuel bed, removes the heat, and is exhausted out of the rocket. The design was found to be neutronically critical and to have tolerable heating rates. Therefore, this particle bed reactor design is suitable as a propulsion unit for this mission.

Lazareth, Otto; Ludewig, Hans; Perkins, K.; Powell, J.

1990-01-01

463

Reactor safety for the space exploration initiative  

SciTech Connect

A task force was created by the National Aeronautics and Space Administration to conduct a 90-day study to support efforts to determine requirements to meet the goals of the Space Exploration Initiative. The task force identified the need for a nuclear reactor to provide the electrical power required as the outpost power demands on the Moon and Mars evolve into hundreds of kilowatts. A preliminary hazards analysis has been performed to examine safety aspects of nuclear reactor power systems for representative missions to the Moon and Mars. Mission profiles were defined for reference lunar and martian flights. Potential alternatives to each mission phase were also defined. Accident scenarios were qualitatively defined for the mission phases. The safety issues decay heat removal, reactor control, disposal, criticality, end-of-mission shutdown, radiation exposure, the martian environment, high speed impact on the surfaces of the Moon or Mars, and return flyby trajectories were identified.

Dix, T.E. (Rockwell International/Rocketdyne Division, 6633 Canoga Avenue, MS HB07, Canoga Park, California 91303 (US))

1991-01-01

464

Nuclear reactor shutdown control rod assembly  

DOEpatents

A temperature responsive, self-actuated nuclear reactor shutdown control rod assembly 10. The upper end 18 of a lower drive line 17 fits within the lower end of an upper drive line 12. The lower end (not shown) of the lower drive line 17 is connected to a neutron absorber. During normal temperature conditions the lower drive line 17 is supported by detent means 22,26. When an overtemperature condition occurs thermal actuation means 34 urges ring 26 upwardly sufficiently to allow balls 22 to move radially outwardly thereby allowing lower drive line 17 to move downwardly toward the core of the nuclear reactor resulting in automatic reduction of the reactor powder.

Bilibin, Konstantin (North Hollywood, CA)

1988-01-01

465

Reactor antineutrino background at Gran Sasso  

NASA Astrophysics Data System (ADS)

The flux of electron antineutrinos arriving at the Gran Sasso Laboratory underground solar-neutrino detector from the world-total 345 operating nuclear power reactors (total capacity 283 GWe) is estimated theoretically, extending and refining the analysis of Lagage (1985). The total is found to be about 450,000/sq cm s, of which about two thirds originate in reactors in Italy, France, and the FRG. From this result it is inferred that reactor electron antineutrinos become predominant below about 4 MeV when searching for elastic-coherent-scattering of high-energy solar neutrionos and below 8 MeV for diffuse-solar-neutrino background measurements, and that the high-energy tail of the earth antineutrino flux (1-3 MeV) cannot be detected. It is concluded that great efforts to lower the energy threshold of detectors may not be justified.

Lagage, P. O.

1986-04-01

466

Gaseous fuel reactors for power systems  

NASA Technical Reports Server (NTRS)

The Los Alamos Scientific Laboratory is participating in a NASA-sponsored program to demonstrate the feasibility of a gaseous uranium fueled reactor. The work is aimed at acquiring experimental and theoretical information for the design of a prototype plasma core reactor which will test heat removal by optical radiation. The basic goal of this work is for space applications, however, other NASA-sponsored work suggests several attractive applications to help meet earth-bound energy needs. Such potential benefits are: small critical mass, on-site fuel processing, high fuel burnup, low fission fragment inventory in reactor core, high temperature for process heat, optical radiation for photochemistry and space power transmission, and high temperature for advanced propulsion systems.

Helmick, H. H.; Schwenk, F. C.

1978-01-01

467

Gaseous fuel reactor systems for aerospace applications  

NASA Technical Reports Server (NTRS)

Research on the gaseous fuel nuclear rocket concept continues under the programs of the U.S. National Aeronautics and Space Administration (NASA) Office for Aeronautics and Space Technology and now includes work related to power applications in space and on earth. In a cavity reactor test series, initial experiments confirmed the low critical mass determined from reactor physics calculations. Recent work with flowing UF6 fuel indicates stable operation at increased power levels. Preliminary design and experimental verification of test hardware for high-temperature experiments have been accomplished. Research on energy extraction from fissioning gases has resulted in lasers energized by fission fragments. Combined experimental results and studies indicate that gaseous-fuel reactor systems have significant potential for providing nuclear fission power in space and on earth.

Thom, K.; Schwenk, F. C.

1977-01-01

468

Generic small modular reactor plant design.  

SciTech Connect

This report gives an overview of expected design characteristics, concepts, and procedures for small modular reactors. The purpose of this report is to provide those who are interested in reducing the cost and improving the safety of advanced nuclear power plants with a generic design that possesses enough detail in a non-sensitive manner to give merit to their conclusions. The report is focused on light water reactor technology, but does add details on what could be different in a more advanced design (see Appendix). Numerous reactor and facility concepts were used for inspiration (documented in the bibliography). The final design described here is conceptual and does not reflect any proposed concept or sub-systems, thus any details given here are only relevant within this report. This report does not include any design or engineering calculations.

Lewis, Tom Goslee,; Cipiti, Benjamin B.; Jordan, Sabina Erteza; Baum, Gregory A.

2012-12-01

469

Particle bed reactor nuclear rocket concept  

NASA Technical Reports Server (NTRS)

The particle bed reactor nuclear rocket concept consists of fuel particles (in this case (U,Zr)C with an outer coat of zirconium carbide). These particles are packed in an annular bed surrounded by two frits (porous tubes) forming a fuel element; the outer one being a cold frit, the inner one being a hot frit. The fuel element are cooled by hydrogen passing in through the moderator. These elements are assembled in a reactor assembly in a hexagonal pattern. The reactor can be either reflected or not, depending on the design, and either 19 or 37 elements, are used. Propellant enters in the top, passes through the moderator fuel element and out through the nozzle. Beryllium used for the moderator in this particular design to withstand the high radiation exposure implied by the long run times.

Ludewig, Hans

1991-01-01

470

Reactor safety for the Space Exploration Initiative  

NASA Technical Reports Server (NTRS)

A task force was created by the National Aeronautics and Space Administration to conduct a 90-day study to support efforts to determine requirements to meet the goals of the Space Exploration Initiative. The task force identified the need for a nuclear reactor to provide the electrical power required as the outpost power demands on the moon and Mars evolve into hundreds of kilowatts. A preliminary hazards analysis has been performed to examine safety aspects of nuclear reactor power systems for representative missions to the moon and Mars. Mission profiles were defined for reference lunar and Martian flights. Potential alternatives to each mission phase were also defined. Accident scenarios were qualitatively defined for the mission phases. The safety issues decay heat removal, reactor control, disposal, criticality, end-of-mission shutdown, radiation exposure, the Martian environment, high speed impact on the surfaces of the moon or Mars, and return flyby trajectories were identified.

Dix, Terry E.

1991-01-01

471

Plant maintenance and advanced reactors, 2005  

SciTech Connect

The focus of the September-October issue is on plant maintenance and advanced reactors. Major articles/reports in this issue include: First U.S. EPRs in 2015, by Ray Ganthner, Framatome ANP; Pursuing several opportunities, by William E. (Ed) Cummins, Westinghouse Electric Company; Vigorous plans to develop advanced reactors, by Yuliang Sun, Tsinghua University, China; Multiple designs, small and large, by Kumiaki Moriya, Hitachi Ltd., Japan; Sealed and embedded for safety and security, by Handa Norihiko, Toshiba Corporation, Japan; Scheduled online in 2010, by Johan Slabber, PMBR (Pty) Ltd., South Africa; Multi-application reactors, by Nikolay G. Kodochigov, OKBM, Russia; Six projects under budget and on schedule, by David F. Togerson, AECL, Canada; Creating a positive image, by Scott Peterson, Nuclear Energy Institute (NEI); Advanced plans for nuclear power's renaissance, by John Cleveland, International Atomic Energy Agency, Austria; and, Plant profile: last five outages in less than 20 days, by Beth Rapczynski, Exelon Nuclear.

Agnihotri, Newal (ed.)

2005-09-15

472

Mirror Advanced Reactor Study interim design report  

SciTech Connect

The status of the design of a tenth-of-a-kind commercial tandem-mirror fusion reactor is described at the midpoint of a two-year study. When completed, the design is to serve as a strategic goal for the mirror fusion program. The main objectives of the Mirror Advanced Reactor Study (MARS) are: (1) to design an attractive tandem-mirror fusion reactor producing electricity and synfuels (in alternate versions), (2) to identify key development and technology needs, and (3) to exploit the potential of fusion for safety, low activation, and simple disposal of radioactive waste. In the first year we have emphasized physics and engineering of the central cell and physics of the end cell. Design optimization and trade studies are continuing, and we expect additional modifications in the end cells to further improve the performance of the final design.

Not Available

1983-04-01

473

Monitoring and control of anaerobic reactors.  

PubMed

The current status in monitoring and control of anaerobic reactors is reviewed. The influence of reactor design and waste composition on the possible monitoring and control schemes is examined. After defining the overall control structure, and possible control objectives, the possible process measurements are reviewed in detail. In the sequel, possible manipulated variables, such as the hydraulic retention time, the organic loading rate, the sludge retention time, temperature, pH and alkalinity are evaluated with respect to the two main reactor types: high-rate and low-rate. Finally, the different control approaches that have been used are comprehensively described. These include simple and adaptive controllers, as well as more recent developments such as fuzzy controllers, knowledge-based controllers and controllers based on neural networks. PMID:12747567

Pind, Peter F; Angelidaki, Irini; Ahring, Birgitte K; Stamatelatou, Katerina; Lyberatos, Gerasimos

2003-01-01

474

The Modular Helium Reactor for Hydrogen Production  

SciTech Connect

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

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

2006-10-01

475

Nonlinear, inelastic fast reactor subassembly interaction analyses  

SciTech Connect

Liquid Metal Fast Breeder Reactor (LMFBR) core structural design is complicated by the trade-offs associated with keeping the subassemblies closely packed for the neutronic considerations and accommodating the volumetric changes associated with irradiation swelling. The environmental variation across the reactor core results in temperature and neutron flux gradients across the subassemblies which in turn cause the subassemblies to bow as well as dilate and grow volumetrically. These deformations in a tightly packed reactor core cause the subassemblies to interact and can potentially result in excessive withdrawal loads during the refueling operations. ABADAN, a general purpose, nonlinear, inelastic, multi-dimensional finite element structural analysis computer code, was developed for the express purpose of solving large nonlinear problems as typified by the above interaction problems. For the subassembly interaction problem ABADAN has been applied to the solution of an interacting radial row of Fast Flux Test Facility (FFTF) fuel assemblies.

Sutherland, W.H.; Bard, F.E.

1983-01-01

476

Modular Stellarator Fusion Reactor (MSR) concept  

SciTech Connect

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

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

1981-01-01

477

Small Inertial Fusion Energy (IFE) demonstration reactors  

SciTech Connect

ICF target design studies done for the Nova Upgrade have identified conditions under which the target ignition ``cliff`` is shifted to much lower drive energy albeit with the penalty that the gain achieved at a given energy is also smaller. These targets would repeatedly produce the output and spectra of a higher gain targets at low yield. They should, thus, allow building much smaller R&D reactors with full thermonuclear effects. Demonstration reactor at the 1 to 100 MW{sub e} level appear to be feasible with driver energies of 0.5 to 2.0 MJ per pulse. These smaller, less expensive test and demonstration facilities should result in lower IFE development cost. If the U.S. government builds a driver and target factory, it is also conceivable that commercial organizations could build their own scaled concepts of IFE reactors using the beams and targets supplied by the government`s facilities.

Hogan, W.J.

1991-10-03

478

Reactor Neutrino Experiments: $\\theta_{13}$ and Beyond  

E-print Network

We review the current-generation short-baseline reactor neutrino experiments that have firmly established the third neutrino mixing angle $\\theta_{13}$ to be non-zero. The relative large value of $\\theta_{13}$ (around 9$^\\circ$) has opened many new and exciting opportunities for future neutrino experiments. Daya Bay experiment with the first measurement of $\\Delta m^2_{ee}$ is aiming for a precision measurement of this atmospheric mass-squared splitting with a comparable precision as $\\Delta m^2_{\\mu\\mu}$ from accelerator muon neutrino experiments. JUNO, a next-generation reactor neutrino experiment, is targeting to determine the neutrino mass hierarchy with medium baselines ($\\sim$50 km). Beside these {\\color{black} opportunities enabled by the large $\\theta_{13}$}, the current-generation (Daya Bay, Double Chooz, and RENO) and the next-generation (JUNO, RENO-50, and PROSPECT) reactor experiments, with their unprecedented statistics, are also leading the precision era of the 3-flavor neutrino oscillation phys...

Qian, X

2014-01-01

479

Gas-Fast Reactor Fuel Fabrication  

SciTech Connect

The gas-cooled fast reactor is a high temperature helium cooled Generation IV reactor concept. Operating parameters for this type of reactor are well beyond those of current fuels so a novel fuel must be developed. One fuel concept calls for UC particles dispersed throughout a SiC matrix. This study examines a hybrid reaction bonding process as a possible fabrication route for this fuel. Processing parameters are also optimized. The process combines carbon and SiC powders and a carbon yielding polymer. In order to obtain dense reaction bonded SiC samples the porosity to carbon ratio in the preform must be large enough to accommodate SiC formation from the carbon present in the sample, however too much porosity reduces mechanical integrity which leads to poor infiltration properties . The porosity must also be of a suitable size to allow silicon transport throughout the sample but keep residual silicon to a minimum.

Randall Fielding; Mitchell Meyer; Ramprashad Prabhakaran; Jim Miller; Sean McDeavitt

2005-11-01

480

Reactor Coolant Pump seal issues and their applicability to new reactor designs  

SciTech Connect

Reactor Coolant Pumps (RCPs) of various types are used to circulate the primary coolant through the reactor in most reactor designs. RCPs generally contain mechanical seals to limit the leakage of pressurized reactor coolant along the pump drive shaft into the containment. The relatively large number of RCP seal and seal auxiliary system failures experienced at US operating plants during the 1970`s and early 1980`s raised concerns from the US Nuclear Regulatory Commission (NRC) that gross failures may lead to reactor core uncovery and subsequent core damage. Some seal failure events resulted in a loss of primary coolant to the containment at flow rates greater than the normal makeup capacity of Pressurized Water Reactor (PWR) plants. This is an example of RCP seal failures resulting in a small Loss of Coolant Accident (LOCA). This paper discusses observed and potential causes of RCP seal failure and the recommendations for limiting the likelihood of a seal induced small LOCA. Issues arising out of the research supporting these recommendations and subsequent public comments by the utility industry on them, serve as lessons learned, which are applicable to the design of new reactor plants.

Ruger, C.J.; Higgins, J.C.

1993-11-01

481

Low exchange element for nuclear reactor  

DOEpatents

A flow exchange element is presented which lowers temperature gradients in fuel elements and reduces maximum local temperature within high temperature gas-cooled reactors. The flow exchange element is inserted within a column of fuel elements where it serves to redirect coolant flow. Coolant which has been flowing in a hotter region of the column is redirected to a cooler region, and coolant which has been flowing in the cooler region of the column is redirected to the hotter region. The safety, efficiency, and longevity of the high temperature gas-cooled reactor is thereby enhanced.

Brogli, Rudolf H. (Aarau, CH); Shamasunder, Bangalore I. (Encinitas, CA); Seth, Shivaji S. (Encinitas, CA)

1985-01-01

482

Simulation of space particle bed reactors  

E-print Network

are isothermal devices. In another design of particle bed called Pellet Bed Reactor, the particles are packed between two frits in such a. manner that they would not move. This is a nevi and innovative concept built on an established pellet bed terrestrial... are isothermal devices. In another design of particle bed called Pellet Bed Reactor, the particles are packed between two frits in such a. manner that they would not move. This is a nevi and innovative concept built on an established pellet bed terrestrial...

Vincendon, Isabelle R.

2012-06-07

483

Perspectives on reactor safety. Revision 1  

SciTech Connect

The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor safety concepts. The course consists of five modules: (1) the development of safety concepts; (2) severe accident perspectives; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

Haskin, F.E. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering; Camp, A.L. [Sandia National Labs., Albuquerque, NM (United States); Hodge, S.A. [Oak Ridge National Lab., TN (United States). Engineering Technology Div.

1997-11-01

484

Summary of the mirror advanced reactor study  

SciTech Connect

The Mirror Advanced Reactor Study (MARS) is a conceptual design of a 1200-MWe commercial tandem mirror reactor for electricity and synfuels (methanol) production. Thermal barrier end plugs of the TMX-U/MFTF-B type allow steady-state ignition of a 130-m-long central-cell DT plasma. Compact, gridless direct converters supply all the plant auxiliary power. The simple lead-lithium eutectic-cooled blanket has high neutron energy multiplication (1.36) as well as a low tritium inventory (< 8 g), and it will not melt in accidents.

Logan, B.G.; Henning, C.D.; Carlson, G.A.; Gordon, J.D.; Maniscalco, J.A.; Kulcinski, G.L.; Perkins, L.J.; Parmer, J.F.; Bilton, J.R.; Glancy, J.E.

1984-07-26

485

Removing Undesired Fine Powder From Silicon Reactor  

NASA Technical Reports Server (NTRS)

Fluidized-bed reactor produces highly pure polycrystalline silicon particles with diameters approximately greater than 400 micrometers. Operates by pyrolysis of silane in reaction zone, which is bed of silicon seed particles fluidized by flow of silane and carrier gas. Above reaction zone, gas mixture flows rapidly enough to entrain silicon powders, but not larger seed and product particles. Entrained particles swept out of reactor. Applicable to other processes such as production of fine metal and ceramic powders where control of sizes of product needed.

Flagella, Robert N.

1992-01-01

486

DOE/NE robotics for advanced reactors  

SciTech Connect

This bimonthly progress report for February through March, 1991, for the US Department of Energy Robotics for Advanced Reactors program contains information on research efforts in the following areas at the noted facilities: modeling of the Fuel Cycle Facility of the Integral Fast Reactor (IFR) program (University of Florida), assembly of infrared range sensors (University of Michigan), demonstration of the autonomous surveillance of a waste storage container for radioactive surface contamination (Oak Ridge National Laboratory), development of a simulation and animation environment for sensor-based robots (University of Tennessee), and machining and installation of the actuator test stand and associated fixtures and adapters (University of Texas). (MHB)

Not Available

1991-01-01

487

Plasma arc process systems, reactors, and applications  

SciTech Connect

Thermal plasma arc systems are being applied to a wide variety of high-temperature applications. The use of electricity is proving to be a cost-effective alternative for the treatment of industrial wastes, recycling of scrap, upgrading of existing metallurgical processes, and new methods for the production of materials. Thermal plasma arc reactors are classified according to: (1) the mode of arc attachment, as non-transferred and transferred: and (2) the site for energy and/or mass transfer, as dispersed phase or condensed phase (bulk, film and packed bed). Unique features in the design of plasma reactor systems and applications in waste treatment and metals production are discussed.

Mac Rae, D.R. (Bethlehem Steel Corp., PA (USA))

1989-03-01

488

Polymerization in emulsion microdroplet reactors  

NASA Astrophysics Data System (ADS)

The goal of this research project is to utilize emulsion droplets as chemical reactors for execution of complex polymerization chemistries to develop unique and functional particle materials. Emulsions are dispersions of immiscible fluids where one fluid usually exists in the form of drops. Not surprisingly, if a liquid-to-solid chemical reaction proceeds to completion within these drops, the resultant solid particles will possess the shape and relative size distribution of the drops. The two immiscible liquid phases required for emulsion polymerization provide unique and complex chemical and physical environments suitable for the engineering of novel materials. The development of novel non-ionic fluorosurfactants allows fluorocarbon oils to be used as the continuous phase in a water-free emulsion. Such emulsions enable the encapsulation of almost any hydrocarbon compound in droplets that may be used as separate compartments for water-sensitive syntheses. Here, we exemplify the promise of this approach by suspension polymerization of polyurethanes (PU), in which the liquid precursor is emulsified into droplets that are then converted 1:1 into polymer particles. The stability of the droplets against coalescence upon removal of the continuous phase by evaporation confirms the formation of solid PU particles. These results prove that the water-free environment of fluorocarbon based emulsions enables high conversion. We produce monodisperse, cross-linked, and fluorescently labeled PU-latexes with controllable mesh size through microfluidic emulsification in a simple one-step process. A novel method for the fabrication of monodisperse mesoporous silica particles is presented. It is based on the formation of well-defined equally sized emulsion droplets using a microfluidic approach. The droplets contain the silica precursor/surfactant solution and are suspended in hexadecane as the continuous oil phase. The solvent is then expelled from the droplets, leading to concentration and micellization of the surfactant. At the same time, the silica solidifies around the surfactant structures, forming equally sized mesoporous particles. The procedure can be tuned to produce well-separated particles or alternatively particles that are linked together. The latter allows us to create 2D or 3D structures with hierarchical porosity. Oil, water, and surfactant liquid mixtures exhibit very complex phase behavior. Depending on the conditions, such mixtures give rise to highly organized structures. A proper selection of the type and concentration of surfactants determines the structuring at the nanoscale level. In this work, we show that hierarchically bimodal nanoporous structures can be obtained by templating silica microparticles with a specially designed surfactant micelle/microemulsion mixture. Tuning the phase state by adjusting the surfactant composition and concentration allows for the controlled design of a system where microemulsion droplets coexist with smaller surfactant micellar structures. The microemulsion droplet and micellar dimensions determine the two types of pore sizes (single nanometers and tens of nanometers). We also demonstrate the fabrication of carbon and carbon/platinum replicas of the silica microspheres using a "lost-wax" approach. Such particles have great potential for the design of electrocatalysts for fuel cells, chromatography separations, and other applications. It was determined that slight variations in microemulsion mixture components (electrolyte concentration, wt% of surfactants, oil to sol ratio, etc.) produces strikingly different pore morphologies and particle surface areas. Control over the size and structure of the smaller micelle-templated pores was made possible by varying the length of the hydrocarbon block within the trimethyl ammonium bromide surfactant and characterized using X-ray diffraction. The effect of emulsion aging was studied by synthesizing particles at progressive time levels from a sample emulsion. It was discovered surface pore size increases after just a few hours, with

Carroll, Nick J.

489

Daddy, What's a Nuclear Reactor?  

SciTech Connect

No matter what we think of the nuclear industry, it is part of mankind's heritage. The decommissioning process is slowly making facilities associated with this industry disappear and not enough is being done to preserve the information for future generations. This paper provides some food for thought and provides a possible way forward. Industrial archaeology is an ever expanding branch of archaeology that is dedicated to preserving, interpreting and documenting our industrial past and heritage. Normally it begins with analyzing an old building or ruins and trying to determine what was done, how it was done and what changes might have occurred during its operation. We have a unique opportunity to document all of these issues and provide them before the nuclear facility disappears. Entombment is an acceptable decommissioning strategy; however we would have to change our concept of entombment. It is proposed that a number of nuclear facilities be entombed or preserved for future generations to appreciate. This would include a number of different types of facilities such as different types of nuclear power and research reactors, a reprocessing plant, part of an enrichment plant and a fuel manufacturing plant. One of the main issues that would require resolution would be that of maintaining information of the location of the buried facility and the information about its operation and structure, and passing this information on to future generations. This can be done, but a system would have to be established prior to burial of the facility so that no information would be lost. In general, our current set of requirements and laws may need to be re-examined and modified to take into account these new situations. As an alternative, and to compliment the above proposal, it is recommended that a study and documentation of the nuclear industry be considered as part of twentieth century industrial archaeology. This study should not only include the power and fuel cycle facilities, but also the nuclear weapons complex and the industrial and research sectors. This would be a large chore due to the considerable number of different types of facilities that have been used in these industries, but it would be a worthwhile endeavor. This study would gather information that would normally be lost due to the decommissioning process and allow future generations to appreciate these industries. Because of the volume and varying types of facilities, it might be more beneficial to produce a set of studies relating to different aspects of the industry. A logical division would be the separation of the commercial nuclear industry and the nuclear weapons complex. The separation of the fuel cycle facilities may also be considered. If done properly, this could result in a set of documents of interest to a wide audience. The current nuclear industry is slowly disappearing through the decommissioning process. This industry is unique and is part of mankind's heritage. It must not be forgotten and the information should be made available for future generations. The U.S. Department of Energy and the National Park Service are doing some limited preservation of information, but I do not believe its enough. It is not being done in a manner that will preserve the true activities that were performed. It is recommended that the American Nuclear Society, along with other organizations, evaluate this proposal and possibly provide funds for a set of studies to be prepared and ensure that this valuable part of our heritage is not lost.

Reisenweaver, Dennis W. [Alion Science and Technology, 1475 Central Ave., Los Alamos, NM 87544 (United States)

2008-01-15

490

Technology gap analysis on sodium-cooled reactor fuel handling system supporting advanced burner reactor development.  

SciTech Connect

The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand in an environmentally sustainable manner, to address nuclear waste management issues without making separated plutonium, and to address nonproliferation concerns. The advanced burner reactor (ABR) is a fast reactor concept which supports the GNEP fuel cycle system. Since the integral fast reactor (IFR) and advanced liquid-metal reactor (ALMR) projects were terminated in 1994, there has been no major development on sodium-cooled fast reactors in the United States. Therefore, in support of the GNEP fast reactor program, the history of sodium-cooled reactor development was reviewed to support the initiation of this technology within the United States and to gain an understanding of the technology gaps that may still remain for sodium fast reactor technology. The fuel-handling system is a key element of any fast reactor design. The major functions of this system are to receive, test, store, and then load fresh fuel into the core; unload from the core; then clean, test, store, and ship spent fuel. Major requirements are that the system must be reliable and relatively easy to maintain. In addition, the system should be designed so that it does not adversely impact plant economics from the viewpoints of capital investment or plant operations. In this gap analysis, information on fuel-handling operating experiences in the following reactor plants was carefully reviewed: EBR-I, SRE, HNPF, Fermi, SEFOR, FFTF, CRBR, EBR-II, DFR, PFR, Rapsodie, Phenix, Superphenix, KNK, SNR-300, Joyo, and Monju. The results of this evaluation indicate that a standardized fuel-handling system for a commercial fast reactor is yet to be established. However, in the past sodium-cooled reactor plants, most major fuel-handling components-such as the rotatable plug, in-vessel fuel-handling machine, ex-vessel fuel transportation cask, ex-vessel sodium-cooled storage, and cleaning stations-have accumulated satisfactory construction and operation experiences. In addition, two special issues for future development are described in this report: large capacity interim storage and transuranic-bearing fuel handling.

Chikazawa, Y.; Farmer, M.; Grandy, C.; Nuclear Engineering Division

2009-03-01

491

Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors  

SciTech Connect

Background: The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of Generation IV reactor systems such as supercritical water reactors (SCWR) for higher efficiency, improved performance and operation, design simplification, enhanced safety and reduced waste and cost. The objective of this Korean / US / laboratory / university collaboration of coupled fundamental computational and experimental studies is to develop the supporting knowledge needed for improved predictive techniques for use in the technology development of Generation IV reactor concepts and their passive safety systems. The present study emphasizes SCWR concepts in the Generation IV program.

D.M. McEligot; K. G. Condie; G. E. McCreery; H. M. McIlroy; R. J. Pink; L.E. Hochreiter; J.D. Jackson; R.H. Pletcher; B.L. Smith; P. Vukoslavcevic; J.M. Wallace; J.Y. Yoo; J.S. Lee; S.T. Ro; S.O. Park

2005-10-01

492

Advanced burner test reactor preconceptual design report.  

SciTech Connect

The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand, to address nuclear waste management concerns and to promote non-proliferation. Implementation of the GNEP requires development and demonstration of three major technologies: (1) Light water reactor (LWR) spent fuel separations technologies that will recover transuranics to be recycled for fuel but not separate plutonium from other transuranics, thereby providing proliferation-resistance; (2) Advanced Burner Reactors (ABRs) based on a fast spectrum that transmute the recycled transuranics to produce energy while also reducing the long term radiotoxicity and decay heat loading in the repository; and (3) Fast reactor fuel recycling technologies to recover and refabricate the transuranics for repeated recycling in the fast reactor system. The primary mission of the ABR Program is to demonstrate the transmutation of transuranics recovered from the LWR spent fuel, and hence the benefits of the fuel cycle closure to nuclear waste management. The transmutation, or burning of the transuranics is accomplished by fissioning and this is most effectively done in a fast spectrum. In the thermal spectrum of commercial LWRs, some transuranics capture neutrons and become even heavier transuranics rather than being fissioned. Even with repeated recycling, only about 30% can be transmuted, which is an intrinsic limitation of all thermal spectrum reactors. Only in a fast spectrum can all transuranics be effectively fissioned to eliminate their long-term radiotoxicity and decay heat. The Advanced Burner Test Reactor (ABTR) is the first step in demonstrating the transmutation technologies. It directly supports development of a prototype full-scale Advanced Burner Reactor, which would be followed by commercial deployment of ABRs. The primary objectives of the ABTR are: (1) To demonstrate reactor-based transmutation of transuranics as part of an advanced fuel cycle; (2) To qualify the transuranics-containing fuels and advanced structural materials needed for a full-scale ABR; and (3) To support the research, development and demonstration required for certification of an ABR standard design by the U.S. Nuclear Regulatory Commission. The ABTR should also address the following additional objectives: (1) To incorporate and demonstrate innovative design concepts and features that may lead to significant improvements in cost, safety, efficiency, reliability, or other favorable characteristics that could promote public acceptance and future private sector investment in ABRs; (2) To demonstrate improved technologies for safeguards and security; and (3) To support development of the U.S. infrastructure for design, fabrication and construction, testing and deployment of systems, structures and components for the ABRs. Based on these objectives, a pre-conceptual design of a 250 MWt ABTR has been developed; it is documented in this report. In addition to meeting the primary and additional objectives listed above, the lessons learned from fast reactor programs in the U.S. and worldwide and the operating experience of more than a dozen fast reactors around the world, in particular the Experimental Breeder Reactor-II have been incorporated into the design of the ABTR to the extent possible.

Chang, Y. I.; Finck, P. J.; Grandy, C.; Cahalan, J.; Deitrich, L.; Dunn, F.; Fallin, D.; Farmer, M.; Fanning, T.; Kim, T.; Krajtl, L.; Lomperski, S.; Moisseytsev, A.; Momozaki, Y.; Sienicki, J.; Park, Y.; Tang, Y.; Reed, C.; Tzanos, C; Wiedmeyer, S.; Yang, W.; Chikazawa, Y.; JAEA

2008-12-16

493

Optimization algorithms in boiling water reactor lattice design  

E-print Network

Given the highly complex nature of neutronics and