Powered by Deep Web Technologies
Note: This page contains sample records for the topic "reactor components shield" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


1

REACTOR SHIELD  

DOE Patents (OSTI)

Radiation shield construction is described for a nuclear reactor. The shield is comprised of a plurality of steel plates arranged in parallel spaced relationship within a peripheral shell. Reactor coolant inlet tubes extend at right angles through the plates and baffles are arranged between the plates at right angles thereto and extend between the tubes to create a series of zigzag channels between the plates for the circulation of coolant fluid through the shield. The shield may be divided into two main sections; an inner section adjacent the reactor container and an outer section spaced therefrom. Coolant through the first section may be circulated at a faster rate than coolant circulated through the outer section since the area closest to the reactor container is at a higher temperature and is more radioactive. The two sections may have separate cooling systems to prevent the coolant in the outer section from mixing with the more contaminated coolant in the inner section.

Wigner, E.P.; Ohlinger, L.E.; Young, G.J.; Weinberg, A.M.

1959-02-17T23:59:59.000Z

2

NEUTRONIC REACTOR SHIELD  

DOE Patents (OSTI)

The reactor radiation shield material is comprised of alternate layers of iron-containing material and compressed cellulosic material, such as masonite. The shielding material may be prefabricated in the form of blocks, which can be stacked together in ary desired fashion to form an effective shield.

Fermi, E.; Zinn, W.H.

1957-09-24T23:59:59.000Z

3

MEANS FOR SHIELDING REACTORS  

DOE Patents (OSTI)

A reactor of the heterageneous, heavy water moderated type is described. The reactor is comprised of a plurality of vertically disposed fuel element tubes extending through a tank of heavy water moderator and adapted to accommodate a flow of coolant water in contact with the fuel elements. A tank containing outgoing coolant water is disposed above the core to function is a radiation shield. Unsaturated liquid hydrocarbon is floated on top of the water in the shield tank to reduce to a minimum the possibility of the occurrence of explosive gaseous mixtures resulting from the neutron bombardment of the water in the shield tank.

Garrison, W.M.; McClinton, L.T.; Burton, M.

1959-03-10T23:59:59.000Z

4

Nuclear reactor shield including magnesium oxide  

DOE Patents (OSTI)

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

Rouse, Carl A. (Del Mar, CA); Simnad, Massoud T. (La Jolla, CA)

1981-01-01T23:59:59.000Z

5

Neutronic reactor thermal shield  

DOE Patents (OSTI)

1. The combination with a plurality of parallel horizontal members arranged in horizontal and vertical rows, the spacing of the members in all horizontal rows being equal throughout, the spacing of the members in all vertical rows being equal throughout; of a shield for a nuclear reactor comprising two layers of rectangular blocks through which the members pass generally perpendicularly to the layers, each block in each layer having for one of the members an opening equally spaced from vertical sides of the block and located closer to the top of the block than the bottom thereof, whereby gravity tends to make each block rotate about the associated member to a position in which the vertical sides of the block are truly vertical, the openings in all the blocks of one layer having one equal spacing from the tops of the blocks, the openings in all the blocks of the other layer having one equal spacing from the tops of the blocks, which spacing is different from the corresponding spacing in the said one layer, all the blocks of both layers having the same vertical dimension or length, the blocks of both layers consisting of relatively wide blocks and relatively narrow blocks, all the narrow blocks having the same horizontal dimension or width which is less than the horizontal dimension or width of the wide blocks, which is the same throughout, each layer consisting of vertical rows of narrow blocks and wide blocks alternating with one another, each vertical row of narrow blocks of each layer being covered by a vertical row of wide blocks of the other layer which wide blocks receive the same vertical row of members as the said each vertical row of narrow blocks, whereby the rectangular perimeters of each block of each layer is completely out of register with that of each block in the other layer.

Lowe, Paul E. (Blue Ash, OH)

1976-06-15T23:59:59.000Z

6

Neutronic reactor thermal shield  

DOE Patents (OSTI)

1. The method of operating a water-cooled neutronic reactor having a graphite moderator which comprises flowing a gaseous mixture of carbon dioxide and helium, in which the helium comprises 40-60 volume percent of the mixture, in contact with the graphite moderator.

Wende, Charles W. J. (West Chester, PA)

1976-06-15T23:59:59.000Z

7

NEUTRON REACTOR HAVING A Xe$sup 135$ SHIELD  

DOE Patents (OSTI)

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

Stanton, H.E.

1957-10-29T23:59:59.000Z

8

Where are we and where are we going in reactor shielding  

SciTech Connect

Information on reactor shielding is presented concerning nuclear data in shielding calculations, analysis methods, water-cooled reactors, fast-breeder reactors, and fusion reactors.

Maienschein, F.C.

1977-01-01T23:59:59.000Z

9

Gravity Scaling of a Power Reactor Water Shield  

Science Conference Proceedings (OSTI)

Water based reactor shielding is being considered as an affordable option for potential use on initial lunar surface reactor power systems. Heat dissipation in the shield from nuclear sources must be rejected by an auxillary thermal hydraulic cooling system. The mechanism for transferring heat through the shield is natural convection between the core surface and an array of thermosyphon radiator elements. Natural convection in a 100 kWt lunar surface reactor shield design has been previously evaluated at lower power levels (Pearson

Robert S. Reid; J. Boise Pearson

2008-01-01T23:59:59.000Z

10

Reactor component automatic grapple  

DOE Patents (OSTI)

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

Greenaway, Paul R. (Bethel Park, PA)

1982-01-01T23:59:59.000Z

11

Shielding considerations for advanced space nuclear reactor systems  

SciTech Connect

To meet the anticipated future space power needs, the Los Alamos National Laboratory is developing components for a compact, 100 kW/sub e/-class heat pipe nuclear reactor. The reactor uses uranium dioxide (UO/sub 2/) as its fuel, and is designed to operate around 1500 k. Heat pipes are used to remove thermal energy from the core without the use of pumps or compressors. The reactor heat pipes transfer mal energy to thermoelectric conversion elements that are advanced versions of the converters used on the enormously successful Voyager missions to the outer planets. Advanced versions of this heat pipe reactor could also be used to provide megawatt-level power plants. The paper reviews the status of this advanced heat pipe reactor and explores the radiation environments and shielding requirements for representative manned and unmanned applications.

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

1982-01-01T23:59:59.000Z

12

Recovery Act Workers Clear Reactor Shields from Brookhaven Lab | Department  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Workers Clear Reactor Shields from Brookhaven Lab Workers Clear Reactor Shields from Brookhaven Lab Recovery Act Workers Clear Reactor Shields from Brookhaven Lab American Recovery and Reinvestment Act workers are in the final stage of decommissioning a nuclear reactor after they recently removed thick steel shields once used to absorb neutrons produced for research. The Brookhaven National Laboratory is using $39 million from the Recovery Act to decommission the Brookhaven Graphite Research Reactor, the world's first reactor built solely for peaceful research purposes. Recovery Act Workers Clear Reactor Shields from Brookhaven Lab More Documents & Publications Brookhaven Graphite Research Reactor Workshop 2011 ARRA Newsletters Idaho Crews Overcome Challenges to Safely Dispose 1-Million-Pound Hot Cell

13

Space Reactor Radiation Shield Design Summary, for Information  

SciTech Connect

The purpose of this letter is to provide a summary of the Prometheus space reactor radiation shield design status at the time of program restructuring.

EC Pheil

2006-02-17T23:59:59.000Z

14

Recovery Act Workers Clear Reactor Shields from Brookhaven Lab...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Brookhaven Lab American Recovery and Reinvestment Act workers are in the final stage of decommissioning a nuclear reactor after they recently removed thick steel shields once...

15

Packed rod neutron shield for fast nuclear reactors  

DOE Patents (OSTI)

A fast neutron nuclear reactor including a core and a plurality of vertically oriented neutron shield assemblies surrounding the core. Each assembly includes closely packed cylindrical rods within a polygonal metallic duct. The shield assemblies are less susceptible to thermal stresses and are less massive than solid shield assemblies, and are cooled by liquid coolant flow through interstices among the rods and duct.

Eck, John E. (Hempfield Township, Westmoreland County, PA); Kasberg, Alvin H. (Murrysville, PA)

1978-01-01T23:59:59.000Z

16

Gravity Scaling of a Power Reactor Water Shield  

SciTech Connect

Water based reactor shielding is being considered as an affordable option for potential use on initial lunar surface reactor power systems. Heat dissipation in the shield from nuclear sources must be rejected by an auxillary thermal hydraulic cooling system. The mechanism for transferring heat through the shield is natural convection between the core surface and an array of thermosyphon radiator elements. Natural convection in a 100 kWt lunar surface reactor shield design has been previously evaluated at lower power levels (Pearson, 2006). The current baseline assumes that 5.5 kW are dissipated in the water shield, the preponderance on the core surface, but with some volumetric heating in the naturally circulating water as well. This power is rejected by a radiator located above the shield with a surface temperature of 370 K. A similarity analysis on a water-based reactor shield is presented examining the effect of gravity on free convection between a radiation shield inner vessel and a radiation shield outer vessel boundaries. Two approaches established similarity: 1) direct scaling of Rayleigh number equates gravity-surface heat flux products, 2) temperature difference between the wall and thermal boundary layer held constant on Earth and the Moon. Nussult number for natural convection (laminar and turbulent) is assumed of form Nu = CRa{sup n}. These combined results estimate similarity conditions under Earth and Lunar gravities. The influence of reduced gravity on the performance of thermosyphon heat pipes is also examined.

Reid, Robert S.; Pearson, J. Boise [NASA Marshall Space Flight Center, Huntsville, AL 35812 (United States)

2008-01-21T23:59:59.000Z

17

Recovery Act Workers Clear Reactor Shields from Brookhaven Lab  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

UPTON, N.Y. - American Recovery and Reinvestment Act UPTON, N.Y. - American Recovery and Reinvestment Act workers are in the final stage of decommissioning a nuclear reactor after they recently removed thick steel shields once used to absorb neutrons produced for research. The Brookhaven National Laboratory is using $39 million from the Recovery Act to decommission the Brookhaven Graphite Research Reactor, the world's first reactor built solely for peaceful research purposes. The decommissioning is slated for completion later this year and will end Office of Environmental Management legacy cleanup activities at the Lab. The neutron shields were located on the north and south sides of a 700-ton graphite pile. The three-inch-thick shields absorbed neutrons that escaped from the graphite pile. The shields also limited movement of the pile when the reactor was in opera-

18

Effect of Aggregate Specification in Heavy Concrete for Fusion Reactor Shield on Neutron Dose Evaluation  

E-Print Network (OSTI)

Effect of Aggregate Specification in Heavy Concrete for Fusion Reactor Shield on Neutron Dose Evaluation

Murata, I; Takahashi, A

2000-01-01T23:59:59.000Z

19

Shielded fluid stream injector for particle bed reactor  

DOE Patents (OSTI)

A shielded fluid-stream injector assembly is provided for particle bed reactors. The assembly includes a perforated pipe injector disposed across the particle bed region of the reactor and an inverted V-shaped shield placed over the pipe, overlapping it to prevent descending particles from coming into direct contact with the pipe. The pipe and shield are fixedly secured at one end to the reactor wall and slidably secured at the other end to compensate for thermal expansion. An axially extending housing aligned with the pipe and outside the reactor and an inline reamer are provided for removing deposits from the inside of the pipe. The assembly enables fluid streams to be injected and distributed uniformly into the particle bed with minimized clogging of injector ports. The same design may also be used for extraction of fluid streams from particle bed reactors.

Notestein, J.E.

1991-12-31T23:59:59.000Z

20

An Experimental Shield Test Facility for the Development of Minimum Weight Shields for Compact Reactor Power Systems  

SciTech Connect

Discussions are given of the characteristics of fission-source plate, graphite reactor, and pool-type reactor facilities applicable to development studies of minimum weight shielding materials. Advantages of a proposed SNAP dual-purpose shielding facility are described in terms of a disk-shaped fission-source plate, reactor, and building. A program for the study of advanced shielding materials is discussed for materials and configuations to be evaluted with the fission-source plate, the testing of the prototype at high-power levels, and full-power tests on the actual reactor.

Tomlinson, R.L.

1959-08-07T23:59:59.000Z

Note: This page contains sample records for the topic "reactor components shield" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


21

A STUDY OF THE PRIMARY SHIELD FOR THE PRDC REACTOR  

SciTech Connect

Temperature distributions, irradiation effects, stacking arrangements, voidage, and economics for the borated-graphite shield of the PRDC reactor were investigated. Of the shield systems considered, four are reported here. System 1 contalns 30 in. of 1% borated graphite, with either ordinary graphite or a cement as a filler for the remaindcr of the volume. The maximum temperature at the flex plates in this system was calculated to be 5OO deg F. Systems 2 and 3 consist of 2 in. of 5% borated graphite near the core vessel and 1/2 in. of Boral at the primary-shield tank. A filler material of carbon blocks is used in System 2 and graphite in System 3. The calculated maximum temperatures were 700 deg F and 35O deg F, respectively. System 4 consists of a laminated structure of Boral and graphite near the primary-shield tank and carbon-block filler. It was calculated to have a maximum temperature of 600 deg F at the flex plates. The maximum temperature at the flex plates recommended by APDA is 500 deg F. Energy storage and radiation damage were found to be within permissible limits in all four systems. However, these conclusions are based on experimental data from the Hanford reactor in which the neutron-energy spectrum differs considerably from the PRDC spectrum. A porosity of less than 740 cu ft is required in order that a sodium leak from the core vessel does not expose the core. The voidages in any of the systems mentioned above is about 400 cu ft excluding absorption effects. these are believed to be small. The systems coataining Boral were found to be less expensive than the ones using only borated graphite. Over-all material costs range between 0,000 for Boral systems and 0,000 for borated- graphite systems. (auth)

Epstein, H.M.; Dingee, D.A.; Chastain, J.W.

1957-04-15T23:59:59.000Z

22

Radiation Shielding Analysis for Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors (DUPIC)  

Science Conference Proceedings (OSTI)

As a part of the compatibility analysis of DUPIC fuel in Canada deuterium uranium (CANDU) reactors, the radiation physics calculations have been performed for the CANDU primary shielding system, which was originally designed for natural uranium core. At first, the conventional CANDU primary shield analysis method was validated using the Monte Carlo code MCNP-4B in order to assess the current analysis code system and the cross-section data. The computational benchmark calculation was performed for the CANDU end shield system, which has shown that the conventional method produces results consistent with the reference calculations as far as the total dose rate and total heat deposition rate are concerned. Second, the primary shield system analysis was performed for the DUPIC fuel core based on the power distribution obtained from the time-average core model, and the results have shown that the dose rates and heat deposition rates through the primary shield of the DUPIC fuel core are not much different from those of the natural uranium core because the power levels on the core periphery are similar for both cores. This study has shown that the current primary shield system is adaptable for the DUPIC fuel CANDU core without design modification.

Roh, Gyuhong; Choi, Hangbok [Korea Atomic Energy Research Institute (Korea, Republic of)

2004-06-15T23:59:59.000Z

23

Nuclear reactor having a polyhedral primary shield and removable vessel insulation  

DOE Patents (OSTI)

A nuclear reactor is provided having a generally cylindrical reactor vessel disposed within an opening in a primary shield. The opening in the primary shield is defined by a plurality of generally planar side walls forming a generally polyhedral-shaped opening. The reactor vessel is supported within the opening in the primary shield by reactor vessel supports which are in communication and aligned with central portions of some of the side walls. The reactor vessel is connected to the central portions of the reactor vessel supports. A thermal insulation polyhedron formed from a plurality of slidably insertable and removable generally planar insulation panels substantially surrounds at least a portion of the reactor vessel and is disposed between the reactor vessel and the side walls of the primary shield. The shape of the insulation polyhedron generally corresponds to the shape of the opening in the primary shield. Reactor monitoring instrumentation may be mounted in the corners of the opening in the primary shield between the side walls and the reactor vessel such that insulation is not disposed between the instrumentation and the reactor vessel. 5 figures.

Ekeroth, D.E.; Orr, R.

1993-12-07T23:59:59.000Z

24

Nuclear reactor having a polyhedral primary shield and removable vessel insulation  

DOE Patents (OSTI)

A nuclear reactor is provided having a generally cylindrical reactor vessel disposed within an opening in a primary shield. The opening in the primary shield is defined by a plurality of generally planar side walls forming a generally polyhedral-shaped opening. The reactor vessel is supported within the opening in the primary shield by reactor vessel supports which are in communication and aligned with central portions of some of the side walls. The reactor vessel is connected to the central portions of the reactor vessel supports. A thermal insulation polyhedron formed from a plurality of slidably insertable and removable generally planar insulation panels substantially surrounds at least a portion of the reactor vessel and is disposed between the reactor vessel and the side walls of the primary shield. The shape of the insulation polyhedron generally corresponds to the shape of the opening in the primary shield. Reactor monitoring instrumentation may be mounted in the corners of the opening in the primary shield between the side walls and the reactor vessel such that insulation is not disposed between the instrumentation and the reactor vessel.

Ekeroth, Douglas E. (Delmont, PA); Orr, Richard (Pittsburgh, PA)

1993-01-01T23:59:59.000Z

25

Solid tags for identifying failed reactor components  

DOE Patents (OSTI)

A solid tag material which generates stable detectable, identifiable, and measurable isotopic gases on exposure to a neutron flux to be placed in a nuclear reactor component, particularly a fuel element, in order to identify the reactor component in event of its failure. Several tag materials consisting of salts which generate a multiplicity of gaseous isotopes in predetermined ratios are used to identify different reactor components.

Bunch, Wilbur L. (Richland, WA); Schenter, Robert E. (Richland, WA)

1987-01-01T23:59:59.000Z

26

Space reactor system and subsystem investigations: assessment of technology issues for the reactor and shield subsystem. SP-100 Program  

SciTech Connect

As part of Rockwell's effort on the SP-100 Program, preliminary assessment has been completed of current nuclear technology as it relates to candidate reactor/shield subsystems for the SP-100 Program. The scope of the assessment was confined to the nuclear package (to the reactor and shield subsystems). The nine generic reactor subsystems presented in Rockwell's Subsystem Technology Assessment Report, ESG-DOE-13398, were addressed for the assessment.

Atkins, D.F.; Lillie, A.F.

1983-06-30T23:59:59.000Z

27

ANS shielding standards for light-water reactors  

SciTech Connect

The purpose of the American Nuclear Society Standards Subcommittee, ANS-6, Radiation Protection and Shielding, is to develop standards for radiation protection and shield design, to provide shielding information to other standards-writing groups, and to develop standard reference shielding data and test problems. A total of seven published ANS-6 standards are now current. Additional projects of the subcommittee, now composed of nine working groups, include: standard reference data for multigroup cross sections, gamma-ray absorption coefficients and buildup factors, additional benchwork problems for shielding problems and energy spectrum unfolding, power plant zoning design for normal and accident conditions, process radiation monitors, and design for postaccident radiological conditions.

Trubey, D.K.

1982-01-01T23:59:59.000Z

28

EXPERIMENTAL EVALUATION OF THE THERMAL PERFORMANCE OF A WATER SHIELD FOR A SURFACE POWER REACTOR  

SciTech Connect

Water based reactor shielding is being investigated for use on initial lunar surface power systems. A water shield may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. Natural convection in a 100 kWt lunar surface reactor shield design is evaluated with 2 kW power input to the water in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The experimental data from the WST is used to validate a CFD model. Performance of the water shield on the lunar surface is then predicted with a CFD model anchored to test data. The experiment had a maximum water temperature of 75 C. The CFD model with 1/6-g predicts a maximum water temperature of 88 C with the same heat load and external boundary conditions. This difference in maximum temperature does not greatly affect the structural design of the shield, and demonstrates that it may be possible to use water for a lunar reactor shield.

REID, ROBERT S. [Los Alamos National Laboratory; PEARSON, J. BOSIE [Los Alamos National Laboratory; STEWART, ERIC T. [Los Alamos National Laboratory

2007-01-16T23:59:59.000Z

29

Experimental Evaluation of the Thermal Performance of a Water Shield for a Surface Power Reactor  

SciTech Connect

Water based reactor shielding is being investigated for use on initial lunar surface power systems. A water shield may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. Natural convection in a 100 kWt lunar surface reactor shield design is evaluated with 2 kW power input to the water in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The experimental data from the WST is used to validate a CFD model. Performance of the water shield on the lunar surface is then predicted with a CFD model anchored to test data. The experiment had a maximum water temperature of 75 deg. C. The CFD model with 1/6-g predicts a maximum water temperature of 88 deg. C with the same heat load and external boundary conditions. This difference in maximum temperature does not greatly affect the structural design of the shield, and demonstrates that it may be possible to use water for a lunar reactor shield.

Pearson, J. Boise; Stewart, Eric T. [NASA Marshall Space Flight Center, Huntsville, AL 35812 (United States); Reid, Robert S. [Los Alamos National Laboratory, Los Alamos, NM 87544 (United States)

2007-01-30T23:59:59.000Z

30

Standard Guide for Dry Lead Glass and Oil-Filled Lead Glass Radiation Shielding Window Components for Remotely Operated Facilities  

E-Print Network (OSTI)

Standard Guide for Dry Lead Glass and Oil-Filled Lead Glass Radiation Shielding Window Components for Remotely Operated Facilities

American Society for Testing and Materials. Philadelphia

2010-01-01T23:59:59.000Z

31

Magnetic shielding system for the tokamak experimental power reactor  

SciTech Connect

A poloidal field system is proposed for superconducting tokamaks which magnetically shields the toroidal field coil windings from pulsed fields and reduces the plasma inductive requirements. The principal elements of the system are a primary air core winding and a shield-vertical-field winding. The properties of the system are investigated using computer codes for MHD equilibrium, inductances, and fields. (auth)

Marcus, F.B.; Peng, Y.K.M.; Dory, R.A.; Moore, J.R.

1975-01-01T23:59:59.000Z

32

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

Science Conference Proceedings (OSTI)

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

2010-06-01T23:59:59.000Z

33

Mechanical cutting of irradiated reactor internal components  

Science Conference Proceedings (OSTI)

Mechanical cutting methods to volume reduce and package reactor internal components are now a viable solution for stakeholders challenged with the retirement of first generation nuclear facilities. The recent completion of the removal of the Reactor Vessel Internals (RVI) from within the Sacramento Municipal Utility District's (SMUD) Rancho Seco Nuclear Power Plant demonstrates that unlike previous methods, inclusive of plasma arc and abrasive water-jet cutting, mechanical cutting minimizes exposure to workers, costly water cleanup, and excessive secondary waste generation. Reactor internal components were segmented, packaged, and removed from the reactor building for shipment or storage, allowing the reactor cavity to be drained and follow-on reactor segmentation activities to proceed in the dry state. Area exposure rates at the work positions during the segmentation process were generally 1 mR per hr. Radiological exposure documented for the underwater segmentation processes totaled 13 person rem. The reactor internals weighing 343,000 pounds were segmented into over 200 pieces for maximum shipping package efficiency and produced 5,600 lb of stainless steel chips and shavings which were packaged in void spaces of existing disposal containers, therefore creating no additional disposal volume. Because no secondary waste was driven into suspension in the reactor cavity water, the water was free released after one pass through a charcoal bed and ion exchange filter system. Mechanical cutting techniques are capable of underwater segmentation of highly radioactive components on a large scale. This method minimized radiological exposure and costly water cleanup while creating no secondary waste.

Anderson, Michael G. [MOTA Corporation: 3410 Sunset Boulevard, West Columbia, SC, 29169 (United States)

2008-01-15T23:59:59.000Z

34

Thermal-Structural Design of a Water Shield For Surface Reactor Missions  

SciTech Connect

Water shielding is an attractive option for an affordable lunar surface fission reactor program. The attractiveness of the water shielding option arises from the relative ease of proto-typing and ground testing, the relatively low development effort needed, as well as the fabrication and operating experience with stainless steel and water. The most significant limitation in using a water shield is temperature: to prevent the formation of voids and the consequent loss of cooling, the water temperature has to be maintained below the saturation temperature corresponding to the shield pressure. This paper examines natural convection for a prototypic water shield design using the computational fluid dynamics (CFD) code CFX-5 as well as analytical modeling. The results show that natural convection is adequate to keep the water well-mixed. The results also show that for the above-ground configuration, shield surface and water temperatures during lunar day conditions are high enough to require shield pressures up to 2.5 atm to prevent void formation. For the buried configuration, a set of ammonia heat pipes attached to the shield outer wall can be used to maintain water temperatures within acceptable limits. Overall the results show that water shielding is feasible for lunar surface applications. The results of the CFD analyses can also be used to guide development of testing plans for shield thermal testing. (authors)

Sadasivan, Pratap; Kapernick, Richard J.; Poston, David I. [D-5 Nuclear Systems Design Group MS K575, Los Alamos National Laboratory, Los Alamos, New Mexico, 87545 (United States)

2006-07-01T23:59:59.000Z

35

Options for Shielding Tokamak Cooling Water Electrical Components against High Magnetic Fields  

SciTech Connect

The Tokamak Cooling Water System (TCWS) Instrumentation and Control (I&C) components of ITER will be located in areas of relatively high magnetic fields. Previous tests on electrical and I&C components have indicated that shielding will be required to protect these components from such magnetic fields. To accomplish this, studies were performed by AREVA Federal Services (AFS) in support of the TCWS Design project with the intent of identifying an optimal solution for shielding I&C components. This report presents a summary of these studies and presents design options for providing magnetic shielding to ITER TCWS I&C components and electrical equipment that are susceptible to the magnetic fields present.

Korsah, Kofi [ORNL; Michael, Smith [AREVA Federal Services LLC; Kim, Seokho H [ORNL; Charles, Neumeyer [Princeton Plasma Physics Laboratory (PPPL)

2011-01-01T23:59:59.000Z

36

Mechanical Cutting of Irradiated Reactor Internal Components  

SciTech Connect

This paper discusses the use of mechanical cutting methods to volume reduce and package irradiated reactor internal components. The recent completion of the removal of the Reactor Vessel Internals (RVI) from within the Sacramento Municipal Utility District's (SMUD) Rancho Seco Nuclear Power Plant demonstrates that unlike previous methods used for similar projects, mechanical cutting minimizes exposure to workers, costly water cleanup, and excessive secondary waste generation. (authors)

Anderson, M.G.; Fennema, J.A. [MOTA Corporation, West Columbia, SC (United States)

2007-07-01T23:59:59.000Z

37

Thermal insulating barrier and neutron shield providing integrated protection for a nuclear reactor vessel  

DOE Patents (OSTI)

The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel to form a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor.

Schreiber, Roger B. (Penn Twp., PA); Fero, Arnold H. (New Kensington, PA); Sejvar, James (Murrysville, PA)

1997-01-01T23:59:59.000Z

38

Thermal insulating barrier and neutron shield providing integrated protection for a nuclear reactor vessel  

DOE Patents (OSTI)

The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel to form a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive valving also includes bistable vents at the upper end of the thermal insulating barrier for releasing steam. A removable, modular neutron shield extending around the upper end of the reactor cavity below the nozzles forms with the upwardly and outwardly tapered transition on the outer surface of the reactor vessel, a labyrinthine channel which reduces neutron streaming while providing a passage for the escape of steam during a severe accident, and for the cooling air which is circulated along the reactor cavity walls outside the thermal insulating barrier during normal operation of the reactor. 8 figs.

Schreiber, R.B.; Fero, A.H.; Sejvar, J.

1997-12-16T23:59:59.000Z

39

Heavy Water Components Test Reactor Decommissioning - Major Component Removal  

SciTech Connect

The Heavy Water Components Test Reactor (HWCTR) facility (Figure 1) 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 facility is on high, well-drained ground, about 30 meters above the water table. The HWCTR was a pressurized heavy water test reactor used to develop candidate fuel designs for heavy water power reactors. It was not a defense-related facility like the materials production reactors at SRS. The reactor was moderated with heavy water and was rated at 50 megawatts thermal power. 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 1965, fuel assemblies were removed, systems that contained heavy water were drained, fluid piping systems were drained, deenergized and disconnected and the spent fuel basin was drained and dried. The doors of the reactor facility were shut and it wasn't until 10 years later that decommissioning plans were considered and ultimately postponed due to budget constraints. In the early 1990s, DOE began planning to decommission HWCTR again. 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. The $1.6 billion allocation from the American Recovery and Reinvestment Act to SRS for site clean up at SRS has opened the doors to the HWCTR again - this time for final decommissioning. During the lifetime of HWCTR, 36 different fuel assemblies were tested in the facility. Ten of these experienced cladding failures as operational capabilities of the different designs were being established. In addition, numerous spills of heavy water occurred within the facility. Currently, radiation and radioactive contamination levels are low within HWCTR with most of the radioactivity contained within the reactor vessel. There are no known insults to the environment, however with the increasing deterioration of the facility, the possibility exists that contamination could spread outside the facility if it is not decommissioned. An interior panoramic view of the ground floor elevation taken in August 2009 is shown in Figure 2. The foreground shows the transfer coffin followed by the reactor vessel and control rod drive platform in the center. Behind the reactor vessel is the fuel pool. Above the ground level are the polar crane and the emergency deluge tank at the top of the dome. Note the considerable rust and degradation of the components and the interior of the containment building. Alternative studies have concluded that the most environmentally safe, cost effective option for final decommissioning is to remove the reactor vessel, steam generators, and all equipment above grade including the dome. Characterization studies along with transport models have concluded that the remaining below grade equipment that is left in place including the transfer coffin will not contribute any significant contamination to the environment in the future. The below grade space will be grouted in place. A concrete cover will be placed over the remaining footprint and the groundwater will be monitored for an indefinite period to ensure compliance with environmental regulations. The schedule for completion of decommissioning is late FY2011. This paper describes the concepts planned in order to remove the major components including the dome, the reactor vessel (RV), the two steam generators (SG), and relocating the transfer coffin (TC).

Austin, W.; Brinkley, D.

2010-05-05T23:59:59.000Z

40

Heavy Water Components Test Reactor Decommissioning - Major Component Removal  

SciTech Connect

The Heavy Water Components Test Reactor (HWCTR) facility (Figure 1) 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 facility is on high, well-drained ground, about 30 meters above the water table. The HWCTR was a pressurized heavy water test reactor used to develop candidate fuel designs for heavy water power reactors. It was not a defense-related facility like the materials production reactors at SRS. The reactor was moderated with heavy water and was rated at 50 megawatts thermal power. 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 1965, fuel assemblies were removed, systems that contained heavy water were drained, fluid piping systems were drained, deenergized and disconnected and the spent fuel basin was drained and dried. The doors of the reactor facility were shut and it wasn't until 10 years later that decommissioning plans were considered and ultimately postponed due to budget constraints. In the early 1990s, DOE began planning to decommission HWCTR again. 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. The $1.6 billion allocation from the American Recovery and Reinvestment Act to SRS for site clean up at SRS has opened the doors to the HWCTR again - this time for final decommissioning. During the lifetime of HWCTR, 36 different fuel assemblies were tested in the facility. Ten of these experienced cladding failures as operational capabilities of the different designs were being established. In addition, numerous spills of heavy water occurred within the facility. Currently, radiation and radioactive contamination levels are low within HWCTR with most of the radioactivity contained within the reactor vessel. There are no known insults to the environment, however with the increasing deterioration of the facility, the possibility exists that contamination could spread outside the facility if it is not decommissioned. An interior panoramic view of the ground floor elevation taken in August 2009 is shown in Figure 2. The foreground shows the transfer coffin followed by the reactor vessel and control rod drive platform in the center. Behind the reactor vessel is the fuel pool. Above the ground level are the polar crane and the emergency deluge tank at the top of the dome. Note the considerable rust and degradation of the components and the interior of the containment building. Alternative studies have concluded that the most environmentally safe, cost effective option for final decommissioning is to remove the reactor vessel, steam generators, and all equipment above grade including the dome. Characterization studies along with transport models have concluded that the remaining below grade equipment that is left in place including the transfer coffin will not contribute any significant contamination to the environment in the future. The below grade space will be grouted in place. A concrete cover will be placed over the remaining footprint and the groundwater will be monitored for an indefinite period to ensure compliance with environmental regulations. The schedule for completion of decommissioning is late FY2011. This paper describes the concepts planned in order to remove the major components including the dome, the reactor vessel (RV), the two steam generators (SG), and relocating the transfer coffin (TC).

Austin, W.; Brinkley, D.

2010-05-05T23:59:59.000Z

Note: This page contains sample records for the topic "reactor components shield" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


41

HEAVY WATER COMPONENTS TEST REACTOR DECOMMISSIONING  

Science Conference Proceedings (OSTI)

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.

2011-10-13T23:59:59.000Z

42

Seismic and cask drop excitation evaluation of the tower shielding reactor  

SciTech Connect

During the current shutdown of the Tower Shielding Reactor II (TSR-II), analyses were performed to determine the effect of nearby cask drops on the structural and mechanical integrity of the reactor. This evaluation was then extended to include the effects of earthquakes. Several analytic models were developed to simulate the effects of earthquake and cask drop excitation. A coupled soil-structure model was developed. As a result of the analyses, several hardware modifications and enhancements were implemented to ensure reactor integrity during future operations. 6 figs.

Harris, S.P.; Stover, R.L.; Johnson, J.J.; Sumodobila, B.N. (EQE, Inc., San Francisco, CA (USA); Oak Ridge National Lab., TN (USA); EQE, Inc., San Francisco, CA (USA))

1989-01-01T23:59:59.000Z

43

Heat insulating system for a fast reactor shield slab  

DOE Patents (OSTI)

Improved thermal insulation for a nuclear reactor deck comprising many helical coil springs disposed in generally parallel, side-by-side laterally overlapping or interfitted relationship to one another so as to define a three-dimensional composite having both metal and voids between the metal, and enclosure means for holding the composite to the underside of the deck.

Kotora, Jr., James (LaGrange Park, IL); Groh, Edward F. (Naperville, IL); Kann, William J. (Park Ridge, IL); Burelbach, James P. (Glen Ellyn, IL)

1986-01-01T23:59:59.000Z

44

Heat insulating system for a fast reactor shield slab  

DOE Patents (OSTI)

Improved thermal insulation for a nuclear reactor deck comprises many helical coil springs disposed in generally parallel, side-by-side laterally overlapping or interfitted relationship to one another so as to define a three-dimensional composite having both metal and voids between the metal, and enclosure means for holding the composite to the underside of the deck.

Kotora, J. Jr.; Groh, E.F.; Kann, W.J.; Burelbach, J.P.

1984-04-10T23:59:59.000Z

45

GCFR shielding design and supporting experimental programs  

SciTech Connect

The shielding for the conceptual design of the gas-cooled fast breeder reactor (GCFR) is described, and the component exposure design criteria which determine the shield design are presented. The experimental programs for validating the GCFR shielding design methods and data (which have been in existence since 1976) are also discussed.

Perkins, R.G.; Hamilton, C.J.; Bartine, D.

1980-05-01T23:59:59.000Z

46

Characterization of Nuclear Reactor Materials and Components with ...  

Science Conference Proceedings (OSTI)

About this Symposium. Meeting, 2013 TMS Annual Meeting & Exhibition. Symposium, Characterization of Nuclear Reactor Materials and Components with  ...

47

Characterization of Nuclear Reactor Materials and Components with ...  

Science Conference Proceedings (OSTI)

About this Symposium. Meeting, 2011 TMS Annual Meeting & Exhibition. Symposium, Characterization of Nuclear Reactor Materials and Components with  ...

48

Nuclear reactor removable radial shielding assembly having a self-bowing feature  

DOE Patents (OSTI)

A removable radial shielding assembly for use in the periphery of the core of a liquid-metal-cooled fast-breeder reactor, for closing interassembly gaps in the reactor core assembly load plane prior to reactor criticality and power operation to prevent positive reactivity insertion. The assembly has a lower nozzle portion for inserting into the core support and a flexible heat-sensitive bimetallic central spine surrounded by blocks of shielding material. At refueling temperature and below the spine is relaxed and in a vertical position so that the tolerances permitted by the interassembly gaps allow removal and replacement of the various reactor core assemblies. During an increase in reactor temperature from refueling to hot standby, the bimetallic spine expands, bowing the assembly toward the core center line, exerting a radially inward gap-closing-force on the above core load plane of the reactor core assembly, closing load plane interassembly gaps throughout the core prior to startup and preventing positive reactivity insertion.

Pennell, William E. (Greensburg, PA); Kalinowski, Joseph E. (Smithton, PA); Waldby, Robert N. (New Stanton, PA); Rylatt, John A. (Monroeville, PA); Swenson, Daniel V. (Greensburg, PA)

1978-01-01T23:59:59.000Z

49

REACTOR AND SHIELD PHYSICS. Comprehensive Technical Report, General Electric Direct-Air-Cycle, Aircraft Nuclear Propulsion Program.  

SciTech Connect

This volume is one of twenty-one summarizing the Aircraft Nuclear Propulsion Program of the General Electric Company. This volume describes the experimental and theoretical work accomplished in the areas of reactor and shield physics.

Edwards, W.E.; Simpson, J.D.

1962-01-01T23:59:59.000Z

50

Characterization of Nuclear Reactor Materials and Components with ...  

Science Conference Proceedings (OSTI)

Mar 6, 2013 ... Characterization of Nuclear Reactor Materials and Components with ... Results are discussed in terms of existing theoretical models for hydride ...

51

Austenitic alloy and reactor components made thereof  

DOE Patents (OSTI)

An austenitic stainless steel alloy is disclosed, having excellent fast neutron irradiation swelling resistance and good post irradiation ductility, making it especially useful for liquid metal fast breeder reactor applications. The alloy contains: about 0.04 to 0.09 wt. % carbon; about 1.5 to 2.5 wt. % manganese; about 0.5 to 1.6 wt. % silicon; about 0.030 to 0.08 wt. % phosphorus; about 13.3 to 16.5 wt. % chromium; about 13.7 to 16.0 wt. % nickel; about 1.0 to 3.0 wt. % molybdenum; and about 0.10 to 0.35 wt. % titanium.

Bates, John F. (Ogden, UT); Brager, Howard R. (Richland, WA); Korenko, Michael K. (Wexford, PA)

1986-01-01T23:59:59.000Z

52

ACDOS1: A COMPUTER CODE TO CALCULATE DOSE RATES FROM NEUTRON ACTIVATION OF NEUTRAL BEAMLINES AND OTHER FUSION REACTOR COMPONENTS  

E-Print Network (OSTI)

BEAMLINES AND OTHER FUSION--REACTOR COMPONENTS Gregory S.Beamlines and Other Fusion-Reactor Compon­ ents By Gregoryin the Tokamak Fusion Test Reactor Test Cell", Nucl.

Keney, G.S.

2010-01-01T23:59:59.000Z

53

Hydraulic balancing of a control component within a nuclear reactor  

DOE Patents (OSTI)

A reactor control component includes an inner conduit, for instance containing neutron absorber elements, adapted for longitudinal movement within an outer guide duct. A transverse partition partially encloses one end of the conduit and meets a transverse wall within the guide duct when the conduit is fully inserted into the reactor core. A tube piece extends from the transverse partition and is coaxially aligned to be received within a tubular receptacle which extends from the transverse wall. The tube piece and receptacle cooperate in engagement to restrict the flow and pressure of coolant beneath the transverse partition and thereby minimize upward forces tending to expel the inner conduit.

Marinos, D.; Ripfel, H.C.F.

1975-10-14T23:59:59.000Z

54

Nuclear reactor spacer grid and ductless core component  

DOE Patents (OSTI)

The invention relates to a nuclear reactor spacer grid member for use in a liquid cooled nuclear reactor and to a ductless core component employing a plurality of these spacer grid members. The spacer grid member is of the egg-shell type and is constructed so that the walls of the cell members of the grid member are formed of a single thickness of metal to avoid tolerance problems. Within each cell member is a hydraulic spring which laterally constrains the nuclear material bearing rod which passes through each cell member against a hardstop in response to coolant flow through the cell member. This hydraulic spring is also suitable for use in a water cooled nuclear reactor. A core component constructed of, among other components, a plurality of these spacer grid members, avoids the use of a full length duct by providing spacer sleeves about the sodium tubes passing through the spacer grid members at locations between the grid members, thereby maintaining a predetermined space between adjacent grid members.

Christiansen, David W. (Kennewick, WA); Karnesky, Richard A. (Richland, WA)

1989-01-01T23:59:59.000Z

55

Prognostics Health Management for Advanced Small Modular Reactor Passive Components  

SciTech Connect

In the United States, sustainable nuclear power to promote energy security is a key national energy priority. Advanced small modular reactors (AdvSMR), which are based on modularization of advanced reactor concepts using non-light-water reactor (LWR) coolants such as liquid metal, helium, or liquid salt may provide a longer-term alternative to more conventional LWR-based concepts. The economics of AdvSMRs will be impacted by the reduced economy-of-scale savings when compared to traditional LWRs and the controllable day-to-day costs of AdvSMRs are expected to be dominated by operations and maintenance costs. Therefore, achieving the full benefits of AdvSMR deployment requires a new paradigm for plant design and management. In this context, prognostic health management of passive components in AdvSMRs can play a key role in enabling the economic deployment of AdvSMRs. In this paper, the background of AdvSMRs is discussed from which requirements for PHM systems are derived. The particle filter technique is proposed as a prognostics framework for AdvSMR passive components and the suitability of the particle filter technique is illustrated by using it to forecast thermal creep degradation using a physics-of-failure model and based on a combination of types of measurements conceived for passive AdvSMR components.

Meyer, Ryan M.; Ramuhalli, Pradeep; Coble, Jamie B.; Mitchell, Mark R.; Wootan, David W.; Hirt, Evelyn H.; Berglin, Eric J.; Bond, Leonard J.; Henager, Charles H.

2013-10-18T23:59:59.000Z

56

Evaluation of an Advanced Radiation Shielding Material for Permanent Installation at an Operating Commercial Nuclear Reactor  

Science Conference Proceedings (OSTI)

The industry continues to investigate, validate, and implement new radiation field reduction measures in response to increased emphasis on reducing dose to workers. Many nuclear plants are interested in permanent shielding applications to further reduce personnel exposure and to reduce the recurring effort associated with temporary installations. In 2008, a flexible, impregnated, layered matrix material was identified as a possible material for incorporating a shielding substance. This report provides an...

2010-09-30T23:59:59.000Z

57

Nuclear reactor heat transport system component low friction support system  

SciTech Connect

A support column for a heavy component of a liquid metal fast breeder reactor heat transport system which will deflect when the pipes leading coolant to and from the heavy component expand or contract due to temperature changes includes a vertically disposed pipe, the pipe being connected to the heavy component by two longitudinally spaced cycloidal dovetail joints wherein the distal end of each of the dovetails constitutes a part of the surface of a large diameter cylinder and the centerlines of these large diameter cylinders intersect at right angles and the pipe being supported through two longitudinally spaced cycloidal dovetail joints wherein the distal end of each of the dovetails constitutes a part of the surface of a large diameter cylinder and the centerlines of these large diameter cylinders intersect at right angles, each of the cylindrical surfaces bearing on a flat and horizontal surface.

Wade, Elman E. (Ruffs Dale, PA)

1980-01-01T23:59:59.000Z

58

Nonthermal effects in two component DT fusion reactors  

SciTech Connect

Net energy generation rates and f-factors are calculated for a variety of two component DT reactor configurations using a computer code that follows the energy distributions of the reactants and products explicitly, utilizing the Fokker--Planck approximation for low-angle Coulomb scattering and a transfer matrix for high-angle Coulomb, nuclear, and radiative processes. The relative importance of such non-thermal effects as alpha particle deposition, non- Maxwellian energy distributions for the target tritons and electrons, and the influence of high-angle Coulomb and nuclear scattering on the energy loss rate of the injected deuterons is explicitly assessed. (auth)

Weaver, T.A.; Chu, T.C.

1975-11-01T23:59:59.000Z

59

Method of detecting leakage of reactor core components of liquid metal cooled fast reactors  

DOE Patents (OSTI)

A method of detecting the failure of a sealed non-fueled core component of a liquid-metal cooled fast reactor having an inert cover gas. A gas mixture is incorporated in the component which includes Xenon-124; under neutron irradiation, Xenon-124 is converted to radioactive Xenon-125. The cover gas is scanned by a radiation detector. The occurrence of 188 Kev gamma radiation and/or other identifying gamma radiation-energy level indicates the presence of Xenon-125 and therefore leakage of a component. Similarly, Xe-126, which transmutes to Xe-127 and Kr-84, which produces Kr-85.sup.m can be used for detection of leakage. Different components are charged with mixtures including different ratios of isotopes other than Xenon-124. On detection of the identifying radiation, the cover gas is subjected to mass spectroscopic analysis to locate the leaking component.

Holt, Fred E. (Richland, WA); Cash, Robert J. (Richland, WA); Schenter, Robert E. (Richland, WA)

1977-01-01T23:59:59.000Z

60

Radiation physics and shielding codes and analyses applied to design-assist and safety analyses of CANDU{sup R} and ACR{sup TM} reactors  

SciTech Connect

This paper discusses the radiation physics and shielding codes and analyses applied in the design of CANDU and ACR reactors. The focus is on the types of analyses undertaken rather than the inputs supplied to the engineering disciplines. Nevertheless, the discussion does show how these analyses contribute to the engineering design. Analyses in radiation physics and shielding can be categorized as either design-assist or safety and licensing (accident) analyses. Many of the analyses undertaken are designated 'design-assist' where the analyses are used to generate recommendations that directly influence plant design. These recommendations are directed at mitigating or reducing the radiation hazard of the nuclear power plant with engineered systems and components. Thus the analyses serve a primary safety function by ensuring the plant can be operated with acceptable radiation hazards to the workers and public. In addition to this role of design assist, radiation physics and shielding codes are also deployed in safety and licensing assessments of the consequences of radioactive releases of gaseous and liquid effluents during normal operation and gaseous effluents following accidents. In the latter category, the final consequences of accident sequences, expressed in terms of radiation dose to members of the public, and inputs to accident analysis, e.g., decay heat in fuel following a loss-of-coolant accident, are also calculated. Another role of the analyses is to demonstrate that the design of the plant satisfies the principle of ALARA (as low as reasonably achievable) radiation doses. This principle is applied throughout the design process to minimize worker and public doses. The principle of ALARA is an inherent part of all design-assist recommendations and safety and licensing assessments. The main focus of an ALARA exercise at the design stage is to minimize the radiation hazards at the source. This exploits material selection and impurity specifications and relies heavily on experience and engineering judgement, consistent with the ALARA philosophy. Special care is taken to ensure that the best estimate dose rates are used to the extent possible when applying ALARA. Provisions for safeguards equipment are made throughout the fuel-handling route in CANDU and ACR reactors. For example, the fuel bundle counters rely on the decay gammas from the fission products in spent-fuel bundles to record the number of fuel movements. The International Atomic Energy Agency (IAEA) Safeguards system for CANDU and ACR reactors is based on item (fuel bundle) accounting. It involves a combination of IAEA inspection with containment and surveillance, and continuous unattended monitoring. The spent fuel bundle counter monitors spent fuel bundles as they are transferred from the fuelling machine to the spent fuel bay. The shielding and dose-rate analysis need to be carried out so that the bundle counter functions properly. This paper includes two codes used in criticality safety analyses. Criticality safety is a unique phenomenon and codes that address criticality issues will demand specific validations. However, it is recognised that some of the codes used in radiation physics will also be used in criticality safety assessments. (authors)

Aydogdu, K.; Boss, C. R. [Atomic Energy of Canada Limited, Sheridan Science and Technology Park, Mississauga, Ont. L5K 1B2 (Canada)

2006-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor components shield" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


61

ACDOS1: A COMPUTER CODE TO CALCULATE DOSE RATES FROM NEUTRON ACTIVATION OF NEUTRAL BEAMLINES AND OTHER FUSION REACTOR COMPONENTS  

E-Print Network (OSTI)

the Tokamak Fusion Test Reactor Test Cell", Nucl. Technol.BEAMLINES AND OTHER FUSION--REACTOR COMPONENTS Gregory S.Beamlines and Other Fusion-Reactor Compon­ ents By Gregory

Keney, G.S.

2010-01-01T23:59:59.000Z

62

Performance of Ni-alloy Components in CANDU Reactors  

Science Conference Proceedings (OSTI)

Environmentally Assisted Cracking Susceptibility Assessment of AP 1000 Reactor Coolant Pump Flywheel Retainer Ring A289 18Cr-18Mn Steel by Slow Strain ...

63

Development of Underwater Laser Cladding and Underwater Laser Seal Welding Techniques for Reactor Components (II)  

SciTech Connect

Stress corrosion cracking (SCC) is one of the major reasons to reduce the reliability of aged reactor components. Toshiba has been developing underwater laser welding onto surface of the aged components as maintenance and repair techniques. Because most of the reactor internal components to apply this underwater laser welding technique have 3-dimensional shape, effect of welding positions and welded shapes are examined and presented in this report. (authors)

Masataka Tamura; Shohei Kawano; Wataru Kouno; Yasushi Kanazawa [Toshiba Corporation (Japan)

2006-07-01T23:59:59.000Z

64

Radiation Shielding Design and Orientation Considerations for a 1 kWe Heat Pipe Cooled Reactor Utilized to Bore Through the Ice Caps of Mars  

SciTech Connect

The goal in designing any space power system is to develop a system able to meet the mission requirements for success while minimizing the overall costs. The mission requirements for the this study was to develop a reactor (with Stirling engine power conversion) and shielding configuration able to fit, along with all the other necessary science equipment, in a Cryobot 3 m high with {approx}0.5 m diameter hull, produce 1 kWe for 5yrs, and not adversely affect the mission science by keeping the total integrated dose to the science equipment below 150 krad. Since in most space power missions the overall system mass dictates the mission cost, the shielding designs in this study incorporated Martian water extracted at the startup site in order to minimize the tungsten and LiH mass loading at launch. Different reliability and mass minimization concerns led to three design configuration evolutions. With the help of implementing Martian water and configuring the reactor as far from the science equipment as possible, the needed tungsten and LiH shield mass was minimized. This study further characterizes the startup dose and the necessary mission requirements in order to ensure integrity of the surface equipment during reactor startup phase.

Fensin, Michael L. [Department of Nuclear and Radiological Engineering, University of Florida, Gainesville, FL 32611 (United States); Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Elliott, John O. [Jet Propulsion Laboratories, California Institute of Technology, Pasedena, Ca 91109 (United States); Lipinski, Ronald J. [Sandia National Laboratory, Albuquerque, NM 87185 (United States); Poston, David I. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

2006-01-20T23:59:59.000Z

65

Experience with non-fuel-bearing components in LWR (light-water reactor) fuel systems  

SciTech Connect

Many non-fuel-bearing components are so closely associated with the spent fuel assemblies that their integrity and behavior must be taken into consideration with the fuel assemblies, when handling spent fuel of planning waste management activities. Presented herein is some of the experience that has been gained over the past two decades from non-fuel-bearing components in light-water reactors (LWRs), both pressurized-water reactors (PWRs) and boiling-water reactors (BWRs). Among the most important of these components are the control rod systems, the absorber and burnable poison rods, and the fuel assembly channels. 15 refs., 5 figs., 2 tabs.

Bailey, W.J.; Berting, F.M.

1990-12-01T23:59:59.000Z

66

Overview of Component Testing Requirements for a Small Fluoride Salt-Cooled High Tempreature Reactor  

Science Conference Proceedings (OSTI)

This article summarizes the information necessary to provide reasonable assurance that components for a small fluoride salt-cooled high temperature reactor will meet their functional requirements. In support of the analysis of testing requirements, a simplified, conceptual description of the systems, structures, and components specific to this reactor class was developed. These reactor system elements were divided into major categories based on their functions: (1) reactor core systems, (2) heat transport system, (3) reactor auxiliary cooling system, and (4) instrumentation and controls system. An assessment of technical maturity for each element was made, and a gap analysis was performed to identify specific areas that require further testing. A prioritized list of the testing requirements was then developed. The prioritization was based on both the relative importance of the system to reactor viability, and performance and time requirements to perform the testing.

Cetiner, Mustafa Sacit [ORNL; Holcomb, David Eugene [ORNL; Flanagan, George F [ORNL; Peretz, Fred J [ORNL; Yoder Jr, Graydon L [ORNL

2010-01-01T23:59:59.000Z

67

Surface erosion of fusion reactor components due to radiation blistering and neutron sputtering  

SciTech Connect

Radiation blistering and neutron sputtering can lead to the surface erosion of fusion reactor components exposed to plasma radiations. Recent studies of methods to reduce the surface erosion caused by these processes are discussed. (auth)

Das, S.K.; Kaminsky, M.

1975-01-01T23:59:59.000Z

68

Modelling of a passive reactor cavity cooling system (RCCS) for a nuclear reactor core subject to environmental changes and the optimisation of the RCCS radiation heat shield heat shield.  

E-Print Network (OSTI)

??ENGLISH ABSTRACT: A reactor cavity cooling system (RCCS) is used in the PBMR to protect the concrete citadel surrounding the reactor from direct nuclear radiation… (more)

Verwey, Aldo

2010-01-01T23:59:59.000Z

69

An Analysis of Testing Requirements for Fluoride Salt Cooled High Temperature Reactor Components  

SciTech Connect

This report provides guidance on the component testing necessary during the next phase of fluoride salt-cooled high temperature reactor (FHR) development. In particular, the report identifies and describes the reactor component performance and reliability requirements, provides an overview of what information is necessary to provide assurance that components will adequately achieve the requirements, and then provides guidance on how the required performance information can efficiently be obtained. The report includes a system description of a representative test scale FHR reactor. The reactor parameters presented in this report should only be considered as placeholder values until an FHR test scale reactor design is completed. The report focus is bounded at the interface between and the reactor primary coolant salt and the fuel and the gas supply and return to the Brayton cycle power conversion system. The analysis is limited to component level testing and does not address system level testing issues. Further, the report is oriented as a bottom-up testing requirements analysis as opposed to a having a top-down facility description focus.

Holcomb, David Eugene [ORNL; Cetiner, Mustafa Sacit [ORNL; Flanagan, George F [ORNL; Peretz, Fred J [ORNL; Yoder Jr, Graydon L [ORNL

2009-11-01T23:59:59.000Z

70

Advanced shield development for a fission surface power system for the lunar surface  

SciTech Connect

A nuclear reactor power system such as the affordable fission surface power system enables a potential outpostonthemoon.Aradiation shieldmustbe included in the reactor system to reduce the otherwise excessive dose to the astronauts and other vital system components. The radiation shield is typically the most massive component of a space reactor system, and thus must be optimized to reduce mass asmuchas possible while still providing the required protection.Various shield options for an on-lander reactor system are examined for outpost distances of 400m and 1 kmfromthe reactor. Also investigated is the resulting mass savings from the use of a high performance cermet fuel. A thermal analysis is performed to determine the thermal behaviours of radiation shields using borated water. For an outpost located 1000m from the core, a tetramethylammonium borohydride shield is the lightest (5148.4 kg), followed by a trilayer shield (boron carbide–tungsten–borated water; 5832.3 kg), and finally a borated water shield (6020.7 kg). In all of the final design cases, the temperature of the borated water remains below 400 K.

A. E. Craft; I. J. Silver; C. M. Clark; S. D. Howe; J. C. King

2011-02-01T23:59:59.000Z

71

Neutronic Assessment of Candidate Materials for TF Coils Shielding in a DEMO Fusion Reactor Based on a DCLL Blanket  

Science Conference Proceedings (OSTI)

Blanket Materials Technology / Proceedings of the Fifteenth International Conference on Fusion Reactor Materials, Part A: Fusion Technology

J. P. Catalán; J. Sanz; F. Ogando; R. Pampin

72

Process, Techniques, and Successes in Welding the Dry Shielded Canister Welds of the TMI-2 Reactor Core Debris  

SciTech Connect

The Idaho National Engineering and Environmental Laboratory (INEEL) is operated by Bechtel-BWXT Idaho LLC (BBWI), which recently completed a very successful $100 million Three-Mile Island-2 (TMI-2) program for the Department of Energy (DOE). This complex and challenging program used an integrated multidisciplinary team approach that loaded, welded, and transported an unprecedented 25 dry shielded canisters (DSC) in seven months, and did so ahead of schedule. The program moved over 340 canisters of TMI-2 core debris that had been in wet storage into a dry storage facility at the INEEL. The main thrust of this paper is relating the innovations, techniques, approaches, and lessons learned associated to welding of the DSC's. This paper shows the synergism of elements to meet program success and shares these lessons learned that will facilitate success with welding of dry shielded canisters in other DOE complex dry storage programs.

Zirker, Laurence R; Rankin, Richard Allen; Ferrell, Larry Joseph

2002-02-01T23:59:59.000Z

73

Processes, Techniques, and Successes in Welding the Dry Shielded Canisters of the TMI-2 Reactor Core Debris  

SciTech Connect

The Idaho National Engineering and Environmental Laboratory (INEEL) is operated by Bechtel-BWXT Idaho LLC (BBWI), which recently completed a very successful $100 million Three-Mile Island-2 (TMI-2) program for the Department of Energy (DOE). This complex and challenging program used an integrated multidisciplinary team approach that loaded, welded, and transported an unprecedented 25 dry shielded canisters (DSC) in seven months, and did so ahead of schedule. The program moved over 340 canisters of TMI-2 core debris that had been in wet storage into a dry storage facility at the INEEL. The main thrust of this paper is relating the innovations, techniques, approaches, and lessons learned associated to welding of the DSC's. This paper shows the synergism of elements to meet program success and shares these lessons learned that will facilitate success with welding of dry shielded canisters in other DOE complex dry storage programs.

Zirker, L.R.; Rankin, R.A.; Ferrell, L.J.

2002-01-29T23:59:59.000Z

74

The Evaluation of Lithium Hydride for Use in a Space Nuclear Reactor Shield, Including a Historical Perspective  

DOE Green Energy (OSTI)

LiH was one of the five primary shield materials the NRPCT intended to develop (along with beryllium, boron carbide, tungsten, and water) for potential Prometheus application. It was also anticipated that {sup 10}B metal would be investigated for feasibility at a low level of effort. LiH historically has been selected as a low mass, neutron absorption material for space shields (Systems for Nuclear Auxiliary Power (SNAP), Topaz, SP-100). Initial NRPCT investigations did not produce convincing evidence that LiH was desirable or feasible for a Prometheus mission due to material property issues (primarily swelling and hydrogen cover gas containment), and related thermal design complexity. Furthermore, if mass limits allowed, an option to avoid use of LiH was being contemplated to lower development costs and associated risks. However, LiH remains theoretically the most efficient neutron shield material per unit mass, and, with sufficient testing and development, could be an optimal material choice for future flights.

D. Poeth

2005-12-09T23:59:59.000Z

75

TA-2 Water Boiler Reactor Decommissioning Project  

Science Conference Proceedings (OSTI)

This final report addresses the Phase 2 decommissioning of the Water Boiler Reactor, biological shield, other components within the biological shield, and piping pits in the floor of the reactor building. External structures and underground piping associated with the gaseous effluent (stack) line from Technical Area 2 (TA-2) Water Boiler Reactor were removed in 1985--1986 as Phase 1 of reactor decommissioning. The cost of Phase 2 was approximately $623K. The decommissioning operation produced 173 m{sup 3} of low-level solid radioactive waste and 35 m{sup 3} of mixed waste. 15 refs., 25 figs., 3 tabs.

Durbin, M.E. (ed.); Montoya, G.M.

1991-06-01T23:59:59.000Z

76

Analysis of seismic sloshing of reactor tanks considering submerged components and seismic isolation  

Science Conference Proceedings (OSTI)

A study of the seismic sloshing response of a large pool-type reactor tank with several deck-mounted components is presented. The main objective of the study is to investigate the effects of internal components on the sloshing response and to determine the sloshing loads on the components. The study shows that the presence of internal components can significantly change the dynamic characteristics of the sloshing motion. The sloshing frequencies of a tank with internal components are considerably higher than those of a tank without internals. The higher sloshing frequencies reduce the sloshing wave height on the free surface but the dynamic pressures of the fluid are increased. The effects of seismic isolation on sloshing response are also presented.

Ma, D.C.; Chang, Y.M.

1985-01-01T23:59:59.000Z

77

Materials Reliability Program: Third International Conference on Fatigue of Reactor Components (MRP-151)  

Science Conference Proceedings (OSTI)

These proceedings contain information presented at the Third International Conference on Fatigue of Reactor Components held October 3-6, 2004, in Seville, Spain. This third conference was again sponsored by EPRI, the Organisation for Economic Co-Operation and Development (OECD) Nuclear Energy Agency/Committee on the Safety of Nuclear Installations (NEA/CSNI), and the U. S. Nuclear Regulatory Commission (NRC) and hosted by Consejo de Segurida Nuclear (CNS) of Spain.

2005-09-12T23:59:59.000Z

78

Adhesive particle shielding  

DOE Patents (OSTI)

An efficient device for capturing fast moving particles has an adhesive particle shield that includes (i) a mounting panel and (ii) a film that is attached to the mounting panel wherein the outer surface of the film has an adhesive coating disposed thereon to capture particles contacting the outer surface. The shield can be employed to maintain a substantially particle free environment such as in photolithographic systems having critical surfaces, such as wafers, masks, and optics and in the tools used to make these components, that are sensitive to particle contamination. The shield can be portable to be positioned in hard-to-reach areas of a photolithography machine. The adhesive particle shield can incorporate cooling means to attract particles via the thermophoresis effect.

Klebanoff, Leonard Elliott (Dublin, CA); Rader, Daniel John (Albuquerque, NM); Walton, Christopher (Berkeley, CA); Folta, James (Livermore, CA)

2009-01-06T23:59:59.000Z

79

COMPONENT DEGRADATION SUSCEPTIBILITIES AS THE BASES FOR MODELING REACTOR AGING RISK  

Science Conference Proceedings (OSTI)

The extension of nuclear power plant operating licenses beyond 60 years in the United States will be necessary if we are to meet national energy needs while addressing the issues of carbon and climate. Characterizing the operating risks associated with aging reactors is problematic because the principal tool for risk-informed decision-making, Probabilistic Risk Assessment (PRA), is not ideally-suited to addressing aging systems. The components most likely to drive risk in an aging reactor - the passives - receive limited treatment in PRA, and furthermore, standard PRA methods are based on the assumption of stationary failure rates: a condition unlikely to be met in an aging system. A critical barrier to modeling passives aging on the wide scale required for a PRA is that there is seldom sufficient field data to populate parametric failure models, and nor is there the availability of practical physics models to predict out-year component reliability. The methodology described here circumvents some of these data and modeling needs by using materials degradation metrics, integrated with conventional PRA models, to produce risk importance measures for specific aging mechanisms and component types. We suggest that these measures have multiple applications, from the risk-screening of components to the prioritization of materials research.

Unwin, Stephen D.; Lowry, Peter P.; Toyooka, Michael Y.

2010-07-18T23:59:59.000Z

80

Modular shield  

DOE Patents (OSTI)

A modular system for containing projectiles has a sheet of material including at least a polycarbonate layer held by a metal frame having a straight frame member corresponding to each straight edge of the sheet. Each frame member has a U-shaped shield channel covering and holding a straight edge of the sheet and an adjacent U-shaped clamp channel rigidly held against the shield channel. A flexible gasket separates each sheet edge from its respective shield channel; and each frame member is fastened to each adjacent frame member only by clamps extending between adjacent clamp channels.

Snyder, Keith W. (Sandia Park, NM)

2002-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor components shield" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


81

Mechanical properties of thermally aged cast stainless steels from Shippingport reactor components  

SciTech Connect

Thermal embrittlement of static-cast CF-8 stainless steel components from the decommissioned Shippingport reactor has been characterized. Cast stainless steel materials were obtained from four cold-leg check valves, three hot-leg main shutoff valves, and two pump volutes. The actual time-at-temperature for the materials was {approximately}13 y at {approximately}281 C (538 F) for the hot-leg components and {approximately}264 C (507 F) for the cold-leg components. Baseline mechanical properties for as-cast material were determined from tests on either recovery-annealed material, i.e., annealed for 1 h at 550 C and then water quenched, or material from the cooler region of the component. The Shippingport materials show modest decreases in fracture toughness and Charpy-impact properties and a small increase in tensile strength because of relatively low service temperatures and ferrite content of the steel. The procedure and correlations developed at Argonne National Laboratory for estimating mechanical properties of cast stainless steels predict accurate or slightly lower values for Charpy-impact energy, tensile flow stress, fracture toughness J-R curve, and J{sub IC} of the materials. The kinetics of thermal embrittlement and degree of embrittlement at saturation, i.e., the minimum impact energy achieved after long-term aging, were established from materials that were aged further in the laboratory. The results were consistent with the estimates. The correlations successfully predicted the mechanical properties of the Ringhals 2 reactor hot and crossover-leg elbows (CF-8M steel) after service of {approximately} 15 y and the KRB reactor pump cover plate (CF-8) after {approximately} 8 y of service.

Chopra, O.K.; Shack, W.J. [Argonne National Lab., IL (United States)

1995-04-01T23:59:59.000Z

82

Radiation shielding composition  

DOE Patents (OSTI)

A composition for use as a radiation shield. The shield is a concrete product containing a stable uranium aggregate for attenuating gamma rays and a neutron absorbing component, the uranium aggregate and neutron absorbing component being present in the concrete product in sufficient amounts to provide a concrete having a density between about 4 and about 15 grams/cm.sup.3 and which will at a predetermined thickness, attenuate gamma rays and absorb neutrons from a radioactive material of projected gamma ray and neutron emissions over a determined time period. The composition is preferably in the form of a container for storing radioactive materials that emit gamma rays and neutrons. The concrete container preferably comprises a metal liner and/or a metal outer shell. The resulting radiation shielding container has the potential of being structurally sound, stable over a long period of time, and, if desired, readily mobile.

Quapp, William J. (Idaho Falls, ID); Lessing, Paul A. (Idaho Falls, ID)

1998-01-01T23:59:59.000Z

83

Radiation shielding composition  

DOE Patents (OSTI)

A composition for use as a radiation shield. The shield is a concrete product containing a stable uranium aggregate for attenuating gamma rays and a neutron absorbing component, the uranium aggregate and neutron absorbing component being present in the concrete product in sufficient amounts to provide a concrete having a density between about 4 and about 15 grams/cm.sup.3 and which will at a predetermined thickness, attenuate gamma rays and absorb neutrons from a radioactive material of projected gamma ray and neutron emissions over a determined time period. The composition is preferably in the form of a container for storing radioactive materials that emit gamma rays and neutrons. The concrete container preferably comprises a metal liner and/or a metal outer shell. The resulting radiation shielding container has the potential of being structurally sound, stable over a long period of time, and, if desired, readily mobile.

Quapp, William J. (Idaho Falls, ID); Lessing, Paul A. (Idaho Falls, ID)

2000-12-26T23:59:59.000Z

84

Radiation Shielding Options for the Affordable Fission Surface Power System  

SciTech Connect

The Affordable Fission Surface Power System (AFSPS) is a proposed power source for an outpost capable of housing six humans for up to six weeks on the lunar surface and emphasizes the design principles of low risk and affordability over high performance. The radiation shield is the most massive component of the reactor system and its effect on launch mass greatly affects the affordability of the AFSPS. Potential shielding materials include lithium hydride, enriched boron-10 carbide, water, borated water, beryllium, boron-doped beryllium and zirconium hydride. Zirconium hydride is the most effective neutron attenuator and also significantly attenuates gamma radiation, but at a significant mass penalty. The other neutron attenuating materials all require the addition of a tungsten layer to provide significant gamma attenuation. Based on neutron radiation alone, lithium hydride is the lightest of the potential attenuators, followed by water and borated water. When gamma radiation is also considered, the lithium hydride/tungsten shield is shown to be the lightest composite shield with a combined mass of 3246 kg, followed by the borated water/tungsten shield (3479 kg). The boron carbide/tungsten shield has a total mass of 4129 kg, but represents significantly less development risk.

Craft, Aaron E.; King, Jeffrey C. [Mining and Nuclear Engineering Department Missouri University of Science and Technology 222 Fulton Hall, 301 W. 14th Street Rolla, MO 65401(United States)

2009-03-16T23:59:59.000Z

85

Packaging, Transportation, and Disposal Logistics for Large Radioactively Contaminated Reactor Decommissioning Components  

Science Conference Proceedings (OSTI)

The packaging, transportation and disposal of large, retired reactor components from operating or decommissioning nuclear plants pose unique challenges from a technical as well as regulatory compliance standpoint. In addition to the routine considerations associated with any radioactive waste disposition activity, such as characterization, ALARA, and manifesting, the technical challenges for large radioactively contaminated components, such as access, segmentation, removal, packaging, rigging, lifting, mode of transportation, conveyance compatibility, and load securing require significant planning and execution. In addition, the current regulatory framework, domestically in Titles 49 and 10 and internationally in TS-R-1, does not lend itself to the transport of these large radioactively contaminated components, such as reactor vessels, steam generators, reactor pressure vessel heads, and pressurizers, without application for a special permit or arrangement. This paper addresses the methods of overcoming the technical and regulatory challenges. The challenges and disposition decisions do differ during decommissioning versus component replacement during an outage at an operating plant. During decommissioning, there is less concern about critical path for restart and more concern about volume reduction and waste minimization. Segmentation on-site is an available option during decommissioning, since labor and equipment will be readily available and decontamination activities are routine. The reactor building removal path is also of less concern and there are more rigging/lifting options available. Radionuclide assessment is necessary for transportation and disposal characterization. Characterization will dictate the packaging methodology, transportation mode, need for intermediate processing, and the disposal location or availability. Characterization will also assist in determining if the large component can be transported in full compliance with the transportation and disposal regulations and criteria or if special authorizations must be granted to transport and/or dispose. The U.S. DOT routinely issues special permits for large components where compliance with regulatory or acceptance criteria is impractical or impossible to meet. Transportation and disposal safety must be maintained even under special permits or authorizations. For example, if transported un-packaged, performance analysis must still be performed to assess the ability of the large component's outer steel shell to contain the internal radioactive contamination under normal transportation conditions and possibly incidence normal to transportation. The dimensions and weight of a large component must be considered when determining the possible modes of transportation (rail, water, or highway). At some locations, rail and/or barge access is unavailable. Many locations that once had an active rail spur to deliver new construction materials and components have let the spur deteriorate to the point that repair and upgrade of the spur is no longer economically feasible. Barge slips that have not been used since new plant construction require significant repair and/or dredging. Short on-site haul routes must be assessed for surface and subsurface conditions, as well as longer off-site routes. Off-site routes require clearance approvals from the regulatory authorities or, in the case of rail transport, the rail lines. Significant engineering planning and analysis must be performed during the pre-mobilization. In conclusion, the packaging, transportation, and disposal of large, oversized radioactively contaminated components removed during plant decommissioning is complex. However, over the last 15 years, a 100 or more components have been safely and compliantly packaged and transported for processing and/or disposal.

Lewis, Mark S. [EnergySolutions: 140 Stoneridge Drive, Columbia, SC 29210 (United States)

2008-01-15T23:59:59.000Z

86

Facility Configuration Study of the High Temperature Gas-Cooled Reactor Component Test Facility  

Science Conference Proceedings (OSTI)

A test facility, referred to as the High Temperature Gas-Cooled Reactor Component Test Facility or CTF, will be sited at Idaho National Laboratory for the purposes of supporting development of high temperature gas thermal-hydraulic technologies (helium, helium-Nitrogen, CO2, etc.) as applied in heat transport and heat transfer applications in High Temperature Gas-Cooled Reactors. Such applications include, but are not limited to: primary coolant; secondary coolant; intermediate, secondary, and tertiary heat transfer; and demonstration of processes requiring high temperatures such as hydrogen production. The facility will initially support completion of the Next Generation Nuclear Plant. It will secondarily be open for use by the full range of suppliers, end-users, facilitators, government laboratories, and others in the domestic and international community supporting the development and application of High Temperature Gas-Cooled Reactor technology. This pre-conceptual facility configuration study, which forms the basis for a cost estimate to support CTF scoping and planning, accomplishes the following objectives: • Identifies pre-conceptual design requirements • Develops test loop equipment schematics and layout • Identifies space allocations for each of the facility functions, as required • Develops a pre-conceptual site layout including transportation, parking and support structures, and railway systems • Identifies pre-conceptual utility and support system needs • Establishes pre-conceptual electrical one-line drawings and schedule for development of power needs.

S. L. Austad; L. E. Guillen; D. S. Ferguson; B. L. Blakely; D. M. Pace; D. Lopez; J. D. Zolynski; B. L. Cowley; V. J. Balls; E.A. Harvego, P.E.; C.W. McKnight, P.E.; R.S. Stewart; B.D. Christensen

2008-04-01T23:59:59.000Z

87

Three-dimensional fluid-structure interaction dynamics of a pool-reactor in-tank component. [LMFBR  

SciTech Connect

The safety evaluation of reactor-components often involves the analysis of various types of fluid/structural components interacting in three-dimensional space. For example, in the design of a pool-type reactor several vital in-tank components such as the primary pumps and the intermediate heat exchangers are contained within the primary tank. Typically, these components are suspended from the deck structure and largely submersed in the sodium pool. Because of this positioning these components are vulnerable to structural damage due to pressure wave propagation in the tank during a CDA. In order to assess the structural integrity of these components it is necessary to perform a dynamic analysis in three-dimensional space which accounts for the fluid-structure coupling. A model is developed which has many of the salient features of this fluid-structural component system.

Kulak, R.F.

1979-01-01T23:59:59.000Z

88

THE COMPONENT TEST FACILITY – A NATIONAL USER FACILITY FOR TESTING OF HIGH TEMPERATURE GAS-COOLED REACTOR (HTGR) COMPONENTS AND SYSTEMS  

DOE Green Energy (OSTI)

The Next Generation Nuclear Plant (NGNP) and other High-Temperature Gas-cooled Reactor (HTGR) Projects require research, development, design, construction, and operation of a nuclear plant intended for both high-efficiency electricity production and high-temperature industrial applications, including hydrogen production. During the life cycle stages of an HTGR, plant systems, structures and components (SSCs) will be developed to support this reactor technology. To mitigate technical, schedule, and project risk associated with development of these SSCs, a large-scale test facility is required to support design verification and qualification prior to operational implementation. As a full-scale helium test facility, the Component Test facility (CTF) will provide prototype testing and qualification of heat transfer system components (e.g., Intermediate Heat Exchanger, valves, hot gas ducts), reactor internals, and hydrogen generation processing. It will perform confirmation tests for large-scale effects, validate component performance requirements, perform transient effects tests, and provide production demonstration of hydrogen and other high-temperature applications. Sponsored wholly or in part by the U.S. Department of Energy, the CTF will support NGNP and will also act as a National User Facility to support worldwide development of High-Temperature Gas-cooled Reactor technologies.

David S. Duncan; Vondell J. Balls; Stephanie L. Austad

2008-09-01T23:59:59.000Z

89

Thermocouple shield  

DOE Patents (OSTI)

A thermocouple shield for use in radio frequency fields. In some embodiments the shield includes an electrically conductive tube that houses a standard thermocouple having a thermocouple junction. The electrically conductive tube protects the thermocouple from damage by an RF (including microwave) field and mitigates erroneous temperature readings due to the microwave or RF field. The thermocouple may be surrounded by a ceramic sheath to further protect the thermocouple. The ceramic sheath is generally formed from a material that is transparent to the wavelength of the microwave or RF energy. The microwave transparency property precludes heating of the ceramic sheath due to microwave coupling, which could affect the accuracy of temperature measurements. The ceramic sheath material is typically an electrically insulating material. The electrically insulative properties of the ceramic sheath help avert electrical arcing, which could damage the thermocouple junction. The electrically conductive tube is generally disposed around the thermocouple junction and disposed around at least a portion of the ceramic sheath. The concepts of the thermocouple shield may be incorporated into an integrated shielded thermocouple assembly.

Ripley, Edward B. (Knoxville, TN)

2009-11-24T23:59:59.000Z

90

Mechanical-property degradation of cast stainless steel components from the Shippingport reactor  

SciTech Connect

The mechanical properties of cast stainless steels from the Shippingport reactor have been characterized. Baseline properties for unaged materials were obtained from tests on either recovery-annealed material or material from a cooler region of the component. The materials exhibited modest decrease in impact energy and fracture toughness and a small increase in tensile strength. The fracture toughness J-R curve, J{sub IC} value, tensile flow stress, and Charpy-impact energy of the materials showed very good agreement with estimations based on accelerated laboratory aging studies. The kinetics of thermal embrittlement and degree of embrittlement at saturation, i.e., the minimum impact energy that would be achieved after long-term aging, were established from materials that were aged further in the laboratory at temperatures between 320 and 400{degrees}C. The results showed very good agreement with estimates; the activation energies ranged from 125 to 250 kJ/mole and the minimum room temperature impact energy was <75 J/cm{sup 2}. The estimated impact energy and fracture toughness J-R curve for materials from the Ringhals reactor hot and crossover-leg elbows are also presented.

Chopra, O.K.

1991-10-01T23:59:59.000Z

91

Shielding Integrity Services, Inc.  

Science Conference Proceedings (OSTI)

... Shield attenuation measurement. ... [12/SE-003] IEEE Std 299 (2006) Measuring the Effectiveness of Electromagnetic Shielding Enclosures. ...

2013-09-27T23:59:59.000Z

92

Induced Radioactivity and Waste Classification of Reactor Zone Components of the Chernobyl Nuclear Power Plant Unit 1 After Final Shutdown  

SciTech Connect

The dismantlement of the reactor core materials and surrounding structural components is a major technical concern for those planning closure and decontamination and decommissioning of the Chernobyl Nuclear Power Plant (NPP). Specific issues include when and how dismantlement should be accomplished and what the radwaste classification of the dismantled system would be at the time it is disassembled. Whereas radiation levels and residual radiological characteristics of the majority of the plant systems are directly measured using standard radiation survey and radiochemical analysis techniques, actual measurements of reactor zone materials are not practical due to high radiation levels and inaccessibility. For these reasons, neutron transport analysis was used to estimate induced radioactivity and radiation levels in the Chernobyl NPP Unit 1 reactor core materials and structures.Analysis results suggest that the optimum period of safe storage is 90 to 100 yr for the Unit 1 reactor. For all of the reactor components except the fuel channel pipes (or pressure tubes), this will provide sufficient decay time to allow unlimited worker access during dismantlement, minimize the need for expensive remote dismantlement, and allow for the dismantled reactor components to be classified as low- or medium-level radioactive waste. The fuel channel pipes will remain classified as high-activity waste requiring remote dismantlement for hundreds of years due to the high concentration of induced {sup 63}Ni in the Zircaloy pipes.

Bylkin, Boris K. [Russian Research Center 'Kurchatov Institute' (Russian Federation); Davydova, Galina B. [Russian Research Center 'Kurchatov Institute' (Russian Federation); Zverkov, Yuri A. [Russian Research Center 'Kurchatov Institute' (Russian Federation); Krayushkin, Alexander V. [Russian Research Center 'Kurchatov Institute' (Russian Federation); Neretin, Yuri A. [Chernobyl Nuclear Power Plant (Ukraine); Nosovsky, Anatoly V. [Slavutych Division of the International Chernobyl Center (Ukraine); Seyda, Valery A. [Chernobyl Nuclear Power Plant (Ukraine); Short, Steven M. [Pacific Northwest National Laboratory (United States)

2001-10-15T23:59:59.000Z

93

Eddy current position indicating apparatus for measuring displacements of core components of a liquid metal nuclear reactor  

DOE Patents (OSTI)

Apparatus for measuring displacements of core components of a liquid metal fast breeder reactor by means of an eddy current probe. The active portion of the probe is located within a dry thimble which is supported on a stationary portion of the reactor core support structure. Split rings of metal, having a resistivity significantly different than sodium, are fixedly mounted on the core component to be monitored. The split rings are slidably positioned around, concentric with the probe and symmetrically situated along the axis of the probe so that motion of the ring along the axis of the probe produces a proportional change in the probes electrical output.

Day, Clifford K. (Richland, WA); Stringer, James L. (Richland, WA)

1977-01-01T23:59:59.000Z

94

Review of Experience With the EPRI DFD Process: Decontamination for Decommissioning of Reactor Coolant Systems and Plant Components  

Science Conference Proceedings (OSTI)

This report describes recent experience using the EPRI DFD (decontamination for decommissioning) dilute chemical decontamination process on reactor coolant systems and components. This report updates previous EPRI reports with additional data on materials corrosion and waste management. A wide range of applications show that EPRI DFD is a versatile, cost-effective process for reducing radiation exposures and radioactive waste.

1998-12-31T23:59:59.000Z

95

Operational safety report for the cleaning of non-radioactive, sodium-wetted reactor components with ethyl alcohol  

DOE Green Energy (OSTI)

The safety aspects of the removal of sodium from nonradioactive reactor components by the alcohol process are described in detail. Pertinent properties of alcohol and hydrogen are presented. Relevant excerpts from the Occupational Safety and Health Act safety codes are presented, and a conceptual system is shown illustrating the application of these safety measures.

Humphrey, L.; Felton, L.; Goodman, L.; Pilicy, G.; Welch, F.

1974-01-01T23:59:59.000Z

96

System Engineering Program Applicability for the High Temperature Gas-Cooled Reactor (HTGR) Component Test Capability (CTC)  

SciTech Connect

This white paper identifies where the technical management and systems engineering processes and activities to be used in establishing the High Temperature Gas-cooled Reactor (HTGR) Component Test Capability (CTC) should be addressed and presents specific considerations for these activities under each CTC alternative

Jeffrey Bryan

2009-06-01T23:59:59.000Z

97

Neutronic reactor  

DOE Patents (OSTI)

A safety rod for a nuclear reactor has an inner end portion having a gamma absorption coefficient and neutron capture cross section approximately equal to those of the adjacent shield, a central portion containing materials of high neutron capture cross section and an outer end portion having a gamma absorption coefficient at least equal to that of the adjacent shield.

Wende, Charles W. J. (West Chester, PA)

1976-08-17T23:59:59.000Z

98

Corium shield  

DOE Patents (OSTI)

A shield for restricting molten corium from flowing into a water sump disposed in a floor of a containment vessel includes upper and lower walls which extend vertically upwardly and downwardly from the floor for laterally bounding the sump. The upper wall includes a plurality of laterally spaced apart flow channels extending horizontally therethrough, with each channel having a bottom disposed coextensively with the floor for channeling water therefrom into the sump. Each channel has a height and a length predeterminedly selected for allowing heat from the molten corium to dissipate through the upper and lower walls as it flows therethrough for solidifying the molten corium therein to prevent accumulation thereof in the sump.

McDonald, Douglas B. (Pleasanton, CA); Buchholz, Carol E. (San Jose, CA)

1994-01-01T23:59:59.000Z

99

Mechanical Properties of Welds in Commercial Alloys for High-Temperature Gas-Cooled Reactor Components  

Science Conference Proceedings (OSTI)

C. 1. Mechanical Property / Status of Metallic Materials Development for Application in Advanced High-Temperature Gas-Cooled Reactor / Material

James R. Lindgren; Brian E. Thurgood; Robin H. Ryder; Chia-Chuan Li

100

Material Reliability Program Technical Basis Document Concerning Irradiation-Induced Stress Relaxation and Void Swelling in Pressuri zed Water Reactor Vessel Internals Components (MRP-50)  

Science Conference Proceedings (OSTI)

Irradiation-induced swelling and irradiation-enhanced stress relaxation are two potential degradation mechanisms that could affect reactor vessel (RV) core internals components in pressurized water reactors (PWRs). This report describes current knowledge of these two potential degradation mechanisms, available relevant data and known functional relationships, and a qualitative assessment of these two mechanisms' combined and separate effects on PWR internals components.

2001-10-18T23:59:59.000Z

Note: This page contains sample records for the topic "reactor components shield" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


101

Radiation Shielding Applications  

NLE Websites -- All DOE Office Websites (Extended Search)

Shielding Radiation Shielding Applications Heavy concrete is standard concrete in which conventional aggregate (typically gravel) is replaced with aggregate composed of a dense...

102

rf_shield.PDF  

NLE Websites -- All DOE Office Websites (Extended Search)

April 2001 Interaction Region RF Shield Issues G. Stupakov and T.O. Raubenheimer Stanford Linear Accelerator Center Stanford, California 102300 Interaction Region RF Shield...

103

John R. Shields  

Science Conference Proceedings (OSTI)

John R. Shields. ... Mr. Shields joined the National Bureau of Standards (NBS) in August 1986 as an Engineering Technician at the Large Fire Lab. ...

2012-07-06T23:59:59.000Z

104

Fusion component design for the moving-ring field-reversed mirror reactor  

DOE Green Energy (OSTI)

This partial report on the reactor design contains sections on the following: (1) burner section magnet system design, (2) plasma ring energy recovery, (3) vacuum system, (4) cryogenic system, (5) tritium flows and inventories, and (6) reactor design and layout. (MOW)

Carlson, G.A.

1981-01-28T23:59:59.000Z

105

Method for Producing Components with Internal Architectures, Such as Micro-Channel Reactors, via Diffusion Bonding Sheets  

NLE Websites -- All DOE Office Websites (Extended Search)

partners partners interested in implementing United States Patent Number 7,900,811 entitled "Method for Producing Components with Internal Architectures, Such as Micro-Channel Reactors, via Diffusion Bonding Sheets."Disclosed in this patent is a method for producing microchannels using graduated diffusion bonding of a stack of precision machined foils or sheets (laminates) to make a micro-channel reactor. The method is a novel multi-step process for the diffusion bonding of laminates, which is independent of the channel width-to-fin lamina thickness (fin aspect ratio) and allows for laminae to uniformly and effectively bond. Unlike conventional hot-pressing methods, the NETL invention increases functional reaction surface area for higher conversion efficiency and reactor performance, and avoids

106

Proceedings of the 2000 International Conference on Fatigue of Reactor Components (MRP-46): PWR Materials Reliability Program (PWRMR P)  

Science Conference Proceedings (OSTI)

This report contains information presented at the First International Conference on Fatigue of Reactor Components held July 31 - August 2, 2000, in Napa, California. The conference -- sponsored by EPRI, the Organisation for Economic Co-operation and Development Nuclear Energy Agency/Committee on the Safety of Nuclear Installations (OECD NEA/CSNI), and the U.S. Nuclear Regulatory Commission (U.S. NRC) -- provided a forum for the technical discussion of fatigue issues that affect the integrity and operatio...

2001-06-25T23:59:59.000Z

107

Shielding analysis of neuron emitting sealed sources in S100 pipe component OVERPACK/TRUPACT-11 geometries using MCNP transport code and empirical measurements  

Science Conference Proceedings (OSTI)

The purpose of this work was to estimate an upper bound for the dose at the mid-plane on the surface of a S100 Overpack, i.e., a specially designed drum for transuranium (TRU) waste containing shielding and sealed neutron sources. In addition, the dose at the surface and at a distance for a TRUPACT-II Container with 14 S100 Overpack drums was also estimated for the Normal Conditions of Transport (NCT), for Normal Condition of Transport with crushed Overpacks and for accident conditions. The source modeled was a Plutonium-Beryllium source with the Plutonium infinitely dilute in the Beryllium. In the calculations, a source of one curie per drum was modeled. The resulting dose from the calculations may be scaled to reflect multi-curie level sources. When these results are scaled to 28 curies per drum, the limiting payload, the dose rates obtained meet the requirements of the TRUPACT-II Safety Analysis Report.

Perry, R. T. (Robert T.); Brown, T. H. (Thomas H.); Tompkins, J. A.

2002-01-01T23:59:59.000Z

108

NEUTRONIC REACTORS  

DOE Patents (OSTI)

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

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

1958-10-14T23:59:59.000Z

109

RELIABILITY MODELS OF AGING PASSIVE COMPONENTS INFORMED BY MATERIALS DEGRADATION METRICS TO SUPPORT LONG-TERM REACTOR OPERATIONS  

Science Conference Proceedings (OSTI)

Paper describes a methodology for the synthesis of nuclear power plant service data with expert-elicited materials degradation information to estimate the future failure rates of passive components. This method should be an important resource to long-term plant operations and reactor life extension. Conventional probabilistic risk assessments (PRAs) are not well suited to addressing long-term reactor operations. Since passive structures and components are among those for which replacement can be least practical, they might be expected to contribute increasingly to risk in an aging plant; yet, passives receive limited treatment in PRAs. Furthermore, PRAs produce only snapshots of risk based on the assumption of time-independent component failure rates. This assumption is unlikely to be valid in aging systems. The treatment of aging passive components in PRA presents challenges. Service data to quantify component reliability models are sparse, and this is exacerbated by the greater data demands of age-dependent reliability models. Another factor is that there can be numerous potential degradation mechanisms associated with the materials and operating environment of a given component. This deepens the data problem since risk-informed management of component aging will demand an understanding of the long-term risk significance of individual degradation mechanisms. In this paper we describe a Bayesian methodology that integrates metrics of materials degradation susceptibility with available plant service data to estimate age-dependent passive component reliabilities. Integration of these models into conventional PRA will provide a basis for materials degradation management informed by predicted long-term operational risk.

Unwin, Stephen D.; Lowry, Peter P.; Toyooka, Michael Y.

2012-05-01T23:59:59.000Z

110

DEGRADATION SUSCEPTIBILITY METRICS AS THE BASES FOR BAYESIAN RELIABILITY MODELS OF AGING PASSIVE COMPONENTS AND LONG-TERM REACTOR RISK  

Science Conference Proceedings (OSTI)

Conventional probabilistic risk assessments (PRAs) are not well-suited to addressing long-term reactor operations. Since passive structures, systems and components are among those for which refurbishment or replacement can be least practical, they might be expected to contribute increasingly to risk in an aging plant. Yet, passives receive limited treatment in PRAs. Furthermore, PRAs produce only snapshots of risk based on the assumption of time-independent component failure rates. This assumption is unlikely to be valid in aging systems. The treatment of aging passive components in PRA does present challenges. First, service data required to quantify component reliability models are sparse, and this problem is exacerbated by the greater data demands of age-dependent reliability models. A compounding factor is that there can be numerous potential degradation mechanisms associated with the materials, design, and operating environment of a given component. This deepens the data problem since the risk-informed management of materials degradation and component aging will demand an understanding of the long-term risk significance of individual degradation mechanisms. In this paper we describe a Bayesian methodology that integrates the metrics of materials degradation susceptibility being developed under the Nuclear Regulatory Commission's Proactive Management of Materials of Degradation Program with available plant service data to estimate age-dependent passive component reliabilities. Integration of these models into conventional PRA will provide a basis for materials degradation management informed by the predicted long-term operational risk.

Unwin, Stephen D.; Lowry, Peter P.; Toyooka, Michael Y.; Ford, Benjamin E.

2011-07-17T23:59:59.000Z

111

Feasibility of underwater welding of highly irradiated in-vessel components of boiling-water reactors: A literature review  

SciTech Connect

In February 1997, the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research (RES), initiated a literature review to assess the state of underwater welding technology. In particular, the objective of this literature review was to evaluate the viability of underwater welding in-vessel components of boiling water reactor (BWR) in-vessel components, especially those components fabricated from stainless steels that are subjected to high neutron fluences. This assessment was requested because of the recent increased level of activity in the commercial nuclear industry to address generic issues concerning the reactor vessel and internals, especially those issues related to repair options. This literature review revealed a preponderance of general information about underwater welding technology, as a result of the active research in this field sponsored by the U.S. Navy and offshore oil and gas industry concerns. However, the literature search yielded only a limited amount of information about underwater welding of components in low-fluence areas of BWR in-vessel environments, and no information at all concerning underwater welding experiences in high-fluence environments. Research reported by the staff of the U.S. Department of Energy (DOE) Savannah River Site and researchers from the DOE fusion reactor program proved more fruitful. This research documented relevant experience concerning welding of stainless steel materials in air environments exposed to high neutron fluences. It also addressed problems with welding highly irradiated materials, and primarily attributed those problems to helium-induced cracking in the material. (Helium is produced from the neutron irradiation of boron, an impurity, and nickel.) The researchers found that the amount of helium-induced cracking could be controlled, or even eliminated, by reducing the heat input into the weld and applying a compressive stress perpendicular to the weld path.

Lund, A.L.

1997-11-01T23:59:59.000Z

112

Analytical Study of High Concentration PCB Paint at the Heavy Water Components Test Reactor  

SciTech Connect

This report provides results of an analytical study of high concentration PCB paint in a shutdown nuclear test reactor located at the US Department of Energy's Savannah River Site (SRS). The study was designed to obtain data relevant for an evaluation of potential hazards associated with the use of and exposure to such paints.

Lowry, N.J.

1998-10-21T23:59:59.000Z

113

Prioritization of reactor control components susceptible to fire damage as a consequence of aging  

SciTech Connect

The Fire Vulnerability of Aged Electrical Components Test Program is to identify and assess issues of plant aging that could lead to an increase in nuclear power plant risk because of fires. Historical component data and prior analyses are used to prioritize a list of components with respect to aging and fire vulnerability and the consequences of their failure on plant safety systems. The component list emphasizes safety system control components, but excludes cables, large equipment, and devices encompassed in the Equipment Qualification (EQ) program. The test program selected components identified in a utility survey and developed test and fire conditions necessary to maximize the effectiveness of the test program. Fire damage considerations were limited to purely thermal effects.

Lowry, W.; Vigil, R. [Science and Engineering Associates, Inc., Albuquerque, NM (United States); Nowlen, S. [Sandia National Labs., Albuquerque, NM (United States)

1994-01-01T23:59:59.000Z

114

Analysis of removal alternatives for the Heavy Water Components Test Reactor at the Savannah River Site. Revision 1  

SciTech Connect

This engineering study evaluates different alternatives for decontamination and decommissioning of the Heavy Water Components Test Reactor (HWCTR). Cooled and moderated with pressurized heavy water, this uranium-fueled nuclear reactor was designed to test fuel assemblies for heavy water power reactors. It was operated for this purpose from march of 1962 until December of 1964. Four alternatives studied in detail include: (1) dismantlement, in which all radioactive and hazardous contaminants would be removed, the containment dome dismantled and the property restored to a condition similar to its original preconstruction state; (2) partial dismantlement and interim safe storage, where radioactive equipment except for the reactor vessel and steam generators would be removed, along with hazardous materials, and the building sealed with remote monitoring equipment in place to permit limited inspections at five-year intervals; (3) conversion for beneficial reuse, in which most radioactive equipment and hazardous materials would be removed and the containment building converted to another use such as a storage facility for radioactive materials, and (4) entombment, which involves removing hazardous materials, filling the below-ground structure with concrete, removing the containment dome and pouring a concrete cap on the tomb. Also considered was safe storage, but this approach, which has, in effect, been followed for the past 30 years, did not warrant detailed evaluation. The four other alternatives were evaluate, taking into account factors such as potential effects on the environment, risks, effectiveness, ease of implementation and cost. The preferred alternative was determined to be dismantlement. This approach is recommended because it ranks highest in the comparative analysis, would serve as the best prototype for the site reactor decommissioning program and would be most compatible with site property reuse plans for the future.

Owen, M.B.

1997-04-01T23:59:59.000Z

115

DUCRETE Shielding: A Cost Effective Alternative Radiation Shield  

NLE Websites -- All DOE Office Websites (Extended Search)

Summary Submitted to Spectrum 2000, Sept 24-28, 2000, Chattanooga, TN Summary Submitted to Spectrum 2000, Sept 24-28, 2000, Chattanooga, TN DUCRETE: A Cost Effective Radiation Shielding Material W. J. Quapp, Starmet CMI W. H. Miller, University of Missouri-Columbia James Taylor, Starmet CMI Colin Hundley, Starmet CMI Nancy Levoy, Starmet Corporation 1. INTRODUCTION A consequence of uranium enrichment in the US has been the accumulation of nearly 740,000 metric tons of depleted uranium hexafluoride (UF 6 ) tails. 1 While this material was once considered a feed stock for the United States Breeder Reactor Program, it is no longer needed. Alternative uses of depleted uranium are few. Some have been used for medical isotope transport casks, some for industrial radioactive source shields, some for military anti-tank

116

Thermal hydraulic considerations in liquid-metal-cooled components of tokamak fusion reactors  

Science Conference Proceedings (OSTI)

The basic considerations of MHD thermal hydraulics for liquid-metal-cooled blankets and first walls of tokamak fusion reactors are discussed. The liquid-metal MHD program of Argonne National Laboratory (ANL) dedicated to analytical and experimental investigations of reactor relevant MHD flows and development of relevant thermal hydraulic design tools is presented. The status of the experimental program and examples of local velocity measurements are given. An account of the MHD codes developed to date at ANL is also presented as is an example of a 3-D thermal hydraulic analysis carried out with such codes. Finally, near term plans for experimental investigations and code development are outlined. 20 refs., 8 figs., 1 tab.

Picologlou, B.F.; Reed, C.B.; Hua, T.Q. (Argonne National Lab., IL (USA))

1989-01-01T23:59:59.000Z

117

PRIMARY SHIELDING CALCULATIONS ON THE IBM 650 (ROC CODES)  

SciTech Connect

Four programs written for the IBM 650 to calculate the gamma dose rates in the primary shielding of thermal reactors are described. Their functions are outlined as follows: Program 651-calculates the core attenuation coefficient and equivalent core gamma volumetric source values for a specific core. Program 652- calculates the activation gamma source data in the shield and prepares tabular data in machine storage for Programs 653 and 654. Program 653- calculates the gamma dose rates in the shield due to gammas arising from activation of shield materials. Program 654calculates the gamma dose rates in the shield due to gammas arising in the core. Gamma photo source values are obtained on the basis of two group neutron flux distributions throughout the reactor core and shield. (W.D.M.)

Rosen, S.S.; Oby, P.V.; Caton, R.L.

1958-10-15T23:59:59.000Z

118

The Tower Shielding Facility: Its glorious past  

Science Conference Proceedings (OSTI)

The Tower Shielding Facility (TSF) is the only reactor facility in the US that was designed and built for radiation-shielding studies in which both the reactor source and shield samples could be raised into the air to allow measurements to be made without interference from ground scattering or other spurious effects. The TSF proved its usefulness as many different programs were successfully completed. It became active in work for the Defense Atomic Support Agency (DASA) Space Nuclear Auxiliary Power, Defense Nuclear Agency, Liquid Metal Fast Breeder Reactor Program, the Gas-Cooled and High-Temperature Gas-Cooled Reactor programs, and the Japanese-American Shielding Program of Experimental Research, just to mention a few of the more extensive ones. The history of the TSF as presented in this report describes the various experiments that were performed using the different reactors. The experiments are categorized as to the programs which they supported and placed in corresponding chapters. The experiments are described in modest detail, along with their purpose when appropriate. Discussion of the results is minimal, but references are given to more extensive topical reports.

Muckenthaler, F.J.

1997-05-07T23:59:59.000Z

119

INERT GAS SHIELD FOR WELDING  

DOE Patents (OSTI)

S>An inert gas shield is presented for arc-welding materials such as zirconium that tend to oxidize rapidly in air. The device comprises a rectangular metal box into which the welding electrode is introduced through a rubber diaphragm to provide flexibility. The front of the box is provided with a wlndow having a small hole through which flller metal is introduced. The box is supplied with an inert gas to exclude the atmosphere, and with cooling water to promote the solidification of the weld while in tbe inert atmosphere. A separate water-cooled copper backing bar is provided underneath the joint to be welded to contain the melt-through at the root of the joint, shielding the root of the joint with its own supply of inert gas and cooling the deposited weld metal. This device facilitates the welding of large workpieces of zirconium frequently encountered in reactor construction.

Jones, S.O.; Daly, F.V.

1958-10-14T23:59:59.000Z

120

Unsteady momentum fluxes in two-phase flow and the vibration of nuclear reactor components  

E-Print Network (OSTI)

The steady and unsteady components of the momentum flux in a twophase flow have been measured at the exit of a vertical pipe. Measured momentum flux data has been machine processed by standard random vibration techniques ...

Yih, Tien Sieh

1968-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor components shield" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


121

Argonne Liquid-Metal Advanced Burner Reactor : components and in-vessel system thermal-hydraulic research and testing experience - pathway forward.  

Science Conference Proceedings (OSTI)

This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition from normal high-flow operation to natural circulation. Low-flow coolant events are the most difficult to design for because they involve the most complex thermal-hydraulic behavior induced by the dominance of thermal-buoyancy forces acting on the coolants. Such behavior can cause multiple-component flow interaction phenomena, which are not adequately understood or appreciated by reactor designers as to their impact on reactor performance and safety. Since the early 1990s, when DOE canceled the U.S. Liquid Metal Fast Breeder Reactor (LMFBR) program, little has been done experimentally to further understand the importance of the complex thermal-buoyancy phenomena and their impact on reactor design or to improve the ability of three-dimensional (3-D) transient computational fluid dynamics (CFD) and structures codes to model the phenomena. An improved experimental data base and the associated improved validated codes would provide needed design tools to the reactor community. The improved codes would also facilitate scale-up from small-scale testing to prototype size and would facilitate comparing performance of one reactor/component design with another. The codes would also have relevance to the design and safety of water-cooled reactors. To accomplish the preceding, it is proposed to establish a national GNEP-LMR research and development center at Argonne having as its foundation state-of-art science-based infrastructure consisting of: (a) thermal-hydraulic experimental capabilities for conducting both water and sodium testing of individual reactor components and complete reactor in-vessel models and (b) a computational modeling development and validation capability that is strongly interfaced with the experimental facilities. The proposed center would greatly advance capabilities for reactor development by establishing the validity of high-fidelity (i.e., close to first principles) models and tools. Such tools could be used directly for reactor design or for qualifying/tuning of lower-fidelity models, which now require costly experimental qualification for each different type of design

Kasza, K.; Grandy, C.; Chang, Y.; Khalil, H.; Nuclear Engineering Division

2007-06-30T23:59:59.000Z

122

Thermal testing of solid neutron shielding materials  

Science Conference Proceedings (OSTI)

Two legal-weight truck casks the GA-4 and GA-9, will carry four PWR and nine BWR spent fuel assemblies, respectively. Each cask has a solid neutron shielding material separating the steel body and the outer steel skin. In the thermal accident specified by NRC regulations in 10CFR Part 71, the cask is subjected to an 800[degree]C environment for 30 minutes. The neutron shield need not perform any shielding function during or after the thermal accident, but its behavior must not compromise the ability of the cask to contain the radioactive contents. In May-June 1989 the first series of full-scale thermal tests was performed on three shielding materials: Bisco Products NS-4-FR, and Reactor Experiments RX-201 and RX-207. The tests are described in Thermal Testing of Solid Neutron Shielding Materials, GA-AL 9897, R. H. Boonstra, General Atomics (1990), and demonstrated the acceptability of these materials in a thermal accident. Subsequent design changes to the cask rendered these materials unattractive in terms of weight or adequate service temperature margin. For the second test series, a material specification was developed for a polypropylene based neutron shield with a softening point of at least 280[degree]F. The neutron shield materials tested were boronated (0.8--4.5%) polymers (polypropylene, HDPE, NS-4). The Envirotech and Bisco materials are not polypropylene, but were tested as potential backup materials in the event that a satisfactory polypropylene could not be found.

Boonstra, R.H.

1992-09-01T23:59:59.000Z

123

A CONCEPTUAL DESIGN OF A SHIELD TESTING AND MATERIALS IRRADIATION FACILITY  

SciTech Connect

A conceptual design is presented for a test reactor facility to be used for shielding experiments and component irradintions necessary for airframe development for the nuclear airplane program. To meet both requirements a modified swimming-pool reactor is used, with a dry irradintion cell of 320 cu ft of useful volume provided for component testing, while shielding experiments are performed in the pool in the usual manner. A BSR-type core is operated at 1 MW to provide a fest neutron flux in the irradiation cell of 10/sup 12/n/cm/sup 2/ sec at the core face and 10/sup 11/at a distance of 4 feet. The irradiation-cell facility is designed to avoid the need of remote operations in making up service connections to the experimental piece. The reactor is contained in a cylindrical building designed for 6 psi internal pressure to meet the conditions of the maximum credible accident. The estimated cost of the facility, including the reactor and the fabrication cost for an initial fuel charge, is 874,000. (auth)

Frankfort, J.H.

1956-11-20T23:59:59.000Z

124

Balance of Plant System Analysis and Component Design of Turbo-Machinery for High Temperature Gas Reactor Systems  

Science Conference Proceedings (OSTI)

The Modular Pebble Bed Reactor system (MPBR) requires a gas turbine cycle (Brayton cycle) as the power conversion system for it to achieve economic competitiveness as a Generation IV nuclear system. The availability of controllable helium turbomachinery and compact heat exchangers are thus the critical enabling technology for the gas turbine cycle. The development of an initial reference design for an indirect helium cycle has been accomplished with the overriding constraint that this design could be built with existing technology and complies with all current codes and standards. Using the initial reference design, limiting features were identified. Finally, an optimized reference design was developed by identifying key advances in the technology that could reasonably be expected to be achieved with limited R&D. This final reference design is an indirect, intercooled and recuperated cycle consisting of a three-shaft arrangement for the turbomachinery system. A critical part of the design process involved the interaction between individual component design and overall plant performance. The helium cycle overall efficiency is significantly influenced by performance of individual components. Changes in the design of one component, a turbine for example, often required changes in other components. To allow for the optimization of the overall design with these interdependencies, a detailed steady state and transient control model was developed. The use of the steady state and transient models as a part of an iterative design process represents a key contribution of this work. A dynamic model, MPBRSim, has been developed. The model integrates the reactor core and the power conversion system simultaneously. Physical parameters such as the heat exchangers; weights and practical performance maps such as the turbine characteristics and compressor characteristics are incorporated into the model. The individual component models as well as the fully integrated model of the power conversion system have been verified with an industry-standard general thermal-fluid code Flownet. With respect to the dynamic model, bypass valve control and inventory control have been used as the primary control methods for the power conversion system. By performing simulation using the dynamic model with the designed control scheme, the combination of bypass and inventory control was optimized to assure system stability within design temperature and pressure limits. Bypass control allows for rapid control system response while inventory control allows for ultimate steady state operation at part power very near the optimum operating point for the system. Load transients simulations show that the indirect, three-shaft arrangement gas turbine power conversion system is stable and controllable. For the indirect cycle the intermediate heat exchanger (IHX) is the interface between the reactor and the turbomachinery systems. As a part of the design effort the IHX was identified as the key component in the system. Two technologies, printed circuit and compact plate-fin, were investigated that have the promise of meeting the design requirements for the system. The reference design incorporates the possibility of using either technology although the compact plate-fin design was chosen for subsequent analysis. The thermal design and parametric analysis with an IHX and recuperator using the plate-fin configuration have been performed. As a three-shaft arrangement, the turbo-shaft sets consist of a pair of turbine/compressor sets (high pressure and low pressure turbines with same-shaft compressor) and a power turbine coupled with a synchronous generator. The turbines and compressors are all axial type and the shaft configuration is horizontal. The core outlet/inlet temperatures are 900/520 C, and the optimum pressure ratio in the power conversion cycle is 2.9. The design achieves a plant net efficiency of approximately 48%.

Ronald G. Ballinger Chunyun Wang Andrew Kadak Neil Todreas

2004-08-30T23:59:59.000Z

125

Structural mechanics of fast spectrum nuclear reactor cores  

NLE Websites -- All DOE Office Websites (Extended Search)

mechanics of fast spectrum nuclear reactor cores A fast reactor core is composed of a closely packed hexagonal arrangement of fuel, control, blanket , and shielding assemblies....

126

Cable shield connecting device  

DOE Patents (OSTI)

A cable shield connecting device for installation on a high voltage cable of the type having a metallic shield, the device including a relatively conformable, looped metal bar for placement around a bared portion of the metallic shield to extend circumferentially around a major portion of the circumference of the metallic shield while being spaced radially therefrom, a plurality of relatively flexible metallic fingers affixed to the bar, projecting from the bar in an axial direction and spaced circumferentially along the bar, each finger being attached to the metallic shield at a portion located remote from the bar to make electrical contact with the metallic shield, and a connecting conductor integral with the bar.

Silva, Frank A. (Basking Ridge, NJ)

1979-01-01T23:59:59.000Z

127

REACTOR  

DOE Patents (OSTI)

A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

Roman, W.G.

1961-06-27T23:59:59.000Z

128

RADIATION SHIELDING DEVICE  

DOE Patents (OSTI)

ABS>A radiation shield that is suitable for the protection of personnel from both gamma rays and nentrons is described. The shield is comprised of a hollow wall and an aggregate consisting of iron and water in approximately equal amounts by volume substantially filling the wall. A means is provided to circulate the water through the wall to cool the shield when in use.

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

1958-09-23T23:59:59.000Z

129

Shield for Water Boiler  

SciTech Connect

Siimplified shielding calculations indicating the proposed design for the water boiler assembly will reduce the radiation at normal operaton to values well below those which are considered tolerable.

Balent, R.

1951-08-08T23:59:59.000Z

130

RADIATION SHIELDING COMPOSITION  

DOE Patents (OSTI)

A light weight radiation shielding composition is described whose mechanical and radiological properties can be varied within wide limits. The composition of this shielding material consists of four basic ingredients: powder of either Pb or W, a plastic resin, a resin plasticizer, and a polymerization catalyst to promote an interaction of the plasticizer with the plastic resin. Air may be mixed into the above ingredients in order to control the density of the final composition. For equivalent gamma attenuation, the shielding composition weighs one-third to one-half as much as conventional Pb shielding. (AEC)

Dunegan, H.L.

1963-01-29T23:59:59.000Z

131

Large Component Removal/Disposal  

Science Conference Proceedings (OSTI)

This paper describes the removal and disposal of the large components from Maine Yankee Atomic Power Plant. The large components discussed include the three steam generators, pressurizer, and reactor pressure vessel. Two separate Exemption Requests, which included radiological characterizations, shielding evaluations, structural evaluations and transportation plans, were prepared and issued to the DOT for approval to ship these components; the first was for the three steam generators and one pressurizer, the second was for the reactor pressure vessel. Both Exemption Requests were submitted to the DOT in November 1999. The DOT approved the Exemption Requests in May and July of 2000, respectively. The steam generators and pressurizer have been removed from Maine Yankee and shipped to the processing facility. They were removed from Maine Yankee's Containment Building, loaded onto specially designed skid assemblies, transported onto two separate barges, tied down to the barges, th en shipped 2750 miles to Memphis, Tennessee for processing. The Reactor Pressure Vessel Removal Project is currently under way and scheduled to be completed by Fall of 2002. The planning, preparation and removal of these large components has required extensive efforts in planning and implementation on the part of all parties involved.

Wheeler, D. M.

2002-02-27T23:59:59.000Z

132

Gamma ray detector shield  

DOE Patents (OSTI)

A gamma ray detector shield comprised of a rigid, lead, cylindrical-shaped vessel having upper and lower portions with an pneumatically driven, sliding top assembly. Disposed inside the lead shield is a gamma ray scintillation crystal detector. Access to the gamma detector is through the sliding top assembly.

Ohlinger, R.D.; Humphrey, H.W.

1985-08-26T23:59:59.000Z

133

Nuclear reactor  

DOE Patents (OSTI)

A nuclear reactor in which the core components, including fuel-rod assemblies, control-rod assemblies, fertile rod-assemblies, and removable shielding assemblies, are supported by a plurality of separate inlet modular units. These units are referred to as inlet module units to distinguish them from the modules of the upper internals of the reactor. The modular units are supported, each removable independently of the others, in liners in the supporting structure for the lower internals of the reactor. The core assemblies are removably supported in integral receptacles or sockets of the modular units. The liners, units, sockets and assmblies have inlet openings for entry of the fluid. The modular units are each removably mounted in the liners with fluid seals interposed between the opening in the liner and inlet module into which the fluid enters and the upper and lower portion of the liner. Each assembly is similarly mounted in a corresponding receptacle with fluid seals interposed between the openings where the fluid enters and the lower portion of the receptacle or fitting closely in these regions. As fluid flows along each core assembly a pressure drop is produced along the fluid so that the fluid which emerges from each core assembly is at a lower pressure than the fluid which enters the core assembly. However because of the seals interposed in the mountings of the units and assemblies the pressures above and below the units and assemblies are balanced and the units are held in the liners and the assemblies are held in the receptacles by their weights as they have a higher specific gravity than the fluid. The low-pressure spaces between each module and its liner and between each core assembly and its module is vented to the low-pressure regions of the vessel to assure that fluid which leaks through the seals does not accumulate and destroy the hydraulic balance.

Pennell, William E. (Greensburg, PA); Rowan, William J. (Monroeville, PA)

1977-01-01T23:59:59.000Z

134

Decontamination of Reactor Systems and Contaminated Components for Disposal or Refurbishment: Developments and Experience with the EPRI DFD Chemical Decontamination Process  

Science Conference Proceedings (OSTI)

This report describes recent developments using the EPRI (Decontamination for Decommissioning (DFD) dilute chemical decontamination process on reactor coolant systems and components. The nuclear industry has now used the process for a variety of applications, from cleaning small components to full system decontamination of retired plants. The report documents substantial savings in radiation exposures during decommissioning work and discusses the benefits of using the process to reduce radioactive waste ...

2001-11-05T23:59:59.000Z

135

Cosmic Ray Interactions in Shielding Materials  

SciTech Connect

This document provides a detailed study of materials used to shield against the hadronic particles from cosmic ray showers at Earth’s surface. This work was motivated by the need for a shield that minimizes activation of the enriched germanium during transport for the MAJORANA collaboration. The materials suitable for cosmic-ray shield design are materials such as lead and iron that will stop the primary protons, and materials like polyethylene, borated polyethylene, concrete and water that will stop the induced neutrons. The interaction of the different cosmic-ray components at ground level (protons, neutrons, muons) with their wide energy range (from kilo-electron volts to giga-electron volts) is a complex calculation. Monte Carlo calculations have proven to be a suitable tool for the simulation of nucleon transport, including hadron interactions and radioactive isotope production. The industry standard Monte Carlo simulation tool, Geant4, was used for this study. The result of this study is the assertion that activation at Earth’s surface is a result of the neutronic and protonic components of the cosmic-ray shower. The best material to shield against these cosmic-ray components is iron, which has the best combination of primary shielding and minimal secondary neutron production.

Aguayo Navarrete, Estanislao; Kouzes, Richard T.; Ankney, Austin S.; Orrell, John L.; Berguson, Timothy J.; Troy, Meredith D.

2011-09-08T23:59:59.000Z

136

POWER REACTOR  

DOE Patents (OSTI)

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

Zinn, W.H.

1958-07-01T23:59:59.000Z

137

SHIELD certification package  

Science Conference Proceedings (OSTI)

Certification as applied to existing computer codes includes the verification and validation process, placing the code in configuration control, establishing user qualification standards and training requirements. All software intended for use in critical calculations must be certified. This report is intended to fulfill the requirements for the certification of the SHIELD, SHLDED, GEDIT, GENPRT, FIPROD, FPCALC, and PROCES modules of the SHIELD system built February, 1992, by W.S. Parks. These modules are used for burnup, cooling, separate, and edit calculations.

Boman, C.

1992-02-01T23:59:59.000Z

138

Cross section space shielding in double heterogeneous HTGR systems  

SciTech Connect

Collision probability methods using rational approximations for space shielding of cross sections in nuclear reactor codes allowing for one level of heterogeneity are first reviewed. A Dancoff factor for a cylindrical HTGR fuel pin with grain structure is then derived. A new simple method for cross section space shielding in a doubly heterogeneous HTGR system using collision probabilities and rational approximations is subsequently presented and calculations based on it are compared with calculations based on more elaborate methods. 39 references. (auth)

Stamatelatos, M.G.

1975-11-01T23:59:59.000Z

139

HWMA/RCRA CLOSURE PLAN FOR THE MATERIALS TEST REACTOR WING (TRA-604) LABORATORY COMPONENTS VOLUNTARY CONSENT ORDER ACTION PLAN VCO-5.8 D REVISION2  

SciTech Connect

This Hazardous Waste Management Act/Resource Conservation and Recovery Act closure plan was developed for the laboratory components of the Test Reactor Area Catch Tank System (TRA-630) that are located in the Materials Test Reactor Wing (TRA-604) at the Reactor Technology Complex, Idaho National Laboratory Site, to meet a further milestone established under Voluntary Consent Order Action Plan VCO-5.8.d. The TRA-604 laboratory components addressed in this closure plan were deferred from the TRA-630 Catch Tank System closure plan due to ongoing laboratory operations in the areas requiring closure actions. The TRA-604 laboratory components include the TRA-604 laboratory warm wastewater drain piping, undersink drains, subheaders, and the east TRA-604 laboratory drain header. Potentially contaminated surfaces located beneath the TRA-604 laboratory warm wastewater drain piping and beneath the island sinks located in Laboratories 126 and 128 (located in TRA-661) are also addressed in this closure plan. The TRA-604 laboratory components will be closed in accordance with the interim status requirements of the Hazardous Waste Management Act/Resource Conservation and Recovery Act as implemented by the Idaho Administrative Procedures Act 58.01.05.009 and 40 Code of Federal Regulations 265, Subparts G and J. This closure plan presents the closure performance standards and the methods for achieving those standards.

KIRK WINTERHOLLER

2008-02-25T23:59:59.000Z

140

Proposed and existing passive and inherent safety-related structures, systems, and components (building blocks) for advanced light-water reactors  

Science Conference Proceedings (OSTI)

A nuclear power plant is composed of many structures, systems, and components (SSCs). Examples include emergency core cooling systems, feedwater systems, and electrical systems. The design of a reactor consists of combining various SSCs (building blocks) into an integrated plant design. A new reactor design is the result of combining old SSCs in new ways or use of new SSCs. This report identifies, describes, and characterizes SSCs with passive and inherent features that can be used to assure safety in light-water reactors. Existing, proposed, and speculative technologies are described. The following approaches were used to identify the technologies: world technical literature searches, world patent searches, and discussions with universities, national laboratories and industrial vendors. 214 refs., 105 figs., 26 tabs.

Forsberg, C.W.; Moses, D.L.; Lewis, E.B.; Gibson, R.; Pearson, R.; Reich, W.J.; Murphy, G.A.; Staunton, R.H.; Kohn, W.E.

1989-10-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor components shield" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


141

Airflow Characteristics of Commonly Used Temperature Radiation Shields  

Science Conference Proceedings (OSTI)

The air temperature radiation shield is a key component in air temperature measurement in weather station networks; however, it is widely recognized that significant errors in the measured air temperature exist due to insufficient airflow past ...

X. Lin; Kenneth G. Hubbard; George E. Meyer

2001-03-01T23:59:59.000Z

142

Materials Reliability Program: Second International Conference on Fatigue of Reactor Components (MRP-84): July 29-August 1, 2002; Sn owbird Ski and Conference Center, Snowbird, Utah  

Science Conference Proceedings (OSTI)

This proceedings contains information presented at the Second International Conference on Fatigue of Reactor Components held 31 July - 1 August, 2002 in Snowbird, Utah. This second conference, again sponsored by EPRI, the Organisation for Economic Co-operation and Development (OECD), the Nuclear Energy Agency/Committee on the Safety of Nuclear Installations (NEA/CSNI), and the U.S. Nuclear Regulatory Commission (NRC), provided a forum for the technical discussion of fatigue issues that affect the integri...

2003-03-04T23:59:59.000Z

143

Composition for radiation shielding  

DOE Patents (OSTI)

A composition for use as a radiation shield. The shield has a depleted urum core for absorbing gamma rays and a bismuth coating for preventing chemical corrosion and absorbing gamma rays. Alternatively, a sheet of gadolinium may be positioned between the uranium core and the bismuth coating for absorbing neutrons. The composition is preferably in the form of a container for storing materials that emit radiation such as gamma rays and neutrons. The container is preferably formed by casting bismuth around a pre-formed uranium container having a gadolinium sheeting, and allowing the bismuth to cool. The resulting container is a structurally sound, corrosion-resistant, radiation-absorbing container.

Kronberg, James W. (Aiken, SC)

1994-01-01T23:59:59.000Z

144

Glove box shield  

DOE Patents (OSTI)

According to the present invention, a shield for a glove box housing radioactive material is comprised of spaced apart clamping members which maintain three overlapping flaps in place therebetween. There is a central flap and two side flaps, the side flaps overlapping at the interior edges thereof and the central flap extending past the intersection of the side flaps in order to insure that the shield is always closed when the user withdraws his hand from the glove box. Lead loaded neoprene rubber is the preferred material for the three flaps, the extent of lead loading depending upon the radiation levels within the glove box.

Brackenbush, Larry W. (Richland, WA); Hoenes, Glenn R. (Richland, WA)

1981-01-01T23:59:59.000Z

145

NEUTRONIC REACTOR BURIAL ASSEMBLY  

DOE Patents (OSTI)

A burial assembly is shown whereby an entire reactor core may be encased with lead shielding, withdrawn from the reactor site and buried. This is made possible by a five-piece interlocking arrangement that may be easily put together by remote control with no aligning of bolt holes or other such close adjustments being necessary.

Treshow, M.

1961-05-01T23:59:59.000Z

146

Lightweight blast shield  

DOE Patents (OSTI)

A tandem warhead missile arrangement that has a composite material housing structure with a first warhead mounted at one end and a second warhead mounted near another end of the composite structure with a dome shaped composite material blast shield mounted between the warheads to protect the second warhead from the blast of the first warhead.

Mixon, Larry C. (Madison, AL); Snyder, George W. (Huntsville, AL); Hill, Scott D. (Toney, AL); Johnson, Gregory L. (Decatur, AL); Wlodarski, J. Frank (Huntsville, AL); von Spakovsky, Alexis P. (Huntsville, AL); Emerson, John D. (Arab, AL); Cole, James M. (Huntsville, AL); Tipton, John P. (Huntsville, AL)

1991-01-01T23:59:59.000Z

147

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.

Wigner, E.P.; Weinberg, A.W.; Young, G.J.

1958-04-15T23:59:59.000Z

148

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A nuclear reactor for isotope production is described. This reactor is designed to provide a maximum thermal neutron flux in a region adjacent to the periphery of the reactor rather than in the center of the reactor. The core of the reactor is generally centrally located with respect tn a surrounding first reflector, constructed of beryllium. The beryllium reflector is surrounded by a second reflector, constructed of graphite, which, in tune, is surrounded by a conventional thermal shield. Water is circulated through the core and the reflector and functions both as a moderator and a coolant. In order to produce a greatsr maximum thermal neutron flux adjacent to the periphery of the reactor rather than in the core, the reactor is designed so tbat the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the materials in the reflector is approximately twice the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the material of the core of the reactor.

Wigner, E.P.

1958-04-22T23:59:59.000Z

149

Assessment of Dancoff adjusted Wigner-Seitz cells for self-shielding LWR lattices  

E-Print Network (OSTI)

The objective of this thesis was to assess the effectiveness of using a Wigner-Seitz (WS) cell with an adjusted moderator thickness to produce more accurate resonance self-shielded cross sections for light water reactor ...

Roomy, Thomas Hayward

2012-01-01T23:59:59.000Z

150

Composition for radiation shielding  

DOE Patents (OSTI)

A composition for use as a radiation shield is disclosed. The shield has a depleted uranium core for absorbing gamma rays and a bismuth coating for preventing chemical corrosion and absorbing gamma rays. Alternatively, a sheet of gadolinium may be positioned between the uranium core and the bismuth coating for absorbing neutrons. The composition is preferably in the form of a container for storing materials that emit radiation such as gamma rays and neutrons. The container is preferably formed by casting bismuth around a pre-formed uranium container having a gadolinium sheeting, and allowing the bismuth to cool. The resulting container is a structurally sound, corrosion-resistant, radiation-absorbing container. 2 figs.

Kronberg, J.W.

1994-08-02T23:59:59.000Z

151

Multilayer radiation shield  

SciTech Connect

A power generation system including: a generator including a rotor including a superconductive rotor coil coupled to a rotatable shaft; a first prime mover drivingly coupled to the rotatable shaft; and a thermal radiation shield, partially surrounding the rotor coil, including at least a first sheet and a second sheet spaced apart from the first sheet by centripetal force produced by the rotatable shaft. A thermal radiation shield for a generator including a rotor including a super-conductive rotor coil including: a first sheet having at least one surface formed from a low emissivity material; and at least one additional sheet having at least one surface formed from a low emissivity material spaced apart from the first sheet by centripetal force produced by the rotatable shaft, wherein each successive sheet is an incrementally greater circumferential arc length and wherein the centripetal force shapes the sheets into a substantially catenary shape.

Urbahn, John Arthur (Saratoga Springs, NY); Laskaris, Evangelos Trifon (Niskayuna, NY)

2009-06-16T23:59:59.000Z

152

Statistically based uncertainty analysis for ranking of component importance in the thermal-hydraulic safety analysis of the Advanced Neutron Source Reactor  

SciTech Connect

The Analytic Hierarchy Process (AHP) has been used to help determine the importance of components and phenomena in thermal-hydraulic safety analyses of nuclear reactors. The AHP results are based, in part on expert opinion. Therefore, it is prudent to evaluate the uncertainty of the AHP ranks of importance. Prior applications have addressed uncertainty with experimental data comparisons and bounding sensitivity calculations. These methods work well when a sufficient experimental data base exists to justify the comparisons. However, in the case of limited or no experimental data the size of the uncertainty is normally made conservatively large. Accordingly, the author has taken another approach, that of performing a statistically based uncertainty analysis. The new work is based on prior evaluations of the importance of components and phenomena in the thermal-hydraulic safety analysis of the Advanced Neutron Source Reactor (ANSR), a new facility now in the design phase. The uncertainty during large break loss of coolant, and decay heat removal scenarios is estimated by assigning a probability distribution function (pdf) to the potential error in the initial expert estimates of pair-wise importance between the components. Using a Monte Carlo sampling technique, the error pdfs are propagated through the AHP software solutions to determine a pdf of uncertainty in the system wide importance of each component. To enhance the generality of the results, study of one other problem having different number of elements is reported, as are the effects of a larger assumed pdf error in the expert ranks. Validation of the Monte Carlo sample size and repeatability are also documented.

Wilson, G.E.

1992-01-01T23:59:59.000Z

153

Justification for Shielded Receiver Tube Additional Lead Shielding  

Science Conference Proceedings (OSTI)

In order to reduce high radiation dose rates encountered when core sampling some radioactive waste tanks the addition of 240 lbs. of lead shielding is being considered to the shielded receiver tube on core sample trucks No.1, No.3 and No.4. The lead shielding is 4 inch diameter x 1/2 inch thick half rounds that have been installed around the SR tube over its' full length. Using three unreleased but independently reviewed structural analyses HNF-6018 justifies the addition of the lead shielding.

BOGER, R.M.

2000-04-11T23:59:59.000Z

154

RADIATION FACILITY FOR NUCLEAR REACTORS  

DOE Patents (OSTI)

A radiation facility is designed for irradiating samples in close proximity to the core of a nuclear reactor. The facility comprises essentially a tubular member extending through the biological shield of the reactor and containing a manipulatable rod having the sample carrier at its inner end, the carrier being longitudinally movable from a position in close proximity to the reactor core to a position between the inner and outer faces of the shield. Shield plugs are provided within the tubular member to prevent direct radiation from the core emanating therethrough. In this device, samples may be inserted or removed during normal operation of the reactor without exposing personnel to direct radiation from the reactor core. A storage chamber is also provided within the radiation facility to contain an irradiated sample during the period of time required to reduce the radioactivity enough to permit removal of the sample for external handling. (AEC)

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

1961-12-12T23:59:59.000Z

155

Actively driven thermal radiation shield  

DOE Patents (OSTI)

A thermal radiation shield for cooled portable gamma-ray spectrometers. The thermal radiation shield is located intermediate the vacuum enclosure and detector enclosure, is actively driven, and is useful in reducing the heat load to mechanical cooler and additionally extends the lifetime of the mechanical cooler. The thermal shield is electrically-powered and is particularly useful for portable solid-state gamma-ray detectors or spectrometers that dramatically reduces the cooling power requirements. For example, the operating shield at 260K (40K below room temperature) will decrease the thermal radiation load to the detector by 50%, which makes possible portable battery operation for a mechanically cooled Ge spectrometer.

Madden, Norman W. (Livermore, CA); Cork, Christopher P. (Pleasant Hill, CA); Becker, John A. (Alameda, CA); Knapp, David A. (Livermore, CA)

2002-01-01T23:59:59.000Z

156

Steam Generator Management Program: Effects of Different pH Control Agents on Pressurized Water Reactor Plant Systems and Components  

Science Conference Proceedings (OSTI)

Corrosion of materials in the condensate, feedwater, and drain systems of PWRs generates a significant amount of corrosion products in the secondary cycle. These corrosion products are generally transported into the steam generators and deposit on tubing surfaces, tubesheets, and tube support plates. Increased corrosion results in elevated levels of undesired corrosion products being deposited into the steam generators. To minimize corrosion of the secondary system components, control of pH in the second...

2009-12-04T23:59:59.000Z

157

shields-98.pdf  

NLE Websites -- All DOE Office Websites (Extended Search)

7 7 The Whole Sky Imager - A Year of Progress J. E. Shields and M. E. Karr Marine Physical Laboratory Scripps Institution of Oceanography University of California, San Diego San Diego, California T. P. Tooman Sandia National Laboratories Livermore, California D. H. Sowle and S. T. Moore Mission Research Corporation Santa Barbara, California Abstract Much progress has been made this last year in realizing the potential of the whole sky imager (WSI). Two imagers are deployed [at the Southern Great Plains (SGP) site and the Surface Heat Budget of the Arctic Ocean (SHEBA)], two are being prepared for deployment in the Tropical Western Pacific (TWP), and more are in production. Data products now include daytime thick cloud fraction and calibrated radiance. Night cloud fraction and daytime thin cloud

158

Shield Volcano | Open Energy Information  

Open Energy Info (EERE)

source source History View New Pages Recent Changes All Special Pages Semantic Search/Querying Get Involved Help Apps Datasets Community Login | Sign Up Search Page Edit History Facebook icon Twitter icon » Shield Volcano Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Print PDF Shield Volcano Dictionary.png Shield Volcano: A dome shaped volcano with gently sloping sides and a broad base characteristic of relatively low viscosity, basaltic lava eruptions. Other definitions:Wikipedia Reegle Topographic Features List of topographic features commonly encountered in geothermal resource areas: Mountainous Horst and Graben Shield Volcano Flat Lava Dome Stratovolcano Cinder Cone Caldera Depression Resurgent Dome Complex Schematic representation of the internal structure of a typical shield

159

Welding shield for coupling heaters  

DOE Patents (OSTI)

Systems for coupling end portions of two elongated heater portions and methods of using such systems to treat a subsurface formation are described herein. A system may include a holding system configured to hold end portions of the two elongated heater portions so that the end portions are abutted together or located near each other; a shield for enclosing the end portions, and one or more inert gas inlets configured to provide at least one inert gas to flush the system with inert gas during welding of the end portions. The shield may be configured to inhibit oxidation during welding that joins the end portions together. The shield may include a hinged door that, when closed, is configured to at least partially isolate the interior of the shield from the atmosphere. The hinged door, when open, is configured to allow access to the interior of the shield.

Menotti, James Louis (Dickinson, TX)

2010-03-09T23:59:59.000Z

160

Portable convertible blast effects shield  

DOE Patents (OSTI)

A rapidly deployable portable convertible blast effects shield/ballistic shield includes a set two or more telescoping cylindrical rings operably connected to each other to convert between a telescopically-collapsed configuration for storage and transport, and a telescopically-extended upright configuration forming an expanded inner volume. In a first embodiment, the upright configuration provides blast effects shielding, such as against blast pressures, shrapnel, and/or fire balls. And in a second embodiment, the upright configuration provides ballistic shielding, such as against incoming weapons fire, shrapnel, etc. Each ring has a high-strength material construction, such as a composite fiber and matrix material, capable of substantially inhibiting blast effects and impinging projectiles from passing through the shield. And the set of rings are releasably securable to each other in the telescopically-extended upright configuration, such as by click locks.

Pastrnak, John W. (Livermore, CA); Hollaway, Rocky (Modesto, CA); Henning, Carl D. (Livermore, CA); Deteresa, Steve (Livermore, CA); Grundler, Walter (Hayward, CA); Hagler, Lisle B. (Berkeley, CA); Kokko, Edwin (Dublin, CA); Switzer, Vernon A (Livermore, CA)

2007-05-22T23:59:59.000Z

Note: This page contains sample records for the topic "reactor components shield" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


161

Portable convertible blast effects shield  

DOE Patents (OSTI)

A rapidly deployable portable convertible blast effects shield/ballistic shield includes a set two or more telescoping cylindrical rings operably connected to each other to convert between a telescopically-collapsed configuration for storage and transport, and a telescopically-extended upright configuration forming an expanded inner volume. In a first embodiment, the upright configuration provides blast effects shielding, such as against blast pressures, shrapnel, and/or fire balls. And in a second embodiment, the upright configuration provides ballistic shielding, such as against incoming weapons fire, shrapnel, etc. Each ring has a high-strength material construction, such as a composite fiber and matrix material, capable of substantially inhibiting blast effects and impinging projectiles from passing through the shield. And the set of rings are releasably securable to each other in the telescopically-extended upright configuration, such as by click locks.

Pastrnak, John W. (Livermore, CA); Hollaway, Rocky (Modesto, CA); Henning, Carl D. (Livermore, CA); Deteresa, Steve (Livermore, CA); Grundler, Walter (Hayward, CA); Hagler,; Lisle B. (Berkeley, CA); Kokko, Edwin (Dublin, CA); Switzer, Vernon A (Livermore, CA)

2010-10-26T23:59:59.000Z

162

Portable convertible blast effects shield  

DOE Patents (OSTI)

A rapidly deployable portable convertible blast effects shield/ballistic shield includes a set two or more frusto-conically-tapered telescoping rings operably connected to each other to convert between a telescopically-collapsed configuration for storage and transport, and a telescopically-extended upright configuration forming an expanded inner volume. In a first embodiment, the upright configuration provides blast effects shielding, such as against blast pressures, shrapnel, and/or fire balls. And in a second embodiment, the upright configuration provides ballistic shielding, such as against incoming weapons fire, shrapnel, etc. Each ring has a high-strength material construction, such as a composite fiber and matrix material, capable of substantially inhibiting blast effects and impinging projectiles from passing through the shield. And the set of rings are releasably securable to each other in the telescopically-extended upright configuration by the friction fit of adjacent pairs of frusto-conically-tapered rings to each other.

Pastrnak, John W. (Livermore, CA); Hollaway, Rocky (Modesto, CA); Henning, Carl D. (Livermore, CA); Deteresa, Steve (Livermore, CA); Grundler, Walter (Hayward, CA); Hagler, Lisle B. (Berkeley, CA); Kokko, Edwin (Dublin, CA); Switzer, Vernon A. (Livermore, CA)

2011-03-15T23:59:59.000Z

163

Status of NAÏADE 1 Shielding Benchmarks  

Science Conference Proceedings (OSTI)

Shielding / Special Issue on the 11th International Conference on Radiation Shielding and the 15th Topical Meeting of the Radiation Protection and Shielding Division (Part 2) / Radiation Protection

Jean-Claude Nimal

164

A (CERN) History of Accelerator Shielding  

Science Conference Proceedings (OSTI)

Shielding / Special Issue on the 11th International Conference on Radiation Shielding and the 15th Topical Meeting of the Radiation Protection and Shielding Division (Part 2) / Radiation Protection

Marco Silari

165

Development of Novel Neutron Shielding Concrete  

Science Conference Proceedings (OSTI)

Shielding Materials / Special Issue on the 11th International Conference on Radiation Shielding and the 15th Topical Meeting of the Radiation Protection and Shielding Division (Part 2) / Radiation Protection

Koichi Okuno; Masayoshi Kawai; Hitoshi Yamada

166

High-Heat-Resistant Neutron Shielding Resin  

Science Conference Proceedings (OSTI)

Shielding Materials / Special Issue on the 11th International Conference on Radiation Shielding and the 15th Topical Meeting of the Radiation Protection and Shielding Division (Part 2) / Radiation Protection

Atsuhiko M. Sukegawa; Koichi Okuno; Shinji Sakurai

167

SNF shipping cask shielding analysis  

Science Conference Proceedings (OSTI)

The Waste Management and Remedial Action Division has planned a modification sequence for storage facility 7827 in the Solid Waste Storage Area (SWSA). The modification cycle is: (1) modify an empty caisson, (2) transfer the spent nuclear fuel (SNF) of an occupied caisson to a hot cell in building 3525 for inspection and possible repackaging, and (3) return the package to the modified caisson in the SWSA. Although the SNF to be moved is in the solid form, it has different levels of activity. Thus, the following 5 shipping casks will be available for the task: the Loop Transport Carrier, the In- Pile Loop LITR HB-2 Carrier, the 6.5-inch HRLEL Carrier, the HFIR Hot Scrap Carrier, and the 10-inch ORR Experiment Removal Shield Cask. This report describes the shielding tasks for the 5 casks: determination of shielding characteristics, any streaming avenues, estimation of thermal limits, and shielding calculational uncertainty for use in the transportation plan.

Johnson, J.O.; Pace, J.V. III

1996-01-01T23:59:59.000Z

168

Transient heat flux shielding using thermal metamaterials  

E-Print Network (OSTI)

We have developed a heat shield based on a metamaterial engineering approach to shield a region from transient diffusive heat flow. The shield is designed with a multilayered structure to prescribe the appropriate spatial profile for heat capacity, density, and thermal conductivity of the effective medium. The heat shield was experimentally compared to other isotropic materials.

Narayana, Supradeep; Sato, Yuki

2013-01-01T23:59:59.000Z

169

THE EFFECT OF A MINIMUM WEIGHT RADIAL REFLECTOR ON SNAP SHIELDING REQUIREMENTS  

SciTech Connect

A model was derived for a minimum-weight radial reflector for SNAP reactors. The effect, which the use of this optimal reflector has on radiation shield weight requirements is investigated. Weights of systems employing conventional and optimal radial reflectors are compared using the FARSE and FARSER computer codes. It is found that for the configuration under study additional shield weight required when the optimal reflector is used is in excess of the reflector weight savings. (auth)

Bernick, R.L.

1963-10-17T23:59:59.000Z

170

Gas-Cooled Fast Reactor Program. Annual progress report for period ending December 31, 1979  

SciTech Connect

Information on the GCFR reactor is presented concerning the Core Flow Test Loop; shielding and physics; pressure vessel and closure studies; and irradiation program.

Gat, U.; Kasten, P.R.

1980-11-01T23:59:59.000Z

171

Radiation shielding materials and containers incorporating same  

DOE Patents (OSTI)

An improved radiation shielding material and storage systems for radioactive materials incorporating the same. The PYRolytic Uranium Compound ("PYRUC") shielding material is preferably formed by heat and/or pressure treatment of a precursor material comprising microspheres of a uranium compound, such as uranium dioxide or uranium carbide, and a suitable binder. The PYRUC shielding material provides improved radiation shielding, thermal characteristic, cost and ease of use in comparison with other shielding materials. The shielding material can be used to form containment systems, container vessels, shielding structures, and containment storage areas, all of which can be used to house radioactive waste. The preferred shielding system is in the form of a container for storage, transportation, and disposal of radioactive waste. In addition, improved methods for preparing uranium dioxide and uranium carbide microspheres for use in the radiation shielding materials are also provided.

Mirsky, Steven M. (Greenbelt, MD); Krill, Stephen J. (Arlington, VA); Murray, Alexander P. (Gaithersburg, MD)

2005-11-01T23:59:59.000Z

172

Radiation Shielding Materials and Containers Incorporating Same  

Science Conference Proceedings (OSTI)

An improved radiation shielding material and storage systems for radioactive materials incorporating the same. The PYRolytic Uranium Compound (''PYRUC'') shielding material is preferably formed by heat and/or pressure treatment of a precursor material comprising microspheres of a uranium compound, such as uranium dioxide or uranium carbide, and a suitable binder. The PYRUC shielding material provides improved radiation shielding, thermal characteristic, cost and ease of use in comparison with other shielding materials. The shielding material can be used to form containment systems, container vessels, shielding structures, and containment storage areas, all of which can be used to house radioactive waste. The preferred shielding system is in the form of a container for storage, transportation, and disposal of radioactive waste. In addition, improved methods for preparing uranium dioxide and uranium carbide microspheres for use in the radiation shielding materials are also provided.

Mirsky, Steven M.; Krill, Stephen J.; and Murray, Alexander P.

2005-11-01T23:59:59.000Z

173

Radiation Shielding Materials and Containers Incorporating Same  

DOE Patents (OSTI)

An improved radiation shielding material and storage systems for radioactive materials incorporating the same. The PYRolytic Uranium Compound (''PYRUC'') shielding material is preferably formed by heat and/or pressure treatment of a precursor material comprising microspheres of a uranium compound, such as uranium dioxide or uranium carbide, and a suitable binder. The PYRUC shielding material provides improved radiation shielding, thermal characteristic, cost and ease of use in comparison with other shielding materials. The shielding material can be used to form containment systems, container vessels, shielding structures, and containment storage areas, all of which can be used to house radioactive waste. The preferred shielding system is in the form of a container for storage, transportation, and disposal of radioactive waste. In addition, improved methods for preparing uranium dioxide and uranium carbide microspheres for use in the radiation shielding materials are also provided.

Mirsky, Steven M.; Krill, Stephen J.; and Murray, Alexander P.

2005-11-01T23:59:59.000Z

174

Program plan for the DOE Office of Fusion Energy First Wall/Blanket/Shield Engineering Technology Program. Volume I. Summary, objectives and management. Revision 2  

SciTech Connect

This document defines a plan for conducting selected aspects of the engineering testing required for magnetic fusion reactor FWBS components and systems. The ultimate product of this program is an established data base that contributes to a functional, reliable, maintainable, economically attractive, and environmentally acceptable commercial fusion reactor first wall, blanket, and shield system. This program plan updates the initial plan issued in November of 1980 by the DOE/Office of Fusion Energy (unnumbered report). The plan consists of two parts. Part I is a summary of activities, responsibilities and program management including reporting and interfaces with other programs. Part II is a compilation of the Detailed Technical Plans for Phase I (1982 to 1984) developed by the participants during Phase 0 of the program (July to December 1981).

Not Available

1982-08-01T23:59:59.000Z

175

Shielding vacuum fluctuations with graphene  

E-Print Network (OSTI)

The Casimir-Polder interaction of ground-state and excited atoms with graphene is investigated with the aim to establish whether graphene systems can be used as a shield for vacuum fluctuations of an underlying substrate. We calculate the zero-temperature Casimir-Polder potential from the reflection coefficients of graphene within the framework of the Dirac model. For both doped and undoped graphene we show limits at which graphene could be used effectively as a shield. Additional results are given for AB-stacked bilayer graphene.

Sofia Ribeiro; Stefan Scheel

2013-10-22T23:59:59.000Z

176

Shielding vacuum fluctuations with graphene  

E-Print Network (OSTI)

The Casimir-Polder interaction of ground-state and excited atoms with graphene is investigated with the aim to establish whether graphene systems can be used as a shield for vacuum fluctuations of an underlying substrate. We calculate the zero-temperature Casimir-Polder potential from the reflection coefficients of graphene within the framework of the Dirac model. For both doped and undoped graphene we show limits at which graphene could be used effectively as a shield. Additional results are given for AB-stacked bilayer graphene.

Ribeiro, Sofia

2013-01-01T23:59:59.000Z

177

Incorporation of Hydride Nuclear Fuels in Commercial Light Water Reactors  

E-Print Network (OSTI)

Fundamental aspects of nuclear reactor fuel elements.Unlike permanent nuclear reactor core components, nuclearof the first nuclear reactors, commercial nuclear fuel still

Terrani, Kurt Amir

2010-01-01T23:59:59.000Z

178

Composite materials for electromagnetic shielding  

Science Conference Proceedings (OSTI)

The paper shows up the research results on processing and characterization of composite materials with polymeric matrix (silicone rubber). The materials obtained in laboratory contain metallized nettling like reinforcement material and powdery graphite ... Keywords: attenuation, filling additions, frequency, plated nettling, polymeric composite on de basis of silicone rubber, shielding effectiveness

Stoian Elena Valentina; Rizescu Cristiana; Iordache Iulian; Ionita Gheorghe; Bacinschi Zorica

2010-07-01T23:59:59.000Z

179

Innovative Approaches to Large Component Packaging  

Science Conference Proceedings (OSTI)

Radioactive waste disposal often times requires creative approaches in packaging design, especially for large components. Innovative design techniques are required to meet the needs for handling, transporting, and disposing of these large packages. Large components (i.e., Reactor Pressure Vessel (RPV) heads and even RPVs themselves) require special packaging for shielding and contamination control, as well as for transport and disposal. WMG Inc designed and used standard packaging for RPV heads without control rod drive mechanisms (CRDMs) attached for five RPV heads and has also more recently met an even bigger challenge and developed the innovative Intact Vessel Head Transport System (IVHTS) for RPV heads with CRDMs intact. This packaging system has been given a manufacturer's exemption by the United States Department of Transportation (USDOT) for packaging RPV heads. The IVHTS packaging has now been successfully used at two commercial nuclear power plants. Another example of innovative packaging is the large component packaging that WMG designed, fabricated, and utilized at the West Valley Demonstration Project (WVDP). In 2002, West Valley's high-level waste vitrification process was shut down in preparation for D and D of the West Valley Vitrification Facility. Three of the major components of concern within the Vitrification Facility were the Melter, the Concentrate Feed Makeup Tank (CFMT), and the Melter Feed Holdup Tank (MFHT). The removal, packaging, and disposition of these three components presented significant radiological and handling challenges for the project. WMG designed, fabricated, and installed special packaging for the transport and disposal of each of these three components, which eliminated an otherwise time intensive and costly segmentation process that WVDP was considering. Finally, WMG has also designed and fabricated special packaging for both the Connecticut Yankee (CY) and San Onofre Nuclear Generating Station (SONGS) RPVs. This paper presents the approach that has been successfully used for planning, implementing, and preparing for the disposition of large components such as those mentioned previously. It addresses the major regulatory and design requirements for packaging, transporting, and disposing of these components. The specific topics that are covered include radiological characterization, shielding, packaging design, on-site handling and movement, off-site transportation options, a brief discussion on disposition, and lessons learned. (authors)

Freitag, A.; Hooper, M.; Posivak, E.; Sullivan, J. [WMG, Inc., Peekskill, NY 10566 (United States)

2006-07-01T23:59:59.000Z

180

Residents and windows. 1. Shielding of windows  

SciTech Connect

In order to assess the influence of the shielding of windows performed by occupants in residential buildings on the heat balance of the building, the shielding of 40,000 windows was determined by observation during two heating seasons. It is shown that the demand for privacy has a large effect on the degree of window-shielding. There are also indications that many occupants trying to save energy use window-shielding as one of their means to achieve this.

Lyrberg, M.D.

1983-06-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor components shield" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


181

Special concrete shield selection using the analytic hierarchy process  

SciTech Connect

Special types of concrete radiation shields that depend on locally available materials and have improved properties for both neutron and gamma-ray attenuation were developed by using plastic materials and heavy ores. The analytic hierarchy process (AHP) is implemented to evaluate these types for selecting the best biological radiation shield for nuclear reactors. Factors affecting the selection decision are degree of protection against neutrons, degree of protection against gamma rays, suitability of the concrete as building material, and economic considerations. The seven concrete alternatives are barite-polyethylene concrete, barite-polyvinyl chloride (PVC) concrete, barite-portland cement concrete, pyrite-polyethylene concrete, pyrite-PVC concrete, pyrite-portland cement concrete, and ordinary concrete. The AHP analysis shows the superiority of pyrite-polyethylene concrete over the others.

Abulfaraj, W.H. (King Abdulaziz Univ., Jeddah (Saudi Arabia). Nuclear Engineering Dept.)

1994-08-01T23:59:59.000Z

182

Catalytic reactor  

DOE Patents (OSTI)

A catalytic reactor is provided with one or more reaction zones each formed of set(s) of reaction tubes containing a catalyst to promote chemical reaction within a feed stream. The reaction tubes are of helical configuration and are arranged in a substantially coaxial relationship to form a coil-like structure. Heat exchangers and steam generators can be formed by similar tube arrangements. In such manner, the reaction zone(s) and hence, the reactor is compact and the pressure drop through components is minimized. The resultant compact form has improved heat transfer characteristics and is far easier to thermally insulate than prior art compact reactor designs. Various chemical reactions are contemplated within such coil-like structures such that as steam methane reforming followed by water-gas shift. The coil-like structures can be housed within annular chambers of a cylindrical housing that also provide flow paths for various heat exchange fluids to heat and cool components.

Aaron, Timothy Mark (East Amherst, NY); Shah, Minish Mahendra (East Amherst, NY); Jibb, Richard John (Amherst, NY)

2009-03-10T23:59:59.000Z

183

Nuclear processes in magnetic fusion reactors with polarized fuel  

E-Print Network (OSTI)

We consider the processes $d +d \\to n +{^3He}$, $d +{^3He} \\to p +{^4He}$, $d +{^3H} \\to n +{^4He}$, ${^3He} +{^3He}\\to p+p +{^4He}$, ${^3H} +{^3He}\\to d +{^4He}$, with particular attention for applications in fusion reactors. After a model independent parametrization of the spin structure of the matrix elements for these processes at thermal colliding energies, in terms of partial amplitudes, we study polarization phenomena in the framework of a formalism of helicity amplitudes. The strong angular dependence of the final nuclei and of the polarization observables on the polarizations of the fuel components can be helpful in the design of the reactor shielding, blanket arrangement etc..We analyze also the angular dependence of the neutron polarization for the processes $\\vec d +\\vec d \\to n +{^3He}$ and $\\vec d +\\vec {^3H} \\to n +{^4He}$.

Rekalo, M P; Rekalo, Michail P.; Tomasi-Gustafsson, Egle

2000-01-01T23:59:59.000Z

184

Nuclear processes in magnetic fusion reactors with polarized fuel  

E-Print Network (OSTI)

We consider the processes $d +d \\to n +{^3He}$, $d +{^3He} \\to p +{^4He}$, $d +{^3H} \\to n +{^4He}$, ${^3He} +{^3He}\\to p+p +{^4He}$, ${^3H} +{^3He}\\to d +{^4He}$, with particular attention for applications in fusion reactors. After a model independent parametrization of the spin structure of the matrix elements for these processes at thermal colliding energies, in terms of partial amplitudes, we study polarization phenomena in the framework of a formalism of helicity amplitudes. The strong angular dependence of the final nuclei and of the polarization observables on the polarizations of the fuel components can be helpful in the design of the reactor shielding, blanket arrangement etc..We analyze also the angular dependence of the neutron polarization for the processes $\\vec d +\\vec d \\to n +{^3He}$ and $\\vec d +\\vec {^3H} \\to n +{^4He}$.

Michail P. Rekalo; Egle Tomasi-Gustafsson

2000-10-16T23:59:59.000Z

185

Brookhaven Graphite Research Reactor | Environmental Restoration Projects |  

NLE Websites -- All DOE Office Websites (Extended Search)

Why Was the BGRR Decommissioned? Why Was the BGRR Decommissioned? BGRR The Brookhaven Graphite Research Reactor (BGRR) at Brookhaven National Laboratory (BNL) was decommissioned to ensure the complex is in a safe and stable condition and to reduce sources of groundwater contamination. The BGRR contained over 8,000 Curies of radioactive contaminants from past operations consisting of primarily nuclear activation products such as hydrogen-3 (tritium) and carbon-14 and fission products cesium-137 and strontium-90. The nature and extent of contamination varied by location depending on historic uses of the systems and components and releases, however, the majority of the contamination (over 99 percent) was bound within the graphite pile and biological shield. Radioactive contamination was identified in the fuel handling system deep

186

Fusion reactor design studies. [ARIES Tokamak  

SciTech Connect

This report discusses the following topics on the ARIES tokamak: systems; plasma power balance; impurity control and fusion ash removal; fusion product ripple loss; energy conversion; reactor fueling; first wall design; shield design; reactor safety; and fuel cost and resources. (LSP)

Emmert, G.A.; Kulcinski, G.L.; Santarius, J.F.

1990-10-12T23:59:59.000Z

187

Spectral Emissivity Measurements of High Temperature Reactor ...  

Science Conference Proceedings (OSTI)

CASL: The Consortium for Advanced Simulation of Light Water Reactors: A U.S. ... Strategies for Studying High Dose Irradiation Effects in Reactor Components.

188

Materials Degradation Issues in Pressurized Water Reactors  

Science Conference Proceedings (OSTI)

CASL: The Consortium for Advanced Simulation of Light Water Reactors: A U.S. ... Strategies for Studying High Dose Irradiation Effects in Reactor Components.

189

THERMAL NUCLEAR REACTOR  

DOE Patents (OSTI)

Nuclear reactors of the graphite moderated air cooled type in which canned slugs or rods of fissile material are employed are discussed. Such a reactor may be provided with a means for detecting dust particles in the exhausted air. The means employed are lengths of dust absorbent cord suspended in vertical holes in the shielding structure above each vertical coolant flow channel to hang in the path of the cooling air issuing from the channels, and associated spindles and drive motors for hauling the cords past detectors, such as Geiger counters, for inspecting the cords periodically. This design also enables detecting the individual channel in which a fault condition may have occurred.

Fenning, F.W.; Jackson, R.F.

1957-09-24T23:59:59.000Z

190

SETTABLE NEUTRON RADIATION SHIELDING MATERIAL  

DOE Patents (OSTI)

A settable, viscous, putty-like shielding composition is described. It consists of an intimate admixture of a major proportion of a compound having a ratio of hydrogen atoms to all other atoms therein within the range of from 0.5: 1 to 2:l. from 0.5 to 10% by weight of boron, and a fluid resinous carrier This composition when cured is adapted to attenuate fast moving neutrons and capture slow moving neutrons.

Axelrad, I.R.

1960-11-22T23:59:59.000Z

191

Nuclear component horizontal seismic restraint  

DOE Patents (OSTI)

A nuclear component horizontal seismic restraint. Small gaps limit horizontal displacement of components during a seismic occurrence and therefore reduce dynamic loadings on the free lower end. The reactor vessel and reactor guard vessel use thicker section roll-forged rings welded between the vessel straight shell sections and the bottom hemispherical head sections. The inside of the reactor guard vessel ring forging contains local vertical dovetail slots and upper ledge pockets to mount and retain field fitted and installed blocks. As an option, the horizontal displacement of the reactor vessel core support cone can be limited by including shop fitted/installed local blocks in opposing alignment with the reactor vessel forged ring. Beams embedded in the wall of the reactor building protrude into apertures in the thermal insulation shell adjacent the reactor guard vessel ring and have motion limit blocks attached thereto to provide to a predetermined clearance between the blocks and reactor guard vessel ring.

Snyder, Glenn J. (Lynchburg, VA)

1988-01-01T23:59:59.000Z

192

VAPOR SHIELD FOR INDUCTION FURNACE  

DOE Patents (OSTI)

This patent relates to a water-cooled vapor shield for an inductlon furnace that will condense metallic vapors arising from the crucible and thus prevent their condensation on or near the induction coils, thereby eliminating possible corrosion or shorting out of the coils. This is accomplished by placing, about the top, of the crucible a disk, apron, and cooling jacket that separates the area of the coils from the interior of the cruclbIe and provides a cooled surface upon whlch the vapors may condense.

Reese, S.L.; Samoriga, S.A.

1958-03-11T23:59:59.000Z

193

Shielding measurements for a 230 MeV proton beam  

SciTech Connect

Energetic secondary neutrons produced as protons interact with accelerator components and patients dominate the radiation shielding environment for proton radiotherapy facilities. Due to the scarcity of data describing neutron production, attenuation, absorbed dose, and dose equivalent values, these parameters were measured for 230 MeV proton bombardment of stopping length Al, Fe, and Pb targets at emission angles of 0{degree}, 22{degree}, 45{degree}, and 90{degree} in a thick concrete shield. Low pressure tissue-equivalent proportional counters with volumes ranging from 1 cm{sup 3} to 1000 cm{sup 3} were used to obtain microdosimetric spectra from which absorbed dose and radiation quality are deduced. Does equivalent values and attenuation lengths determined at depth in the shield were found to vary sharply with angle, but were found to be independent of target material. Neutron dose and radiation length values are compared with Monte Carlo neutron transport calculations performed using the Los Alamos High Energy Transport Code (LAHET). Calculations used 230 MeV protons incident upon an Fe target in a shielding geometry similar to that used in the experiment. LAHET calculations overestimated measured attenuation values at 0{degree}, 22{degree}, and 45{degree}, yet correctly predicted the attenuation length at 90{degree}. Comparison of the mean radiation quality estimated with the Monte Carlo calculations with measurements suggest that neutron quality factors should be increased by a factor of 1.4. These results are useful for the shielding design of new facilities as well as for testing neutron production and transport calculations.

Siebers, J.V.

1990-01-01T23:59:59.000Z

194

Hot Cell Window Shielding Analysis Using MCNP  

SciTech Connect

The Idaho National Laboratory Materials and Fuels Complex nuclear facilities are undergoing a documented safety analysis upgrade. In conjunction with the upgrade effort, shielding analysis of the Fuel Conditioning Facility (FCF) hot cell windows has been conducted. This paper describes the shielding analysis methodology. Each 4-ft thick window uses nine glass slabs, an oil film between the slabs, numerous steel plates, and packed lead wool. Operations in the hot cell center on used nuclear fuel (UNF) processing. Prior to the shielding analysis, shield testing with a gamma ray source was conducted, and the windows were found to be very effective gamma shields. Despite these results, because the glass contained significant amounts of lead and little neutron absorbing material, some doubt lingered regarding the effectiveness of the windows in neutron shielding situations, such as during an accidental criticality. MCNP was selected as an analysis tool because it could model complicated geometry, and it could track gamma and neutron radiation. A bounding criticality source was developed based on the composition of the UNF. Additionally, a bounding gamma source was developed based on the fission product content of the UNF. Modeling the windows required field inspections and detailed examination of drawings and material specifications. Consistent with the shield testing results, MCNP results demonstrated that the shielding was very effective with respect to gamma radiation, and in addition, the analysis demonstrated that the shielding was also very effective during an accidental criticality.

Chad L. Pope; Wade W. Scates; J. Todd Taylor

2009-05-01T23:59:59.000Z

195

Waste Package Component Design Methodology Report  

Science Conference Proceedings (OSTI)

This Executive Summary provides an overview of the methodology being used by the Yucca Mountain Project (YMP) to design waste packages and ancillary components. This summary information is intended for readers with general interest, but also provides technical readers a general framework surrounding a variety of technical details provided in the main body of the report. The purpose of this report is to document and ensure appropriate design methods are used in the design of waste packages and ancillary components (the drip shields and emplacement pallets). The methodology includes identification of necessary design inputs, justification of design assumptions, and use of appropriate analysis methods, and computational tools. This design work is subject to ''Quality Assurance Requirements and Description''. The document is primarily intended for internal use and technical guidance for a variety of design activities. It is recognized that a wide audience including project management, the U.S. Department of Energy (DOE), the U.S. Nuclear Regulatory Commission, and others are interested to various levels of detail in the design methods and therefore covers a wide range of topics at varying levels of detail. Due to the preliminary nature of the design, readers can expect to encounter varied levels of detail in the body of the report. It is expected that technical information used as input to design documents will be verified and taken from the latest versions of reference sources given herein. This revision of the methodology report has evolved with changes in the waste package, drip shield, and emplacement pallet designs over many years and may be further revised as the design is finalized. Different components and analyses are at different stages of development. Some parts of the report are detailed, while other less detailed parts are likely to undergo further refinement. The design methodology is intended to provide designs that satisfy the safety and operational requirements of the YMP. Four waste package configurations have been selected to illustrate the application of the methodology during the licensing process. These four configurations are the 21-pressurized water reactor absorber plate waste package (21-PWRAP), the 44-boiling water reactor waste package (44-BWR), the 5 defense high-level radioactive waste (HLW) DOE spent nuclear fuel (SNF) codisposal short waste package (5-DHLWDOE SNF Short), and the naval canistered SNF long waste package (Naval SNF Long). Design work for the other six waste packages will be completed at a later date using the same design methodology. These include the 24-boiling water reactor waste package (24-BWR), the 21-pressurized water reactor control rod waste package (21-PWRCR), the 12-pressurized water reactor waste package (12-PWR), the 5 defense HLW DOE SNF codisposal long waste package (5-DHLWDOE SNF Long), the 2 defense HLW DOE SNF codisposal waste package (2-MC012-DHLW), and the naval canistered SNF short waste package (Naval SNF Short). This report is only part of the complete design description. Other reports related to the design include the design reports, the waste package system description documents, manufacturing specifications, and numerous documents for the many detailed calculations. The relationships between this report and other design documents are shown in Figure 1.

D.C. Mecham

2004-07-12T23:59:59.000Z

196

An Evaluation of Shadow Shielding for Lunar System Waste Heat Rejection  

E-Print Network (OSTI)

Shadow shielding is a novel and practical concept for waste heat rejection from lunar surface spacecraft systems. A shadow shield is a light shield that shades the radiator from parasitic thermal radiation emanating from the sun or lunar surface. Radiator size and mass can reduce if the radiator is not required to account for parasitic heat loads in addition to system energy rejection requirements. The lunar thermal environment can be very harsh towards radiative heat rejection. Parasitic heat loads force the radiator to expand in size and mass to compensate. On the Moon, there are three types: surface infrared, solar insulation, and albedo. This thesis tests shadow shielding geometry and its effect on the radiator and nuclear reactor in a reactor-powered Carnot heat engine. Due to the nature of cooling by radiative heat transfer, the maximum shaft work a Carnot system can produce and the minimal required radiator area occurs when the Carnot efficiency is 25%. First, a case for shadow shielding is made using an isothermal, control radiator model in Thermal Desktop. Six radiator temperatures and three latitudes are considered in the tests. Test variables in this section include radiator shapes and shade geometry. The simulations found that shadow shielding is best suited for a low-temperature radiator at the lunar equator. Optimized parabolic shade geometry includes a focus right above or at the top of the radiator and full to three-quarters shade height. The most useful rectangular radiator shape for shadow shielding is that which has a low height and long width. All simulations were conducted using a shade with a 10 kg/m2 area mass. A sensitivity study was conducted for different shade area masses using high and low values found in the literature. The shade is the most useful when the shade's area mass is less than or equal to that of the radiator. If the shade mass is below this threshold, the shade would be applicable to all radiator temperatures tested. Optimized shade and radiator geometry results were then factored into a second model where the radiator is comprised of heat pipes which is similar to radiators from actual system designs. Further simulations were conducted implementing the SAFE-4001 fast fission nuclear reactor design. The study found that shadow shielding allowed the system to use a low-temperature radiator where other configurations were not viable because shadow shielding drastically improves radiative heat transfer from the radiator, but at the consequence of raising radiator mass.

Worn, Cheyn

2012-05-01T23:59:59.000Z

197

Atlas SCT/Pixel Grounding and Shielding ATLAS SCT/Pixel Grounding and Shielding Note  

E-Print Network (OSTI)

cylinder with heater tapes 600 CFRP (carbon fiber reinforced polymer) Castellated thermal barrier Schematic This would provide data temperature variation resistance TEMP1 path. DG_SENSE would loaded high) Conductors outside shield (6) Temperature monitoring The Module Array Shield The main purpose shield provide

California at Santa Cruz, University of

198

CORROSION OF LEAD SHIELDING IN NUCLEAR MATERIALS PACKAGES  

SciTech Connect

Inspection of United States-Department of Energy (US-DOE) model 9975 nuclear materials shipping package revealed corrosion of the lead shielding that was induced by off-gas constituents from organic components in the package. Experiments were performed to determine the corrosion rate of lead when exposed to off-gas or degradation products of these organic materials. The results showed that the room temperature vulcanizing (RTV) sealant was the most corrosive organic species used in the construction of the packaging, followed by polyvinyl acetate (PVAc) glue. Fiberboard material, also used in the construction of the packaging induced corrosion to a much lesser extent than the PVAc glue and RTV sealant, and only in the presence of condensed water. The results indicated faster corrosion at temperatures higher than ambient and with condensed water. In light of these corrosion mechanisms, the lead shielding was sheathed in a stainless steel liner to mitigate against corrosion.

Subramanian, K; Kerry Dunn, K; Joseph Murphy, J

2008-07-18T23:59:59.000Z

199

Thermal neutron shield and method of manufacture  

DOE Patents (OSTI)

A thermal neutron shield comprising concrete with a high percentage of the element Boron. The concrete is least 54% Boron by weight which maximizes the effectiveness of the shielding against thermal neutrons. The accompanying method discloses the manufacture of Boron loaded concrete which includes enriching the concrete mixture with varying grit sizes of Boron Carbide.

Brindza, Paul Daniel; Metzger, Bert Clayton

2013-05-28T23:59:59.000Z

200

Memorandum on Chemical Reactors and Reactor Hazards  

SciTech Connect

Two important problems in the investigation of reactor hazards are the chemical reactivity of various materials employed in reactor construction and the chracteristics of heat transfer under transient conditions, specifically heat transfer when driven by an exponentially increasing heat source (exp t/T). Although these problems are independent of each other, when studied in relation to reactor hazards they may occur in a closely coupled sequence. For example the onset of a dangerous chemical reactor may be due to structural failure of various reactor components under an exponentially rising heat source originating with a runaway nuclear reactor. For this reason, these two problems should eventually be studied together after an exploratory experimental survey has been made in which they are considered separately.

Mills, M.M.; Pearlman, H.; Ruebsamen, W.; Steele, G., Chrisney, J.

1951-07-05T23:59:59.000Z

Note: This page contains sample records for the topic "reactor components shield" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


201

Shielding calculations at dismantled synchrocyclotron  

SciTech Connect

The Space Radiation Effects Laboratory located in Newport News, Virginia, was operated by the College of William and Mary for the National Aeronautics and Space Administration. A synchrocyclotron which was formerly in operation in this building was removed in 1980. At several locations, the scattered radiation caused an induced radioactivity within the walls of the cyclotron room. A radiological survey has been performed to determine the amount of residual radioactivity on the walls. Calculations were performed to determine the thickness of the concrete walls and floor for shielding the residual radiation in the cyclotron room. Recommendations are made to minimize exposures from the residual radioactivity on the walls and floor of the cyclotron room to potential occupants working in the building. 19 refs., 1 fig., 2 tabs.

Yalcintas, M.G.

1987-01-01T23:59:59.000Z

202

Design considerations in inertially-confined fusion reactors  

SciTech Connect

This paper discusses the effects of short time pulses of energetic particles and waves typical of inertially-confined thermonuclear reactions on the first wall, blanket and shield of conceptual reactors. Several reactor designs are presented which attempt to cope with the various problems from the microexplosion debris. Fusion-fission hybrid reactors are also discussed. Emphasis is placed on the first-wall problems of laser-initiated, inertially confined fusion reactors using the deuterium-tritium fuel cycle.

Hovingh, J.

1976-08-01T23:59:59.000Z

203

Reactor Materials Program - Baseline Material Property Handbook - Mechanical Properties of 1950's Vintage Stainless Steel Weldment Components, Task Number 89-23-A-1  

SciTech Connect

The Process Water System (primary coolant) piping of the nuclear production reactors constructed in the 1950''s at Savannah River Site is comprised primarily of Type 304 stainless steel with Type 308 stainless steel weld filler. A program to measure the mechanical properties of archival PWS piping and weld materials (having approximately six years of service at temperatures between 25 and 100 degrees C) has been completed. The results from the mechanical testing has been synthesized to provide a mechanical properties database for structural analyses of the SRS piping.

Stoner, K.J.

1999-11-05T23:59:59.000Z

204

Shield Design of the Materials Test Station's Camera Room  

Science Conference Proceedings (OSTI)

Shielding / Special Issue on the 11th International Conference on Radiation Shielding and the 15th Topical Meeting of the Radiation Protection and Shielding Division (Part 2) / Radiation Protection

G. Muhrer; M. Wilson; Ch. Kelsey; E. Pitcher; F. Gallmeier; M. Wohlmuther

205

Radiation Shielding of a Beta-Beam Rapid Cycling Synchrotron  

Science Conference Proceedings (OSTI)

Shielding / Special Issue on the 11th International Conference on Radiation Shielding and the 15th Topical Meeting of the Radiation Protection and Shielding Division (Part 2) / Radiation Protection

Stefania Trovati; Matteo Magistris; Marco Silari

206

Transmission of Shielding Materials for Particle Therapy Facilities  

Science Conference Proceedings (OSTI)

Shielding Materials / Special Issue on the 11th International Conference on Radiation Shielding and the 15th Topical Meeting of the Radiation Protection and Shielding Division (Part 2) / Radiation Protection

Nisy E. Ipe

207

ACDOS1: a computer code to calculate dose rates from neutron activation of neutral beamlines and other fusion-reactor components  

Science Conference Proceedings (OSTI)

A computer code has been written to calculate neutron induced activation of neutral-beam injector components and the corresponding dose rates as a function of geometry, component composition, and time after shutdown. The code, ACDOS1, was written in FORTRAN IV to calculate both activity and dose rates for up to 30 target nuclides and 50 neutron groups. Sufficient versatility has also been incorporated into the code to make it applicable to a variety of general activation problems due to neutrons of energy less than 20 MeV.

Keney, G.S.

1981-08-01T23:59:59.000Z

208

Gamma and X-ray shielding compositions utilizing bauxite - Red Mud regional research laboratory (CSIR), Bhopal, India  

SciTech Connect

Available in abstract form only. Full text of publication follows: The application spectrum of X-ray and Gamma radiation is increasing exponentially in the area of diagnostic, nuclear medicine, food preservation, nuclear power plants and strategic utilities. To prevent the harmful effects of these radiations, shielding materials based on lead metal and its compounds are being used historically, which are toxic in nature. To protect environment it has become necessary to develop non-toxic lead free shielding materials. The use of titanium metal and its compounds as synthetic rock i.e. SYNROC are reported to be very effective non-toxic shielding materials for various applications. Red mud waste generated in aluminum producing industries possesses a unique mineralogical compositions containing fairly high quantity of titanium oxide and iron oxide useful for making non toxic shielding compositions and therefore red mud has been utilized for the first time in the world for making radiation shielding materials. The red mud based compositions developed have been characterized for their various physico-mechanical properties namely compressive strength, impact strength, density and X-ray and gamma radiation shielding capacity in terms of shielding thickness i.e. HVT. Based on the characterization results it is found that the red mud based materials can be used for the construction of X-ray diagnostic and CT-Scanner room and as a substitute shielding material for concrete in the nuclear reactors and other radiation based applications. Studies on the identification of shielding phases and their morphology present, in the red mud based shielding compositions has been carried out using X-ray diffraction and SEM technique. The results of these studies are presented in this paper. (authors)

Anshul, Avneesh; Amritphale, Sudhir Sitaram; Chandra, Navin; Ramakrishnan, N. [Regional Research Laboratory, Council of Scientific and Industrial Research, Bhubaneswar 751013, Orissa (India)

2007-07-01T23:59:59.000Z

209

Drip shield Structural Response to Rock Fall  

Science Conference Proceedings (OSTI)

The objective of this calculation is to determine areas over the drip shield (DS) top plate and side-walls where the residual stress values exceed 50% of Ti-7 yield strength. These areas will also be referred to as the damaged areas throughout this document. The scope of this document is limited to reporting the calculation results in terms of the damaged areas based on a chosen set of stress components. This calculation is intended for use in support of the preliminary design activities for the license application design of the DS. This calculation is associated with the DS design and was performed by the Waste Package and Components. AP-3.12Q, ''Design Calculations and Analyses'' is used to perform the calculation and develop the document. The DS is classified as a safety category item. Therefore, this calculation is subject to the Quality Assurance Requirements and Description. The information provided by the sketches attached to this calculation is that of the potential design of the type of DS considered in this calculation and provides the potential dimensions and materials for the DS design. The finite element (FE) calculation was performed by using the commercially available LS-DYNA Version (V)960 (Software Tracking Number [STN] 10300-960.1106-00, Ref. 7) FE code. The results of this calculation were evaluated using residual first principal stress. Subsequent analysis of areas determined by residual stresses have been reported in the results section of this document. The finite element mesh adequacy was determined based on the maximum stress intensity and maximum first principal stress. The current work processes and procedures for the control of the electronic management of data for this activity were conducted in accordance with AP-3.13Q, ''Design Control'' (Section 5.1.2).

Z. Ceylan

2004-04-01T23:59:59.000Z

210

Evolution of Shielding Computations for the ITER Upper ECH Launcher  

Science Conference Proceedings (OSTI)

Technical Paper / Special Issue on the 16th Biennial Topical Meeting of the Radiation Protection and Shielding Division / Photon and Neutron Transport and Shielding (DETERMINISTIC or Mc)

A. Serikov; U. Fischer; D. Grosse; P. Spaeh; D. Strauss

211

Hot cell shield plug extraction apparatus  

DOE Patents (OSTI)

An apparatus is provided for moving shielding plugs into and out of holes in concrete shielding walls in hot cells for handling radioactive materials without the use of external moving equipment. The apparatus provides a means whereby a shield plug is extracted from its hole and then swung approximately 90 degrees out of the way so that the hole may be accessed. The apparatus uses hinges to slide the plug in and out and to rotate it out of the way, the hinge apparatus also supporting the weight of the plug in all positions, with the load of the plug being transferred to a vertical wall by means of a bolting arrangement.

Knapp, Philip A. (Moore, ID); Manhart, Larry K. (Pingree, ID)

1995-01-01T23:59:59.000Z

212

ADVANTG Shielding Analysis for Closure Operations in an Open-Mode Repository  

Science Conference Proceedings (OSTI)

en-mode repository concepts could require worker entry into access drifts after placement of fuel casks in order to perform activities related to backfill, plug emplacement, routine maintenance, or performance confirmation. An ideal emplacement-drift shielding configuration would minimize dose to workers while maximizing airflow through the emplacement drifts. This paper presents a preliminary investigation of the feasibility and effectiveness of radiation shielding concepts that could be employed to facilitate worker operations in an open-mode repository. The repository model for this study includes pressurized-water reactor fuel assemblies (60 GWd/MTU burnup, 40 year post-irradiation cooldown) in packages of 32 assemblies. The closest fuel packages are 5 meters from dosimetry voxels in the access drift. The unshielded dose to workers in the access drift is 73.7 rem/hour. Prior work suggests that open-mode repository concepts similar to this one would require 15 m3/s of ventilation airflow. Shielding concepts considered here include partial concrete plugs, labyrinthine shields, and stainless steel photon attenuator grids. Maximum dose to workers in the access drift was estimated for each shielding concept using MCNP5 with variance reduction parameters generated by ADVANTG. Because airflow through the shielding is important for open-mode repositories, a semi-empirical estimate of the head loss due to each shielding configuration was also calculated. Airflow and shielding performance vary widely among the proposed shielding configurations. Although the partial plug configuration had the best airflow performance, it allowed dose rates 1500 greater than the specified target. Labyrinthine shielding concepts yield doses on the order of 1 mrem/hour with configurations that impose 3 to 11 J/kg head loss. Adding 1 cm lead lining to the airflow channels of labyrinthine designs further reduces the worker dose by 65% to 95%. Photon-attenuator concepts may reduce worker dose to as low as 29 mrem/hour with head loss on the order of 1.9 J/kg.

Bevill, Aaron M [ORNL; Radulescu, Georgeta [ORNL; Scaglione, John M [ORNL; Howard, Rob L [ORNL

2013-01-01T23:59:59.000Z

213

Deep Space Mission Radiation Shielding Optimization  

Science Conference Proceedings (OSTI)

Providing protection against the hazards of space radiation is a major challenge to the exploration and development of space. The great cost of added radiation shielding is a potential limiting factor in deep space missions. In the present report, we ...

Tripathi R. K.; Wilson J. W.; Cucinotta F. A.; Nealy J. E.; Clowdsley M. S.; Kim M-H. Y.

2001-07-01T23:59:59.000Z

214

Minimizing Errors Associated with Multiplate Radiation Shields  

Science Conference Proceedings (OSTI)

Multiplate radiation shield errors are examined using the following techniques: 1) ray tracing analysis, 2) wind tunnel experiments, 3) numerical flow simulations, and 4) field testing. The authors’ objectives are to develop guidelines for ...

Scott J. Richardson; Fred V. Brock; Steven R. Semmer; Cathy Jirak

1999-11-01T23:59:59.000Z

215

Reliability Methods for Shield Design Process  

Science Conference Proceedings (OSTI)

Providing protection against the hazards of space radiation is a major challenge to the exploration and development of space. The great cost of added radiation shielding is a potential limiting factor in deep space operations. In this enabling technology

R. K. Tripathi; J. W. Wilson

2003-01-01T23:59:59.000Z

216

Instability of Shielded Surface Temperature Vortices  

Science Conference Proceedings (OSTI)

The stability characteristics of the surface quasigeostrophic shielded Rankine vortex are found using a linearized contour dynamics model. Both the normal modes and nonmodal evolution of the system are analyzed and the results are compared with ...

Benjamin J. Harvey; Maarten H. P. Ambaum; Xavier J. Carton

2011-05-01T23:59:59.000Z

217

Thermomagnetic burn control for magnetic fusion reactor  

DOE Patents (OSTI)

Apparatus is provided for controlling the plasma energy production rate of a magnetic-confinement fusion reactor, by controlling the magnetic field ripple. The apparatus includes a group of shield sectors (30a, 30b, etc.) formed of ferromagnetic material which has a temperature-dependent saturation magnetization, with each shield lying between the plasma (12) and a toroidal field coil (18). A mechanism (60) for controlling the temperature of the magnetic shields, as by controlling the flow of cooling water therethrough, thereby controls the saturation magnetization of the shields and therefore the amount of ripple in the magnetic field that confines the plasma, to thereby control the amount of heat loss from the plasma. This heat loss in turn determines the plasma state and thus the rate of energy production.

Rawls, John M. (Del Mar, CA); Peuron, Unto A. (Solana Beach, CA)

1982-01-01T23:59:59.000Z

218

Thermomagnetic burn control for magnetic fusion reactor  

DOE Patents (OSTI)

Apparatus is provided for controlling the plasma energy production rate of a magnetic-confinement fusion reactor, by controlling the magnetic field ripple. The apparatus includes a group of shield sectors formed of ferromagnetic material which has a temperature-dependent saturation magnetization, with each shield lying between the plasma and a toroidal field coil. A mechanism for controlling the temperature of the magnetic shields, as by controlling the flow of cooling water therethrough, thereby controls the saturation magnetization of the shields and therefore the amount of ripple in the magnetic field that confines the plasma, to thereby control the amount of heat loss from the plasma. This heat loss in turn determines the plasma state and thus the rate of energy production.

Rawls, J.M.; Peuron, A.U.

1980-07-01T23:59:59.000Z

219

Formal derivation of optimal active shielding for low-power on-chip buses  

Science Conference Proceedings (OSTI)

Passive shielding has been used to reduce capacitive coupling effects of adjacent bus lines by inserting passive ground or power lines (shields) between the bus lines. Active shielding is another shielding technique, in which the shield is allowed to ...

M. Ghonemia; Y. Ismail

2004-11-01T23:59:59.000Z

220

Phase Conductor and Shield Wire Corrosion  

Science Conference Proceedings (OSTI)

As new inspection technologies for conductors and shield wires develop and mature, the assessment process becomes more accurate in identifying and quantifying flaws. With better accuracy, condition assessment can be less conservative, thus allowing utilities to save money by keeping conductors in service for longer periods. This report, which includes the results of a survey of EPRI member utilities on phase conductor and shield wire installations, inspections, and failures, updates recent EPRI work on c...

2007-12-20T23:59:59.000Z

Note: This page contains sample records for the topic "reactor components shield" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


221

X-ray transmissive debris shield  

DOE Patents (OSTI)

A composite window structure is described for transmitting x-ray radiation and for shielding radiation generated debris. In particular, separate layers of different x-ray transmissive materials are laminated together to form a high strength, x-ray transmissive debris shield which is particularly suited for use in high energy fluences. In one embodiment, the composite window comprises alternating layers of beryllium and a thermoset polymer.

Spielman, Rick B. (Albuquerque, NM)

1994-01-01T23:59:59.000Z

222

Method to Reduce Neutron Production in Small Clean Fusion Reactors...  

NLE Websites -- All DOE Office Websites (Extended Search)

and activation of reactor components and, in doing so, can advance the development of fusion reactors for electrical power and propulsion applications by alleviating the need for...

223

Health physics support for thermal shield repair at Connecticut Yankee  

Science Conference Proceedings (OSTI)

This article describes the radiation and safety controls used by Connecticut Yankee Atomic Power Company to support underwater repair work on the Haddam Neck Plant's core barrel thermal shield. The work was conducted by divers in the reactor cavity using remote tools and protected by a specially-constructed physical barrier that restricted their movements to a carefully defined and thoroughly surveyed area. A unique dosimetry test rig was used to determine the dose rate profiles within the work areas, and all underwater survey equipment was qualified against personnel dosimetric devices. Underwater operations were monitored and controlled by remote means (video surveillance and dosimetry telemetry), and health physics technicians were rotated through job coverage to avoid complacency and maximize training opportunities. A single, hot particle event occurred during one dive, but this was identified almost immediately and controlled to prevent excessive exposure to the diver.

Nevelos, W.F.; Gates, W.J. (Connecticut Yankee Atomic Power Co., East Hampton, CT (United States))

1988-10-01T23:59:59.000Z

224

Hewlett and Duncan - Atomic Shield | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Duncan - Atomic Shield Duncan - Atomic Shield Hewlett and Duncan - Atomic Shield Hewlett, Richard G. and Francis Duncan. Atomic Shield, 1947-1952. U.S. Atomic Energy Comission, 1972. The second volume of the three volume A History of the United States Atomic Energy Commission. Text in each PDF is fully searchable. "Hewlett and Duncan - Atomic Shield (complete).pdf" contains the complete text and images from Atomic Shield. 12mb "Hewlett and Duncan - Atomic Shield (figures only).pdf" contains hi-res (600dpi) scans of the images from Atomic Shield. 30mb Hewlett and Duncan - Atomic Shield (complete).pdf Hewlett and Duncan - Atomic Shield (figures only).pdf More Documents & Publications A History of the Atomic Energy Commission Hewlett and Duncan, Nuclear Navy, 1946-1962

225

AIR COOLED NEUTRONIC REACTOR  

DOE Patents (OSTI)

A nuclear reactor of the air-cooled, graphite moderated type is described. The active core consists of a cubicle mass of graphite, approximately 25 feet in each dimension, having horizontal channels of square cross section extending between two of the opposite faces, a plurality of cylindrical uranium slugs disposed in end to end abutting relationship within said channels providing a space in the channels through which air may be circulated, and a cadmium control rod extending within a channel provided in the moderator. Suitable shielding is provlded around the core, as are also provided a fuel element loading and discharge means, and a means to circulate air through the coolant channels through the fuel charels to cool the reactor.

Fermi, E.; Szilard, L.

1958-05-27T23:59:59.000Z

226

Reactor closure design for a pool-type fast reactor  

SciTech Connect

The reactor closure is the topmost structural part of a reactor module. For a pool-type fast reactor it is an especially important structure because it provides the interface between the primary coolant system and the main access area above the closure. The reactor closure comprises a stationary deck, a rotatable plug, the boundary elements of primary system and containment penetrations for equipment and auxiliary systems. This paper evaluates two different reactor closure design concepts, referred to as ''warm'' deck and ''hot'' deck, for a pool-type fast reactor with respect to their design features, technical merits, and economic benefits. The evaluation also includes functional, structural, and thermal analyses of the two deck design concepts. Issues related to their fabrication and shipping to the plant site are also addressed. The warm deck is a thick solid steel plate with under-the-deck insulation consisting of many layers of steel plates. The hot deck is a box-type structure consisting of a bottom plate reinforced with vertical ribs and cylinders. For insulation and radiation shielding, the region of the hot deck above the bottom plate is filled with steel balls. Conventional insulation is added on the top to further reduce heat loss into area above the deck. The design choice of the closure deck is strongly dependent on design features of the reactor; especially on the reactor module support. While the warm deck is preferable with the top support, the hot deck is better suited for the bottom support design of the module.

Chung, H.; Seidensticker, R.W.; Kann, W.J.; Bump, T.R.; Schatmeier, C.

1986-01-01T23:59:59.000Z

227

Review Article RADIATION SHIELDING TECHNOLOGY  

E-Print Network (OSTI)

written in 1965. The decade of the 1970's The Nuclear Non-Proliferation Treaty (NPT) of 1968 Test Ban Treaty of 1963 prevented the deployment of a nuclear rocket system and the program was ended for the non-hydrogen components. Because the hydrogen cross section increases with decreasing fast

Shultis, J. Kenneth

228

Advances toward a transportable antineutrino detector system for reactor monitoring and safeguards  

SciTech Connect

Nuclear reactors have served as the neutrino source for many fundamental physics experiments. The techniques developed by these experiments make it possible to use these very weakly interacting particles for a practical purpose. The large flux of antineutrinos that leaves a reactor carries information about two quantities of interest for safeguards: the reactor power and fissile inventory. Our SNL/LLNL collaboration has demonstrated that such antineutrino based monitoring is feasible using a relatively small cubic meter scale liquid scintillator detector at tens of meters standoff from a commercial Pressurized Water Reactor (PWR). With little or no burden on the plant operator we have been able to remotely and automatically monitor the reactor operational status (on/off), power level, and fuel burnup. The initial detector was deployed in an underground gallery that lies directly under the containment dome of an operating PWR. The gallery is 25 meters from the reactor core center, is rarely accessed by plant personnel, and provides a muon-screening effect of some 20-30 meters of water equivalent earth and concrete overburden. Unfortunately, many reactor facilities do not contain an equivalent underground location. We have therefore attempted to construct a complete detector system which would be capable of operating in an aboveground location and could be transported to a reactor facility with relative ease. A standard 6-meter shipping container was used as our transportable laboratory - containing active and passive shielding components, the antineutrino detector and all electronics, as well as climate control systems. This aboveground system was deployed and tested at the San Onofre Nuclear Generating Station (SONGS) in southern California in 2010 and early 2011. We will first present an overview of the initial demonstrations of our below ground detector. Then we will describe the aboveground system and the technological developments of the two antineutrino detectors that were deployed. Finally, some preliminary results of our aboveground test will be shown. (authors)

Reyna, D. [Sandia National Laboratories, Livermore, CA 94550 (United States); Bernstein, A. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Lund, J.; Kiff, S.; Cabrera-Palmer, B. [Sandia National Laboratories, Livermore, CA 94550 (United States); Bowden, N. S.; Dazeley, S.; Keefer, G. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States)

2011-07-01T23:59:59.000Z

229

Hydrogen Induced Cracking of Drip Shield  

DOE Green Energy (OSTI)

One potential failure mechanism for titanium and its alloys under repository conditions is via the absorption of atomic hydrogen in the metal crystal lattice. The resulting decreased ductility and fracture toughness may lead to brittle mechanical fracture called hydrogen-induced cracking (HIC) or hydrogen embrittlement. For the current design of the engineered barrier without backfill, HIC may be a problem since the titanium drip shield can be galvanically coupled to rock bolts (or wire mesh), which may fall onto the drip shield, thereby creating conditions for hydrogen production by electrochemical reaction. The purpose of this scientific analysis and modeling activity is to evaluate whether the drip shield will fail by HIC or not under repository conditions within 10,000 years of emplacement. This Analysis and Model Report (AMR) addresses features, events, and processes related to hydrogen induced cracking of the drip shield. REV 00 of this AMR served as a feed to ''Waste Package Degradation Process Model Report'' and was developed in accordance with the activity section ''Hydrogen Induced Cracking of Drip Shield'' of the development plan entitled ''Analysis and Model Reports to Support Waste Package PMR'' (CRWMS M&O 1999a). This AMR, prepared according to ''Technical Work Plan for: Waste Package Materials Data Analyses and Modeling'' (BSC 2002), is to feed the License Application.

G. De

2003-02-24T23:59:59.000Z

230

Microscreen radiation shield for thermoelectric generator  

DOE Patents (OSTI)

The present invention provides a microscreen radiation shield which reduces radiative heat losses in thermoelectric generators such as sodium heat engines without reducing the efficiency of operation of such devices. The radiation shield is adapted to be interposed between a reaction zone and a means for condensing an alkali metal vapor in a thermoelectric generator for converting heat energy directly to electrical energy. The radiation shield acts to reflect infrared radiation emanating from the reaction zone back toward the reaction zone while permitting the passage of the alkali metal vapor to the condensing means. The radiation shield includes a woven wire mesh screen or a metal foil having a plurality of orifices formed therein. The orifices in the foil and the spacing between the wires in the mesh is such that radiant heat is reflected back toward the reaction zone in the interior of the generator, while the much smaller diameter alkali metal atoms such as sodium pass directly through the orifices or along the metal surfaces of the shield and through the orifices with little or no impedance.

Hunt, Thomas K. (Ann Arbor, MI); Novak, Robert F. (Farmington Hills, MI); McBride, James R. (Ypsilanti, MI)

1990-01-01T23:59:59.000Z

231

ANL/APS/TB-21 Radiation Shielding of Insertion Device Beamlines  

NLE Websites -- All DOE Office Websites (Extended Search)

1 1 Radiation Shielding of Insertion Device Beamlines Using a Mirror as the First Optical Element W. Yun, B. Lai, K. J. Randall, S. Davey, D. R. Haeffner, P. K. Job, and D. Shu February 1995 Abstract The radiation shielding for an Advanced Photon Source (APS) insertion device beamline using a mirror as the first optical component is discussed. The beamline layout for a specific Synchrotron Radiation Instrumentation Collaborative Access Team beamline (sector 2 of SRI CAT) is described, and the methodology used to determine the radiation shielding is presented. Results indicate that, by using a x-ray mirror with a critical energy of 32 keV for total reflection, an undulator beam containing nearly all x-rays in the 0 - 32 keV spectral range can be delivered

232

Overview of SNS accelerator shielding analyses  

Science Conference Proceedings (OSTI)

The Spallation Neutron Source is an accelerator driven neutron scattering facility for materials research. During all phases of SNS development, including design, construction, commissioning and operation, extensive neutronics work was performed in order to provide adequate shielding, to assure safe facility operation from radiation protection point of view, and to optimize performance of the accelerator and target facility. Presently, most of the shielding work is concentrated on the beam lines and instrument enclosures to prepare for commissioning, safe operation and adequate radiation background in the future. Although the accelerator is built and in operation mode, there is extensive demand for shielding and activation analyses. It includes redesigning some parts of the facility, facility upgrades, designing additional structures, storage and transport containers for accelerator structures taken out of service, and performing radiation protection analyses and studies on residual dose rates inside the accelerator. (authors)

Popova, I.; Gallmeier, F. X.; Ferguson, P.; Iverson, E.; Lu, W. [ORNL/SNS, MS6475, PO Box 2008, Oak Ridge, TN 37831-6471 (United States)

2012-07-01T23:59:59.000Z

233

X-10 Graphite Reactor | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

X-10 Graphite Reactor X-10 Graphite Reactor X-10 Graphite Reactor X-10 Graphite Reactor When President Roosevelt in December 1942 authorized the Manhattan Project, the Oak Ridge site in eastern Tennessee had already been obtained and plans laid for an air-cooled experimental pile, a pilot chemical separation plant, and support facilities. The X-10 Graphite Reactor, designed and built in ten months, went into operation on November 4, 1943. The X-10 used neutrons emitted in the fission of uranium-235 to convert uranium-238 into a new element, plutonium-239. The reactor consists of a huge block of graphite, measuring 24 feet on each side, surrounded by several feet of high-density concrete as a radiation shield. The block is pierced by 1,248 horizontal diamond-shaped channels in

234

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Initial Assessment of Thermal Annealing Needs and Challenges Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Initial Assessment of Thermal Annealing Needs and Challenges The most life-limiting structural component in light-water reactors (LWR) is the reactor pressure vessel (RPV) because replacement of the RPV is not considered a viable option at this time. LWR licenses are now being extended from 40y to 60y by the U.S. Nuclear Regulatory Commission (NRC) with intentions to extend licenses to 80y and beyond. The RPV materials exhibit varying degrees of sensitivity to irradiation-induced embrittlement (decreased toughness) , as shown in Fig. 1.1, and extending operation from

235

LOS ALAMOS MOLTEN PLUTONIUM REACTOR EXPERIMENT (LAMPRE) HAZARD REPORT  

SciTech Connect

This report supersedes K-1-3425 and LA-2327(Prelim). The first experiment (LAMPRE I) in a program to develop molten plutonium fuels for fast reactors is described and the hazards associated with reactor operation are discussed and evaluated. The reactor desc=iption includes fuel element design, core configuration, sodium coolant system control, safety systems, fuel capsule charger, cover gas system, and shielding. Information of the site comprises population in surrounding areas, meteorological data, geology, and details of the reactor building. The hazmalfunction of the several elements comprising the reactor system. A calculation on the effect of fuel element bowiing appears in an appendix. (auth)

Swickard, E.O. comp.

1959-06-01T23:59:59.000Z

236

Engineering the fusion reactor first wall  

SciTech Connect

Recently the National Academy of Engineering published a set of Grand Challenges in Engineering in which the second item listed was entitled 'Provide energy from fusion'. Clearly a key component of this challenge is the science and technology associated with creating and maintaining burning plasmas. This is being vigorously addressed with both magnetic and inertial approaches with various experiments such as ITER and NIF. Considerably less attention is being given to another key component of this challenge, namely engineering the first wall that will contain the burning plasma. This is a daunting problem requiring technologies and materials that can not only survive, but also perform multiple essential functions in this extreme environment. These functions are (1) shield the remainder of the device from radiation. (2) convert of neutron energy to useful heat and (3) breed and extract tritium to maintain the reactor fuel supply. The first wall must not contaminate the plasma with impurities. It must be infused with cooling to maintain acceptable temperatures on plasma facing and structural components. It must not degrade. It must avoid excessive build-up of tritium on surfaces, and, if surface deposits do form, must be receptive to cleaning techniques. All these functions and constraints must be met while being subjected to nuclear and thermal radiation, particle bombardment, high magnetic fields, thermal cycling and occasional impingement of plasma on the surface. And, operating in a nuclear environment, the first wall must be fully maintainable by remotely-operated manipulators. Elements of the first wall challenge have been studied since the 1970' s both in the US and internationally. Considerable foundational work has been performed on plasma facing materials and breeding blanket/shield modules. Work has included neutronics, materials fabrication and joining, fluid flow, tritium breeding, tritium recovery and containment, energy conversion, materials damage and magnetohydrodynamics. While work to date has been quite valuable, no blanket concept has been built and operated in anything approaching a realistic fusion reactor environment. Rather, work has been limited to isolated experiments on first wall components and paper studies. The need now is to complete necessary R&D on first wall components, assemble components into a practical design, and test the first wall in a realistic fusion environment. Besides supporting work, major prototype experiments could be performed in non-nuclear experiments, as part of the ITER project and as part of the Component Test Facility. The latter is under active consideration and is a proposed machine which would use a driven plasma to expose an entire first wall to a fusion environment. Key US contributors to first wall research have been UCLA, UCSD, U of Wisconsin, LANL, ORNL, PNNL, Argonne and Idaho National Lab. Current efforts have been coordinated by UCLA. It is recognized that when this work progresses to a larger scale, leadership from a national laboratory will be required. LANL is well-prepared to provide such leadership.

Wurden, Glen [Los Alamos National Laboratory; Scott, Willms [Los Alamos National Laboratory

2008-01-01T23:59:59.000Z

237

Shielding analysis and design of the KIPT experimental neutron source facility of Ukraine.  

SciTech Connect

Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the conceptual design development of an experimental neutron source facility based on the use of an electron accelerator driven subcritical (ADS) facility [1]. The facility uses the existing electron accelerators of KIPT in Ukraine. The neutron source of the sub-critical assembly is generated from the interaction of 100 KW electron beam with a natural uranium target. The electron beam has a uniform spatial distribution and the electron energy in the range of 100 to 200 MeV, [2]. The main functions of the facility are the production of medical isotopes and the support of the Ukraine nuclear power industry. Reactor physics experiments and material performance characterization will also be carried out. The subcritical assembly is driven by neutrons generated by the electron beam interactions with the target material. A fraction of these neutrons has an energy above 50 MeV generated through the photo nuclear interactions. This neutron fraction is very small and it has an insignificant contribution to the subcritical assembly performance. However, these high energy neutrons are difficult to shield and they can be slowed down only through the inelastic scattering with heavy isotopes. Therefore the shielding design of this facility is more challenging relative to fission reactors. To attenuate these high energy neutrons, heavy metals (tungsten, iron, etc.) should be used. To reduce the construction cost, heavy concrete with 4.8 g/cm{sup 3} density is selected as a shielding material. The iron weight fraction in this concrete is about 0.6. The shape and thickness of the heavy concrete shield are defined to reduce the biological dose equivalent outside the shield to an acceptable level during operation. At the same time, special attention was give to reduce the total shield mass to reduce the construction cost. The shield design is configured to maintain the biological dose equivalent during operation {le} 0.5 mrem/h inside the subcritical hall, which is five times less than the allowable dose for working forty hours per week for 50 weeks per year. This study analyzed and designed the thickness and the shape of the radial and top shields of the neutron source based on the biological dose equivalent requirements inside the subcritical hall during operation. The Monte Carlo code MCNPX is selected because of its capabilities for transporting electrons, photons, and neutrons. Mesh based weight windows variance reduction technique is utilized to estimate the biological dose outside the shield with good statistics. A significant effort dedicated to the accurate prediction of the biological dose equivalent outside the shield boundary as a function of the shield thickness without geometrical approximations or material homogenization. The building wall was designed with ordinary concrete to reduce the biological dose equivalent to the public with a safety factor in the range of 5 to 20.

Zhong, Z.; Gohar, M. Y. A.; Naberezhnev, D.; Duo, J.; Nuclear Engineering Division

2008-10-31T23:59:59.000Z

238

5/6/03-5/12/03 Monochromator goniometer and shielding ...  

Science Conference Proceedings (OSTI)

... North short walls and column shields in are place. Top shield is positioned and leveled. The laminated shielding is installed above the shutter ...

239

Nuclear Reactors  

NLE Websites -- All DOE Office Websites (Extended Search)

Reactors Nuclear reactors created not only large amounts of plutonium needed for the weapons programs, but a variety of other interesting and useful radioisotopes. They produced...

240

Structural analysis of an LMFBR shield assembly duct under thermo-mechanical and seismic loads  

SciTech Connect

This paper describes the stress analysis performed to assess structural adequacy of the Clinch River Breeder Reactor (CRBR) core removable shield assemblies. Removable shield assemblies are located in the peripheral region of the core (between blanket assemblies and the fixed radial shield), and are subjected to severe cross-sectional thermal gradients and seismic loads requiring a relatively complex duct load pad design. For cost-effectiveness, the analysis was conducted in two stages. First, an elasto-plastic seismic stress analysis was performed using a detailed nonlinear finite element model (with gaps) of the load pad configuration. Next, in order to determine the total strain accumulation and the creep-fatigue damage the maximum seismic stresses combined with the ''worst'' thermal stresses from a single assembly model were used to perform a simplified inelastic analysis using two sets of material properties to bound the changing material conditions during reactor operation. This work demonstrated the necessity and applicability of the two simplified analysis techniques in elevated temperature structural design, i.e., the treatment of time-dependent degradation of material properties due to temperature and nuclear irradiation, and the use of time-independent finite element stress analysis results to perform a simplified creep-fatigue analysis.

Malik, S.N.; Sazawal, V.K.

1984-06-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor components shield" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


241

Electrically shielded enclosure with magnetically retained removable cover  

DOE Patents (OSTI)

Disclosed is an electrically shielded enclosure having electrical components therein and a removable electrically shielded cover over an opening in the enclosure with a magnetic securement mechanism provided to removably secure the cover to the enclosure in a manner which will provide easy access, yet also provide an electrical seal between the cover and the enclosure capable of preventing the passage of electrical radiation through the joint between the cover and the enclosure. Magnets are provided on the enclosure peripherally around the opening and facing the cover, and a ferromagnetic surface is provided on the mating surface of the cover facing the magnets, with a continuous electrical seal provided between the magnets and the ferromagnetic surface on the cover to prevent the leakage of electromagnetic radiation therethrough. In one embodiment the electrical seal includes a flexible metal casing or surface, which is attached to the enclosure and positioned between the magnets and the ferromagnetic surface on the cover, and which is sufficiently flexible to be capable of conforming to the ferromagnetic surface to provide an electrical seal between the cover and the enclosure. In another embodiment, the electrical seal includes a metal mesh associated with the enclosure and positioned between the magnets on the enclosure and the ferromagnetic surface on the cover. The metal mesh is also capable of conforming to the surface of the ferromagnetic surface to thereby provide an electrical seal between the cover and the enclosure.

Rivers, Craig J. (Livermore, CA); Lee, Roanne A. (San Francisco, CA); Jones, Glenn E. (Livermore, CA)

1996-01-01T23:59:59.000Z

242

Spherical torus fusion reactor  

DOE Patents (OSTI)

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.

1985-10-03T23:59:59.000Z

243

Hot cell shield plug extraction apparatus  

DOE Patents (OSTI)

A hot cell installation for the handling of highly radioactive material may comprise a dozen or more interconnected high density concrete vaults, the concrete vault walls having a thickness of approximately three feet. Typically, hot cells are constructed in rows so as to share as many shielding walls as possible. A typical overall length of a row of cells might be 70 yards. A secondary mechanism exists for placing certain objects into a cell. A typical hot cell has been constructed with 8 inch diameter holes through the exterior shielded walls in the vicinity of, and usually above, the viewing windows. It became evident that if the hot cell plugs could be removed and replaced conveniently significant savings in time and personnel exposure could be realized by using these 8 inch holes as entry ports. Fifteen inch cylindrical steel plugs with a diameter of eight inches weigh about two hundred pounds. The shield plug swing mechanism comprises a steel shielding plug mounted on a retraction device that enables the plug to be pulled out of the wall and supports the weight of the pulled out plug. The retraction device is mounted on a hinge, which allows the plug to be swung out of the way so that an operator can insert material into or remove it from the interior of the hot cell and then replace the plug quickly. The hinge mounting transmits the load of the retracted plug to the concrete wall.

Knapp, P.A.; Manhart, L.K.

1994-12-31T23:59:59.000Z

244

Summary of Prometheus Radiation Shielding Nuclear Design Analysis  

SciTech Connect

This report transmits a summary of radiation shielding nuclear design studies performed to support the Prometheus project. Together, the enclosures and references associated with this document describe NRPCT (KAPL & Bettis) shielding nuclear design analyses done for the project.

J. Stephens

2006-01-13T23:59:59.000Z

245

Interactions of Baroclinic Isolated Vortices: The Dominant Effect of Shielding  

Science Conference Proceedings (OSTI)

The interactions of two quasigeostrophic isolated shielded vortices are considered in a two-layer model and in the reduced-gravity approximation. Each shielded vortex is defined by a realistic horizontal profile of relative vorticity in the upper ...

S. Valcke; J. Verron

1997-04-01T23:59:59.000Z

246

Diffraction Studies of Irradiated Cladding and Duct Reactor Materials  

Science Conference Proceedings (OSTI)

Symposium, Characterization of Nuclear Reactor Materials and Components with Neutron and Synchrotron Radiation. Presentation Title, Diffraction Studies of ...

247

Preliminary Nuclear Calculations for the Shield Test Facility  

SciTech Connect

To find the critical size of the proposed shield test facility based upon available data and present construction concepts.

Baucom, H.H.

1960-01-11T23:59:59.000Z

248

Design of Transport Casks with Depleted Uranium Gamma Shield and Advanced Safety  

NLE Websites -- All DOE Office Websites (Extended Search)

Transport Casks with Depleted Uranium Gamma Shield and Advanced Safety Transport Casks with Depleted Uranium Gamma Shield and Advanced Safety Matveev V.Z., Morenko A.I., Shapovalov V.I. Russian Federal Nuclear Center - All-Russian Research Institute of Experimental Physics (RFNC-VNIIEF) 37 Mira Prospect, Sarov, Russia, 607190, matveev@vniief.ru Maslov A.A., Orlov V.K., Semenov A.G., Sergeev V.M., Yuferov O.I., Visik A.M. Bochvar Institute of Inorganic Materials (VNIINM) 5-A Rogova street, p.b. 369, Moscow, Russia, 123060, majul2000@mail.ru Abstract - The report is dedicated to a problem of creation of a new generation of dual-purpose transport packing complete sets (TPCS) 1 with advanced safety. These sets are intended for transportation and storage of spent nuclear fuel assemblies (SNFA) 2 of VVER reactors and spent spark elements (SSE)

249

Design guide for category V reactors transient reactors  

SciTech Connect

The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification, operation, maintenance, and decommissioning of DOE-owned reactors be in accordance with generally uniform standards, guides, and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirements of Category V reactor structures, components, and systems.

Brynda, W J; Karol, R C; Lobner, P R; Powell, R W; Straker, E A

1979-03-01T23:59:59.000Z

250

Construction of a Post-Irradiated Fuel Examination Shielded Enclosure Facility  

SciTech Connect

The U.S. Department of Energy (DOE) has committed to provide funding to the Idaho National Laboratory (INL) for new post-irradiation examination (PIE) equipment in support of advanced fuels development. This equipment will allow researchers at the INL to accurately characterize the behavior of experimental test fuels after they are removed from an experimental reactor also located at the INL. The accurate and detailed characterization of the fuel from the reactor, when used in conjunction with computer modeling, will allow DOE to more quickly understand the behavior of the fuel and to guide further development activities consistent with the missions of the INL and DOE. Due to the highly radioactive nature of the specimen samples that will be prepared and analyzed by the PIE equipment, shielded enclosures are required. The shielded cells will be located in the existing Analytical Laboratory (AL) basement (Rooms B-50 and B-51) at the INL Material and Fuels Complex (MFC). AL Rooms B-50 and B-51 will be modified to establish an area where sample containment and shielding will be provided for the analysis of radioactive fuels and materials while providing adequate protection for personnel and the environment. The area is comprised of three separate shielded cells for PIE instrumentation. Each cell contains an atmosphere interface enclosure (AIE) for contamination containment. The shielding will provide a work area consistent with the as-low-as-reasonably-achievable (ALARA) concept, assuming a source term of 10 samples in each of the three shielded areas. Source strength is assumed to be a maximum of 3 Ci at 0.75 MeV gamma for each sample. Each instrument listed below will be installed in an individual shielded enclosure: Shielded electron probe micro-analyzer (EPMA) Focused ion beam instrument (FIB) Micro-scale x-ray diffractometer (MXRD). The project is designed and expected to be built incrementally as funds are allocated. The initial phase will be to fund the construction activities, which will include facility modifications and construction of one shielded enclosure. Follow-up activities will be to construct two additional shielded enclosures to complete the suite of three separate but connected remote operated examination areas. Equipment purchases are to be capital procurement spread out over several years on a funded schedule. This paper discusses safety and operational considerations given during the conceptual design phase of the project. The paper considers such things as project material at risk (MAR), new processes and equipment, potential hazards, and the major modification evaluation process to determine if a preliminary Documented Safety Analysis (PDSA) is required. As part of that process, an evaluation was made of the potential hazards with the new project compared to the existing and historical work and associated hazards in the affected facility.

Michael A. Lehto, Ph.D.; Boyd D. Christensen

2008-05-01T23:59:59.000Z

251

Repair welding of fusion reactor components  

SciTech Connect

Experiments have shown that irradiated Type 316 stainless steel is susceptible to heat-affected-zone (HAZ) cracking upon cooling when welded using the gas tungsten arc (GTA) process under lateral constraint. The cracking has been hypothesized to be caused by stress-assisted helium bubble growth and rupture at grain boundaries. This study utilized an experimental welding setup which enabled different compressive stresses to be applied to the plates during welding. Autogenous GTA welds were produced in Type 316 stainless steel doped with 256 appm helium. The application of a compressive stress, 55 MPa, during welding suppressed the previously observed catastrophic cracking. Detailed examinations conducted after welding showed a dramatic change in helium bubble morphology. Grain boundary bubble growth along directions parallel to the weld was suppressed. Results suggest that stress-modified welding techniques may be used to suppress or eliminate helium-induced cracking during joining of irradiated materials.

Chin, B.A.

1993-05-15T23:59:59.000Z

252

Materials Challenges in Next Generation Nuclear Reactors  

Science Conference Proceedings (OSTI)

Materials under active consideration for use in different reactor components ... A Theoretical Model of Corrosion Rate Distribution in Liquid LBE Flow Loop at ...

253

Reactor Core Assembly - HFIR Technical Parameters | ORNL Neutron Sciences  

NLE Websites -- All DOE Office Websites (Extended Search)

Home › Facilities › HFIR › Reactor Core Assembly Home › Facilities › HFIR › Reactor Core Assembly Reactor Core Assembly The reactor core assembly is contained in an 8-ft (2.44-m)-diameter pressure vessel located in a pool of water. The top of the pressure vessel is 17 ft (5.18 m) below the pool surface, and the reactor horizontal mid-plane is 27.5 ft (8.38 m) below the pool surface. The control plate drive mechanisms are located in a subpile room beneath the pressure vessel. These features provide the necessary shielding for working above the reactor core and greatly facilitate access to the pressure vessel, core, and reflector regions. In-core irradiation and experiment locations (cross section at horizontal midplane) Reactor core assembly Reactor core assembly: (1) in-core irradiation and experiment locations,

254

Atlas SCT/Pixel Grounding and Shielding Note 1 ATLAS SCT/Pixel Grounding and Shielding Note  

E-Print Network (OSTI)

: Aluminum 150 µm thick cylinder with heater tapes and 600 µm CFRP (carbon fiber reinforced polymer, Schematic 4. This tap would provide data on the temperature variation of the resistance of the TEMP1 path inside the shield (5) Conductors outside the shield (6) Temperature monitoring The Module Array Shield

California at Santa Cruz, University of

255

Advanced Nuclear Reactors | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Advanced Nuclear Advanced Nuclear Reactors Advanced Nuclear Reactors Turbulent Flow of Coolant in an Advanced Nuclear Reactor Visualizing Coolant Flow in Sodium Reactor Subassemblies Sodium-cooled Fast Reactor (SFR) Coolant Flow At the heart of a nuclear power plant is the reactor. The fuel assembly is placed inside a reactor vessel where all the nuclear reactions occur to produce the heat and steam used for power generation. Nonetheless, an entire power plant consists of many other support components and key structures like coolant pipes; pumps and tanks including their surrounding steel framing; and concrete containment and support structures. The Reactors Product Line within NEAMS is concerned with modeling the reactor vessel as well as those components of a complete power plant that

256

Criticality safety and shielding design issues in the development of a high-capacity cask for truck transport  

SciTech Connect

General Atomics (GA) will be submitting an application for certification to the US Nuclear Regulatory Commission (NRC) for the GA-4 and GA-9 Casks In 1992. The GA-4 and GA-9 Casks are high-capacity legal weight truck casks designed to transport light water reactor spent fuel assemblies. To maintain a capacity of four pressurized-water-reactor (PWR) spent fuel assemblies, the GA-4 Cask uses burnup credit as part of the criticality control for initial enrichments over 3.0 wt% U-235. Using the US Department of Energy (DOE) Burnup Credit Program as a basis, GA has performed burnup credit analysis which is included in the Safety Analysis Report for Packaging (SARP). The GA-9 Cask can meet the criticality safety requirements using the ``fresh fuel`` assumption. Our approach to shielding design is to optimize the GA-4 and GA-9 Cask shielding configurations for minimum weights and maximum payloads. This optimization involves the use of the most effective shielding material, square cross-section geometry with rounded corners and tapered neutron shielding sections in the non-fuel regions.

Boshoven, J.K.

1992-08-01T23:59:59.000Z

257

Criticality safety and shielding design issues in the development of a high-capacity cask for truck transport  

Science Conference Proceedings (OSTI)

General Atomics (GA) will be submitting an application for certification to the US Nuclear Regulatory Commission (NRC) for the GA-4 and GA-9 Casks In 1992. The GA-4 and GA-9 Casks are high-capacity legal weight truck casks designed to transport light water reactor spent fuel assemblies. To maintain a capacity of four pressurized-water-reactor (PWR) spent fuel assemblies, the GA-4 Cask uses burnup credit as part of the criticality control for initial enrichments over 3.0 wt% U-235. Using the US Department of Energy (DOE) Burnup Credit Program as a basis, GA has performed burnup credit analysis which is included in the Safety Analysis Report for Packaging (SARP). The GA-9 Cask can meet the criticality safety requirements using the fresh fuel'' assumption. Our approach to shielding design is to optimize the GA-4 and GA-9 Cask shielding configurations for minimum weights and maximum payloads. This optimization involves the use of the most effective shielding material, square cross-section geometry with rounded corners and tapered neutron shielding sections in the non-fuel regions.

Boshoven, J.K.

1992-01-01T23:59:59.000Z

258

Conceptual innovations in hybrid reactors  

SciTech Connect

A number of innovations in the conception of fusion-fission hybrid reactors, including the blanket, the fusion driver, the coupling of the fusion and the fission components as well as the application of hybrid reactors are described, and their feasibility assessed.

Greenspan, E.; Miley, G.H.

1980-01-01T23:59:59.000Z

259

AN ADVANCED SODIUM-GRAPHITE REACTOR NUCLEAR POWER PLANT  

SciTech Connect

An advanced sodium-cooled, graphite-moderated nuclear power plant is described which utilizes high-pressure, high-temperature steam to generate electricity at a high thermal efficiency. Steam is generated at 2400 psig, superheated to 1050 deg F and, after partial expansion in the turbine, reheated to 1000 deg F. Net thermal efficiency of the plant is 42.3%. In a plant sized to produce a net electrical output of 256 Mw, the estimated cost is 8232/kw. Estimated cost of power generation is 6.7 mills/kwh. In a similar plant with a net electrical output of 530 Mw, the estimated power generating cost is 5.4 mills/ kwh. Most of the components of the plant are within the capability of current technology. The major exception is the fuel material, uranium carbide. Preliminary results of the development work now in progress indicate that uranium carbide would be an excellent fuel for high-temperature reactors, but temperature and burnup limitation have yet to be firmly established. Additional development work is also required on the steam generators. These are the single-barrier type similar to those which will be used in the Enrico Fernri Fast Breeder Reactor plant but produce steam at higher pressure and temperature. Questions also remain regarding the use of nitrogen as a cover gas over sodium at 1200 deg F and compatibility of the materials used in the primary neutron shield. All of these questions are currently under investigation. (auth)

Churchill, J.R.; Renard, J.

1960-03-15T23:59:59.000Z

260

High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1980  

SciTech Connect

Research activities are described concerning HTGR chemistry; fueled graphite development; prestressed concrete pressure vessel development; structural materials; HTGR graphite studies; HTR core evaluation; reactor physics; shielding; application and project assessments; and HTR Core Flow Test Loop studies.

Not Available

1981-08-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor components shield" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


261

Electronically shielded solid state charged particle detector  

DOE Patents (OSTI)

An electronically shielded solid state charged particle detector system having enhanced radio frequency interference immunity includes a detector housing with a detector entrance opening for receiving the charged particles. A charged particle detector having an active surface is disposed within the housing. The active surface faces toward the detector entrance opening for providing electrical signals representative of the received charged particles when the received charged particles are applied to the active surface. A conductive layer is disposed upon the active surface. In a preferred embodiment, a nonconductive layer is disposed between the conductive layer and the active surface. The conductive layer is electrically coupled to the detector housing to provide a substantially continuous conductive electrical shield surrounding the active surface. The inner surface of the detector housing is supplemented with a radio frequency absorbing material such as ferrite.

Balmer, D.K.; Haverty, T.W.; Nordin, C.W.; Tyree, W.H.

1995-12-31T23:59:59.000Z

262

Supplemental heating of deposition tooling shields  

DOE Patents (OSTI)

A method of reducing particle generation from the thin coating deposited on the internal surfaces of a deposition chamber which undergoes temperature variation greater than 100.degree. C. comprising maintaining the temperature variation of the internal surfaces low enough during the process cycle to keep thermal expansion stresses between the coating and the surfaces under 500 MPa. For titanium nitride deposited on stainless steel, this means keeping temperature variations under approximately 70.degree. C. in a chamber that may be heated to over 350.degree. C. during a typical processing operation. Preferably, a supplemental heater is mounted behind the upper shield and controlled by a temperature sensitive element which provides feedback control based on the temperature of the upper shield.

Ohlhausen, James A. (Albuquerque, NM); Peebles, Diane E. (Albuquerque, NM); Hunter, John A. (Albuquerque, NM); Eckelmeyer, Kenneth H. (Albuquerque, NM)

2000-01-01T23:59:59.000Z

263

Enrichment Determination of Uranium in Shielded Configurations  

Science Conference Proceedings (OSTI)

The determination of the enrichment of uranium is required in many safeguards and security applications. Typical methods of determining the enrichment rely on detecting the 186 keV gamma ray emitted by {sup 235}U. In some applications, the uranium is surrounded by external shields, and removal of the shields is undesirable. In these situations, methods relying on the detection of the 186 keV gamma fail because the gamma ray is shielded easily. Oak Ridge National Laboratory (ORNL) has previously measured the enrichment of shielded uranium metal using active neutron interrogation. The method consists of measuring the time distribution of fast neutrons from induced fissions with large plastic scintillator detectors. To determine the enrichment, the measurements are compared to a calibration surface that is created from Monte Carlo simulations where the enrichment in the models is varied. In previous measurements, the geometry was always known. ORNL is extending this method to situations where the geometry and materials present are not known in advance. In the new method, the interrogating neutrons are both time and directionally tagged, and an array of small plastic scintillators measures the uncollided interrogating neutrons. Therefore, the attenuation through the item along many different paths is known. By applying image reconstruction techniques, an image of the item is created which shows the position-dependent attenuation. The image permits estimating the geometry and materials present, and these estimates are used as input for the Monte Carlo simulations. As before, simulations predict the time distribution of induced fission neutrons for different enrichments. Matching the measured time distribution to the closest prediction from the simulations provides an estimate of the enrichment. This presentation discusses the method and provides results from recent simulations that show the importance of knowing the geometry and materials from the imaging system.

Crye, Jason Michael [ORNL; Hall, Howard L [ORNL; McConchie, Seth M [ORNL; Mihalczo, John T [ORNL; Pena, Kirsten E [ORNL

2011-01-01T23:59:59.000Z

264

SIGNAL GROUPING FOR CONDITION MONITORING OF NUCLEAR POWER PLANT COMPONENTS  

E-Print Network (OSTI)

SIGNAL GROUPING FOR CONDITION MONITORING OF NUCLEAR POWER PLANT COMPONENTS Piero Baraldi between those used to monitor the reactor coolant pump of a Pressurized Water Reactor (PWR) is considered Monitoring, Empirical Modeling, Power Plants, Safety Critical Nuclear Instrumentation, Autoassociative models

265

Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A)  

Science Conference Proceedings (OSTI)

The Materials Reliability Program (MRP) developed inspection and evaluation (I&E) guidelines for managing long-term aging reactor vessel internal components of pressurized water reactors (PWRs) reactor internals. Specifically, the guidelines are applicable to reactor vessel internal structural components; they do not address fuel assemblies, reactivity control assemblies, or welded attachments to the reactor vessel.

2011-12-23T23:59:59.000Z

266

Reactivity Control Schemes for Fast Spectrum Space Nuclear Reactors  

Science Conference Proceedings (OSTI)

Several different reactivity control schemes are considered for future space nuclear reactor power systems. Each of these control schemes uses a combination of boron carbide absorbers and/or beryllium oxide reflectors to achieve sufficient reactivity swing to keep the reactor subcritical during launch and to provide sufficient excess reactivity to operate the reactor over its expected 7–15 year lifetime. The size and shape of the control system directly impacts the size and mass of the space reactor's reflector and shadow shield

Aaron E. Craft; Jeffrey C. King

2008-01-01T23:59:59.000Z

267

FELIX: an experimental facility to study electromagnetic effects for first wall, blanket, and shield systems  

Science Conference Proceedings (OSTI)

As part of the DOE First Wall/Blanket/Shield (FW/B/S) Engineering Test Program, Argonne National Laboratory (ANL) is carrying out an experimental program to study electromagnetic effects. Electromagnetic effect in a FW/B/S system include time delay of equilibrium field penetration to the plasma, forces and torques, eddy current heating, and electrical arcing. Such effects can arise in a fusion reactor from plasma disruption and normal and abnormal charge and discharge of the magnets. The facility will consist primarily of a solenoid magnet, a surrounding pulsed dipole magnet, and associated power supplies and support structure.

Praeg, W.F.; Turner, L.R.; Biggs, J.; Bywater, J.; Fuja, R.; Knott, M.; Lari, R.J.; McGhee, D.G.; Wehrle, R.B.

1981-01-01T23:59:59.000Z

268

Measurements of Materials Shielding Properties with 1 GeV/nuc56Fe  

DOE Green Energy (OSTI)

The design of future spacecraft such as the Crew ExplorationVehicle must take into account the radiation shielding properties of boththe structural components as well as dedicated shielding materials. Sincemodest depths of shielding stop the vast majority of Solar EnergeticParticles (SEP), the greater challenge is posed by the need to shieldcrew from the Galactic Cosmic Rays (GCR), which include highly-chargedand highly-energetic particles. Here, we report on results from testsperformed with beams of 1 GeV/nuc 56Fe at the Brookhaven NationalLaboratory. A wide variety of targets, both elemental and composite, wereplaced in the particle beams, and the spectra of particles emerging fromthe targets were measured using a stack of silicon detectors. Results arepresented primarily in terms of dose reduction per g cm-2 of targetmaterial, and support the conclusions of an earlier calculation by Wilsonet al. showing that performance improves as the shield's mass numberdecreases, with hydrogen being by far the most effective. The data alsoshow that, as depth increases, the incremental benefit of addingshielding decreases, particularlyfor aluminum and other elements withhigher atomic mass numbers.

Zeitlin, Cary; Guetersloh, Stephen B.; Heilbronn, Lawrence H.; Miller, Jack

2006-05-15T23:59:59.000Z

269

Increasing transcurium production efficiency through direct resonance shielding  

Science Conference Proceedings (OSTI)

The Radiochemical Engineering Development Center at Oak Ridge National Laboratory is the world s leader in production of 252Cf. This and other heavy actinides are produced by irradiation of mixed curium/americium targets in the High Flux Isotope Reactor. Due to the strong dependence of isotopic cross sections upon incoming neutron energy, the efficiency with which an isotope is transmuted is highly dependent upon the neutron flux energy spectrum and intensities. There are certain energy ranges in which the rate of fissions in feedstock materials can be minimized relative to the rate of (n, ) absorptions. It is shown that by perturbing the flux spectrum, it is possible to alter the net consumption of curium feedstock, as well as the yields of key isotopes for the heavy element research program, such as 249Bk and 252Cf. This flux spectrum perturbation is accomplished by means of focused resonance shielding through the use of filter materials. It is further shown that these perturbations can alter the target yields in a significant way, increasing the amount of 252Cf produced per unit curium consumption by over 40%.

Hogle, Susan L [ORNL; Maldonado, G Ivan [ORNL; Alexander, Charles W [ORNL

2013-01-01T23:59:59.000Z

270

Distributed resonance self-shielding using the equivalence principle  

SciTech Connect

This paper presents an extension of the equivalence principle to allow distributed resonance self-shielding in a multi-region fuel configuration. Rational expansion of fuel-to-fuel collision probability is applied in order to establish equivalence between the actual fuel configuration and a homogeneous mixture of hydrogen and resonant absorber, which is a commonly used model to calculate library tables of resonance integrals. The main steps in derivation are given along with the basic physics assumptions on which the presented approach relies. The method has been implemented in the lattice code WIMS-AECL and routinely used for calculation of CANDU-type reactor lattices. Its capabilities are illustrated by comparison of WIMS-AECL and MCNP results of {sup 238}U resonance capture in a CANDU lattice cell. In order to determine optimal rational expansion of fuel-to-fuel collision probability, the calculations were carried out by varying the number of rational terms from 1 to 6. The results show that 4 terms are sufficient. The further increase of the number of terms affects the computing time, while the impact on accuracy is negligible. To illustrate the convergence of the results, the fuel subdivision is gradually refined varying the number of fuel pin subdivisions from 1 to 32 equal-area annuli. The results show very good agreement with the reference MCNP calculation. (authors)

Altiparmakov, D. [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, ON K0J 1J0 (Canada)

2012-07-01T23:59:59.000Z

271

Structural characteristics of proposed ITER (International Thermonuclear Experimental Reactor) TF (toroidal field) coil conductor  

SciTech Connect

This paper analyzes the effect of transverse loading on a cable-in-conduit conductor which has been proposed for the toroidal field coils of the International Thermonuclear Experimental Reactor. The primary components of this conductor are a loose cable of superconducting wires, a thin-wall tube for helium containment, and a U-shaped structural channel. A method is given where the geometry of this conductor can be optimized for a given set of operating conditions. It is shown, using finite-element modeling, that the structural channel is effective in supporting loads due to transverse forces and internal pressure. In addition, it is shown that the superconducting cable is effectively shielded from external transverse loads that might otherwise degrade its current carrying capacity. 10 refs., 10 figs., 3 tabs.

Gibson, C.R.; Miller, J.R.

1988-08-19T23:59:59.000Z

272

NUCLEAR REACTOR  

DOE Patents (OSTI)

A boiling-water nuclear reactor is described wherein control is effected by varying the moderator-to-fuel ratio in the reactor core. This is accomplished by providing control tubes containing a liquid control moderator in the reactor core and providing means for varying the amount of control moderatcr within the control tubes.

Treshow, M.

1961-09-01T23:59:59.000Z

273

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A reactor in which at least a portion of the moderator is in the form of movable refractory balls is described. In addition to their moderating capacity, these balls may serve as carriers for fissionable material or fertile material, or may serve in a coolant capacity to remove heat from the reactor. A pneumatic system is used to circulate the balls through the reactor.

Daniels, F.

1959-10-27T23:59:59.000Z

274

Stack Components  

NLE Websites -- All DOE Office Websites (Extended Search)

Stack Components Stack Components Nancy L. Garland Energy Efficiency and Renewable Energy Hydrogen, Fuel Cells and Infrastructure Technologies Program Fuel Cell Team FORS 5G-086 (202) 586-5673 nancy.garland@ee.doe.gov Stack Components F u e l P r o c e s s o r Bipolar Plate Cathode + Anode - Electrolyte H+ H+ HYDROGEN OXYGEN Example shown is for acidic electrolytes Bipolar Plate e - e - O 2 O 2 O 2 e - H+ Bipolar Plate Bipolar Plate Cathode + Anode - Electrolyte H+ H+ H+ H+ HYDROGEN OXYGEN Example shown is for acidic electrolytes Bipolar Plate Bipolar Plate e - e - e - e - O 2 O 2 O 2 O 2 O 2 O 2 e - e - H+ H+ Power Stack Component Barriers $10 Other Bipolar Plates Membranes Electrodes $25 $5 $5 Fuel Cell Power Systems $45/kW BARRIERS * Stack material cost/manufacturing * Durability * Electrode performance * Thermal and water management Stack Component Targets

275

Sandia National Laboratories: Research: Facilities: Sandia Pulsed Reactor  

NLE Websites -- All DOE Office Websites (Extended Search)

Sandia Pulsed Reactor Facility - Critical Experiments Sandia Pulsed Reactor Facility - Critical Experiments Sandia scientist John Ford places fuel rods in the Seven Percent Critical Experiment (7uPCX) at the Sandia Pulsed Reactor Facility Critical Experiments (SPRF/CX) test reactor - a reactor stripped down to its simplest form. The Sandia Pulsed Reactor Facility - Critical Experiments (SPRF/CX) provides a flexible, shielded location for performing critical experiments that employ different reactor core configurations and fuel types. The facility is also available for hands-on nuclear criticality safety training. Research and other activities The 7% series, an evaluation of various core characteristics for higher commercial-fuel enrichment, is currently under way at the SPRF/CX. Past critical experiments at the SPRF/CX have included the Burnup Credit

276

CONVECTION REACTOR  

DOE Patents (OSTI)

An homogeneous nuclear power reactor utilizing convection circulation of the liquid fuel is proposed. The reactor has an internal heat exchanger looated in the same pressure vessel as the critical assembly, thereby eliminating necessity for handling the hot liquid fuel outside the reactor pressure vessel during normal operation. The liquid fuel used in this reactor eliminates the necessity for extensive radiolytic gas rocombination apparatus, and the reactor is resiliently pressurized and, without any movable mechanical apparatus, automatically regulates itself to the condition of criticality during moderate variations in temperature snd pressure and shuts itself down as the pressure exceeds a predetermined safe operating value.

Hammond, R.P.; King, L.D.P.

1960-03-22T23:59:59.000Z

277

REACTOR DEVELOPMENT PROGRAM, PROGRESS REPORT, MAY 1961  

SciTech Connect

General research and development on water-cooled and sodium-cooled reactors are reported along with specific developments on EBWR, BORAK-V, EBR-I, and EBR-H. Thermal and fast reactor safety studies are summarized in terms of fuel-coolant chemical reactions, kinetics of oxidation and ignition of reactor materials, core meltdown studies, and a sodium vapor pressure furnace. Evaluations were made of improved fast reactors for central station power and of a 50-Mwe Prototype Organic Power Reactor (POPR). Developments in instruments, reactcr fuels and materials, reactor components, heat engineering, separations processes, and advanced reactcrs are discussed. (M.C.G.)

1961-06-15T23:59:59.000Z

278

Radiation shielding for underground low-background experiments  

E-Print Network (OSTI)

The design task of creating an efficient radiation shield for the new COBRA double-beta decay experiment led to a comprehensive study of commercially available shielding materials. The aim was to find the most efficient combination of materials under the constraints of an extreme low-background experiment operating in a typical underground laboratory. All existing shield configurations for this type of experiment have been found to perform sub-optimally in comparison to the class of multilayered configurations proposed in this study. The method used here to create a specific shield configuration should yield a close to optimal result when applied to any experiment utilising a radiation shield. In particular, the survey of single material response to a given radiation source turns out to give a guideline for the construction of efficient multilayer shields.

D. Y Stewart; P. F. Harrison; B. Morgan; Y. A. Ramachers

2006-07-31T23:59:59.000Z

279

Shielding Design of the Spallation Neutron Source (SNS)  

Science Conference Proceedings (OSTI)

The shielding design is important for the construction of an intense high-energy accelerator facility like the proposed Spallation Neutron Source (SNS) due to its impact on conventional facility design, maintenance operations, and since the cost for the radiation shielding shares a considerable part of the total facility costs. A calculational strategy utilizing coupled high energy Monte Carlo calculations and multi-dimensional discrete ordinates calculations, along with semi-empirical calculations, was implemented to perform the conceptual design shielding assessment of the proposed SNS. Biological shields have been designed and assessed for the proton beam transport system and associated beam dumps, the target station, and the target service cell and general remote maintenance cell. Shielding requirements have been assessed with respect to weight, space, and dose-rate constraints for operating, shutdown, and accident conditions. A discussion of the proposed facility design, conceptual design shielding requirements, calculational strategy, source terms, preliminary results and conclusions, and recommendations for additional analyses are presented.

Johnson, J.O.

1998-09-17T23:59:59.000Z

280

The Spallation Neutron Source (SNS) conceptual design shielding analysis  

SciTech Connect

The shielding design is important for the construction of an intense high-energy accelerator facility like the proposed Spallation Neutron Source (SNS) due to its impact on conventional facility design, maintenance operations, and since the cost for the radiation shielding shares a considerable part of the total facility costs. A calculational strategy utilizing coupled high energy Monte Carlo calculations and multi-dimensional discrete ordinates calculations, along with semi-empirical calculations, was implemented to perform the conceptual design shielding assessment of the proposed SNS. Biological shields have been designed and assessed for the proton beam transport system and associated beam dumps, the target station, and the target service cell and general remote maintenance cell. Shielding requirements have been assessed with respect to weight, space, and dose-rate constraints for operating, shutdown, and accident conditions. A discussion of the proposed facility design, conceptual design shielding requirements calculational strategy, source terms, preliminary results and conclusions, and recommendations for additional analyses are presented.

Johnson, J.O.; Odano, N.; Lillie, R.A.

1998-03-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor components shield" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


281

Self-sealing solar collector shield  

SciTech Connect

Window structure for a solar collector shield or the like having a frame defining a row of separate window openings. A selfcontained window unit is installed in each opening. Each window unit comprises a pair of rectangular glass panes held in laterally spaced parallel relation by a peripheral window frame. A mounting frame is provided for mounting each window unit in its window opening. Each mounting frame has overlapping frame members. Adjacent side frame members of adjacent window units also overlap. The overlap is for the purpose of excluding moisture and preventing heat loss. A peripheral window seal is provided for the same purpose.

Katona, J.W.

1980-03-25T23:59:59.000Z

282

X-ray transmissive debris shield  

DOE Patents (OSTI)

An X-ray debris shield for use in X-ray lithography that is comprised of an X-ray window having a layer of low density foam exhibits increased longevity without a substantial increase in exposure time. The low density foam layer serves to absorb the debris emitted from the X-ray source and attenuate the shock to the window so as to reduce the chance of breakage. Because the foam is low density, the X-rays are hardly attenuated by the foam and thus the exposure time is not substantially increased.

Spielman, R.B.

1996-05-21T23:59:59.000Z

283

X-ray transmissive debris shield  

DOE Patents (OSTI)

An X-ray debris shield for use in X-ray lithography that is comprised of an X-ray window having a layer of low density foam exhibits increased longevity without a substantial increase in exposure time. The low density foam layer serves to absorb the debris emitted from the X-ray source and attenuate the shock to the window so as to reduce the chance of breakage. Because the foam is low density, the X-rays are hardly attenuated by the foam and thus the exposure time is not substantially increased.

Spielman, Rick B. (Albuquerque, NM)

1996-01-01T23:59:59.000Z

284

Nuclear Reactor Severe Accident Experiments  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Reactor Severe Accident Experiments Nuclear Reactor Severe Accident Experiments Capabilities Engineering Experimentation Reactor Safety Testing and Analysis Overview Nuclear Reactor Severe Accident Experiments MAX NSTF SNAKE Aerosol Experiments System Components Laser Applications Robots Applications Other Facilities Other Capabilities Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Nuclear Reactor Severe Accident Experiments 1 2 3 4 5 6 7 We perform experiments simulating reactor core melt phenomena in which molten core debris ("corium") erodes the concrete floor of a containment building. This occurred during the Fukushima nuclear power plant accident though the extent of concrete damage is yet unknown. This video shows the top view of a churning molten pool of uranium oxide at 2000°C (3600°F) seen during an experiment at Argonne. Corium behaves much like lava.

285

Heat pipe nuclear reactor for space power  

SciTech Connect

A heat-pipe cooled nuclear reactor has been designed to provide 3.2 MW(t) to an out-of-core thermionic conversion system. The reactor is a fast reactor designed to operate at a nominal heat pipe temperature of 1675/sup 0/K. Each reactor fuel element consists of a hexagonal molybdenum block which is bonded along its axis to one end of a molybdenum, lithium vapor, heat pipe. The block is perforated with an array of longitudinal holes which are loaded with UO/sub 2/ pellets. The heat pipe transfers heat directly to a string of six thermionic converters which are bonded along the other end of the heat pipe. An assembly of 90 such fuel elements forms a hexagonal core. The core is surrounded by a thermal radiation shield, a thin thermal neutron absorber and a BeO reflector containing boron loaded control drums.

Koenig, D.R.

1976-01-01T23:59:59.000Z

286

Technology Development to Improve Radiation Shielding for Material Inspections  

Science Conference Proceedings (OSTI)

Improvements in shielding are necessary to reduce radiation exposure to workers in the nuclear power industry, especially as new inspection and maintenance work tests the industry's ability to further reduce individual and collective radiation doses. To date, new shielding technologies for the nuclear industry have been limited. However, interest in developing new shielding devices has been increasing in other industry sectors such as space exploration and medicine. This report presents a thorough review...

2008-09-22T23:59:59.000Z

287

Neutronics at Wisconsin, ORNL advances ITER shielding and internationa...  

NLE Websites -- All DOE Office Websites (Extended Search)

Neutronics at Wisconsin, ORNL advances ITER shielding and international collaboration American Fusion News Category: U.S. ITER Link: Neutronics at Wisconsin, ORNL advances ITER...

288

Treatment of Shielding in Real-Time Source Tracking Software  

Science Conference Proceedings (OSTI)

Technical Paper / Special Issue on the 16th Biennial Topical Meeting of the Radiation Protection and Shielding Division / Radiation Measurements and General Instrumentation

Raymond T. Klann; Sergio C. de la Barrera; Richard B. Vilim

289

GMAW Shielding Gas Flow Optimisation by Refinement of Nozzle ...  

Science Conference Proceedings (OSTI)

However, sufficient shielding gas coverage of the weld region is essential for the generation of high quality welds, and drafts can be detrimental to its efficiency.

290

APS Radiation Safety Shielding Committee for Design Reviews  

NLE Websites -- All DOE Office Websites (Extended Search)

Minutes 2008 (restricted) 2007 (restricted) 2006 (restricted) 2005 (restricted) APS Radiation Safety Shielding Committee for Design Reviews (RSSCDR) (February 9, 2012) 1. Purpose...

291

REACTOR DEVELOPMENT PROGRAM PROGRESS REPORT, FEBRUARY 1961  

SciTech Connect

Design, development, and testing efforts were continued on BORAX-V, EBR- I, EBR-II, EBWR, JUGGERNAUT ZPRIII, ZPR-VI, and ZPR-W. An evaluation program is outlined for Pebble Bed Reactor designs. Fast and thermal reactor safety studies were conducted. Experimental and theoretical studies in applied nuclear and reactor physics are dsscribed. Developments made in reactor components, fuels, and materials are discussed. Heat engineering studies were conducted on steam separation, and velocity and void distributions in two-phase systems. Fluidization and fluoride volatility separation, and chemical-metallurgical separation processes were studied. Advanced reactor concepts that were discusssd includsed. Basic Radiation Effects Beactor, Biogeonuclear Reactor, Fast Reactor Test Facility, compact high-power density fast reactors, AHFR hydraulic test loop, Packed Bed Reactor, and direct conversion. (For preceding period see ANL- 6328.) (B.O.G.)

1961-03-15T23:59:59.000Z

292

SHIELDING REQUIREMENTS FOR NSLS-II.  

Science Conference Proceedings (OSTI)

Brookhaven National Laboratory is in the process of designing a new Electron Synchrotron for scientific research using synchrotron radiation. This facility, called the 'National Synchrotron Light Source II' (NSLS-II), will provide x-ray radiation of ultra-high brightness and exceptional spatial and energy resolution. It will also provide advanced insertion devices, optics, detectors, and robotics, and a suite of scientific instruments designed to maximize the scientific output of the facility. The project scope includes the design, construction, installation, and commissioning of the following accelerators: a 200 MeV linac, a booster accelerator operating from 200 MeV to 3.0 GeV, and the storage ring which stores a maximum of 500 mA current of electrons at an energy of 3.0 GeV. It is planned to operate the facility primarily in a top-off mode, thereby maintaining the maximum variation in stored beam current to < 1%. Because of the very demanding requirements for beam emittance and synchrotron radiation brilliance, the beam life-time is expected to be low, on the order of 2-3 hours. Analysis of the bulk shielding for operating this facility and the input parameters used for this analysis are discussed in this paper. The characteristics of each of the accelerators and their operating modes are summarized with the input assumptions for the bulk shielding analysis.

JOB,P.K.; CASEY, W.R.

2008-01-02T23:59:59.000Z

293

COMBINING MAGNETIC SHIELDING AND CRYOPUMPING FOR A NEUTRAL BEAM SOURCE  

E-Print Network (OSTI)

generated by the fusion reactor while the cold surface willfor the Toka. -nak Fusion Test Reactor", in Proceed­ ings of

Tanabe, J.

2010-01-01T23:59:59.000Z

294

Pulsed, Photonuclear-induced, Neutron Measurements of Nuclear Materials with Composite Shielding  

SciTech Connect

Active measurements were performed using a 10-MeV electron accelerator with inspection objects containing various nuclear and nonnuclear materials available at the Idaho National Laboratory’s Zero Power Physics Reactor (ZPPR) facility. The inspection objects were assembled from ZPPR reactor plate materials to evaluate the measurement technologies for the characterization of plutonium, depleted uranium or highly enriched uranium shielded by both nuclear and non-nuclear materials. A series of pulsed photonuclear, time-correlated measurements were performed with unshielded calibration materials and then compared with the more complex composite shield configurations. The measurements used multiple 3He detectors that are designed to detect fission neutrons between pulses of an electron linear accelerator. The accelerator produced 10-MeV bremsstrahlung X-rays at a repetition rate of 125 Hz (8 ms between pulses) with a 4-us pulse width. All inspected objects were positioned on beam centerline and 100 cm from the X-ray source. The time-correlated data was collected in parallel using both a Los Alamos National Laboratory-designed list-mode acquisition system and a commercial multichannel scaler analyzer. A combination of different measurement configurations and data analysis methods enabled the identification of each object. This paper describes the experimental configuration, the ZPPR inspection objects used, and the various measurement and analysis results for each inspected object.

James Jones; Kevin Haskell; Rich Waston; William Geist; Jonathan Thron; Corey Freeman; Martyn Swinhoe; Seth McConchie; Eric Sword; Lee Montierth; John Zabriskie

2011-07-01T23:59:59.000Z

295

Final Report on Isotope Ratio Techniques for Light Water Reactors  

SciTech Connect

The Isotope Ratio Method (IRM) is a technique for estimating the energy or plutonium production in a fission reactor by measuring isotope ratios in non-fuel reactor components. The isotope ratios in these components can then be directly related to the cumulative energy production with standard reactor modeling methods.

Gerlach, David C.; Gesh, Christopher J.; Hurley, David E.; Mitchell, Mark R.; Meriwether, George H.; Reid, Bruce D.

2009-07-01T23:59:59.000Z

296

Gaps at the periphery of N Reactor graphite stack: A photographic survey  

Science Conference Proceedings (OSTI)

Miniature camera techniques were used to examine regions at the perimeter of the N Reactor graphite stack. Extensive coverage in previously unexplored regions was obtained. Gaps at the top of the unit show no visible evidence of graphite growth related distortion or displacement. Debris apparently associated with leaks in the subcritical monitors is observed but there are no indications that the material adversely affects the functions of safety, cooling, or monitoring systems. Gaps inside the right primary shield reveal graphite cooling tubes and other components in undisturbed condition. The only changes since construction are a thin layer of graphite dust from earlier rod channel renovation and oxide films on steel components. In the steam vent between moderator and reflector graphite, displaced and broken graphite blocks are observed. Again, no functional impairment of rod channels or other systems is observed. There is no evidence of distortion or deterioration of the graphite cooling tubes. The surveillance summarized here explores many previously hidden recesses. Some findings were unpredicted but none have given cause to change the safety status of the ball channel or control rod systems or to initiate any new safety enhancement measures. This report establishes a baseline for visual surveillance in perimeter regions of the N Reactor stack if operations resume. 10 refs., 22 figs., 1 tab.

Woodruff, E.M.; Weber, J.R.; Sinclair, R.B.

1988-09-01T23:59:59.000Z

297

Three-dimensional analysis of AP600 standard plant shield building roof  

SciTech Connect

The AP600 passive containment vessel is surrounded by a concrete cylindrical shell covered with a truncated conical roof. This roof supports the passive containment cooling system (PCS) annular tank, shield plate and other nonstructural attachments. When the shield building is subjected to different loading combinations as defined in the Standard Review Plan (SRP), some of the sections in the shield building could experience forces in excess of their design values. This report summarized the three-dimensional finite element analysis that was conducted to review the adequacy of the proposed Westinghouse shield building design. The ANSYS finite element software was utilized to analyze the Shield Building Roof (SBR) under dead, snow, wind, thermal and seismic loadings. A three-dimensional model that included a portion of the shield building cylindrical shell, the conical roof and its attachments, the eccentricities at the cone-cylinder connection and at the compression ring and the PCS tank was developed. Mesh sensitivity studies were conducted to select appropriate element size in the cylinder, cone, near air intakes and in the vicinity of the eccentricities. Also, a study was carried out to correctly idealize the water-structure interaction in the PCS tank. Response spectrum analysis was used to calculate the internal forces at different sections in the SBR under Safe Shutdown Earthquake (SSE). Forty-nine structural modes and twenty sloshing modes were used. Two horizontal components of the SSE together with a vertical component were used. Modal stress resultants were combined taking into account the effects of closely spaced modes. The three earthquake directions were combined by the Square Root of the Sum Squares method. Two load combinations were studied. The load combination that included dead, snow, fluid, thermal and seismic loads was selected to be the most critical. Interaction diagrams for critical sections were developed and used to check the design adequacy. The results demonstrated that provided area of steal on each face of several sections of the AP600 SBR was inadequate. This was also noticed when comparing the total provided area of steel per section, i.e., the area of steel on both faces. The discrepancy between Westinghouse results and these reported herein could have resulted from the different finite element mesh sizes and the assumption used in Westinghouse design.

Greimann, L.; Fanous, F.; Safar, S.; Khalil, A.; Bluhm, D.

1999-06-01T23:59:59.000Z

298

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A neutronic reactor in which neutron moderation is achieved primarily in its reflector is described. The reactor structure consists of a cylindrical central "island" of moderator and a spherical moderating reflector spaced therefrom, thereby providing an annular space. An essentially unmoderated liquid fuel is continuously passed through the annular space and undergoes fission while contained therein. The reactor, because of its small size, is particularly adapted for propulsion uses, including the propulsion of aircraft. (AEC)

Fraas, A.P.; Mills, C.B.

1961-11-21T23:59:59.000Z

299

REACTOR COOLING  

DOE Patents (OSTI)

A nuclear reactor with provisions for selectively cooling the fuel elements is described. The reactor has a plurality of tubes extending throughout. Cylindrical fuel elements are disposed within the tubes and the coolant flows through the tubes and around the fuel elements. The fuel elements within the central portion of the reactor are provided with roughened surfaces of material. The fuel elements in the end portions of the tubes within the reactor are provlded with low conduction jackets and the fuel elements in the region between the central portion and the end portions are provided with smooth surfaces of high heat conduction material.

Quackenbush, C.F.

1959-09-29T23:59:59.000Z

300

General Corrosion and Localized Corrosion of the Drip Shield  

DOE Green Energy (OSTI)

The repository design includes a drip shield (BSC 2004 [DIRS 168489]) that provides protection for the waste package both as a barrier to seepage water contact and a physical barrier to potential rockfall. The purpose of the process-level models developed in this report is to model dry oxidation, general corrosion, and localized corrosion of the drip shield plate material, which is made of Ti Grade 7. This document is prepared according to ''Technical Work Plan For: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]). The models developed in this report are used by the waste package degradation analyses for TSPA-LA and serve as a basis to determine the performance of the drip shield. The drip shield may suffer from other forms of failure such as the hydrogen induced cracking (HIC) or stress corrosion cracking (SCC), or both. Stress corrosion cracking of the drip shield material is discussed in ''Stress Corrosion Cracking of the Drip Shield, the Waste Package Outer Barrier, and the Stainless Steel Structural Material'' (BSC 2004 [DIRS 169985]). Hydrogen induced cracking of the drip shield material is discussed in ''Hydrogen Induced Cracking of Drip Shield'' (BSC 2004 [DIRS 169847]).

F. Hua

2004-09-16T23:59:59.000Z

Note: This page contains sample records for the topic "reactor components shield" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


301

Field Measurement Studies on Shunt Coefficient of Lightning Shield Line  

Science Conference Proceedings (OSTI)

The shunt coefficient of lightning shield line is an important parameter, which needs to be considered when we are designing or reforming the substation grounding network, however, it is difficult to be calculated. This paper is to introduce a method ... Keywords: Short circuit current, Lightning shield line, Shunt coefficient, Shunt impedance

Dechao Li; Jinglu Li; Yuhuan Zhang

2008-10-01T23:59:59.000Z

302

Plasma shield for in-air beam processes  

SciTech Connect

A novel concept/apparatus, the Plasma Shield, is introduced in this paper. The purpose of the Plasma Shield is designed to shield a target object chemically and thermally by engulfing an area subjected to beam treatment with inert plasma. The shield consists of a vortex-stabilized arc that is employed to shield beams and workpiece area of interaction from an atmospheric or liquid environment. A vortex-stabilized arc is established between a beam generating device (laser, ion or electron gun) and a target object. The arc, which is composed of a pure noble gas, engulfs the interaction region and shields it from any surrounding liquids like water or reactive gases. The vortex is composed of a sacrificial gas or liquid that swirls around and stabilizes the arc. The successful Plasma Shield was experimentally established and very high-quality electron beam welding with partial plasma shielding was performed. The principle of the operation and experimental results are discussed in the paper.

Hershcovitch, Ady [Collider-Accelerator Department, Building 901A, Brookhaven National Laboratory, Upton, New York 11973 (United States)

2008-05-15T23:59:59.000Z

303

An Account of Oak Ridge National Laboratory's Thirteen Research Reactors  

Science Conference Proceedings (OSTI)

The Oak Ridge National Laboratory has built and operated 13 nuclear reactors in its 66-year history. The first was the graphite reactor, the world's first operational nuclear reactor, which served as a plutonium production pilot plant during World War II. It was followed by two aqueous-homogeneous reactors and two red-hot molten-salt reactors that were parts of power-reactor development programs and by eight others designed for research and radioisotope production. One of the eight was an all-metal fast burst reactor used for health physics studies. All of the others were light-water cooled and moderated, including the famous swimming-pool reactor that was copied dozens of times around the world. Two of the reactors were hoisted 200 feet into the air to study the shielding needs of proposed nuclear-powered aircraft. The final reactor, and the only one still operating today, is the High Flux Isotope Reactor (HFIR) that was built particularly for the production of californium and other heavy elements. With the world's highest flux and recent upgrades that include the addition of a cold neutron source, the 44-year-old HFIR continues to be a valuable tool for research and isotope production, attracting some 500 scientific visitors and guests to Oak Ridge each year. This report describes all of the reactors and their histories.

Rosenthal, Murray Wilford [ORNL

2009-08-01T23:59:59.000Z

304

Shielding Integral Benchmark Archive and Database (SINBAD)  

Science Conference Proceedings (OSTI)

The Shielding Integral Benchmark Archive and Database (SINBAD) collection of benchmarks was initiated in the early 1990 s. SINBAD is an international collaboration between the Organization for Economic Cooperation and Development s Nuclear Energy Agency Data Bank (OECD/NEADB) and the Radiation Safety Information Computational Center (RSICC) at Oak Ridge National Laboratory (ORNL). SINBAD is a major attempt to compile experiments and corresponding computational models with the goal of preserving institutional knowledge and expertise that need to be handed down to future scientists. SINBAD is also a learning tool for university students and scientists who need to design experiments or gain expertise in modeling and simulation. The SINBAD database is currently divided into three categories fission, fusion, and accelerator benchmarks. Where possible, each experiment is described and analyzed using deterministic or probabilistic (Monte Carlo) radiation transport software.

Kirk, Bernadette Lugue [ORNL; Grove, Robert E [ORNL; Kodeli, I. [International Atomic Energy Agency (IAEA); Sartori, Enrico [ORNL; Gulliford, J. [OECD Nuclear Energy Agency

2011-01-01T23:59:59.000Z

305

Fan-fold shielded electrical leads  

DOE Patents (OSTI)

Disclosed are fan-folded electrical leads made from copper cladded Kapton, for example, with the copper cladding on one side serving as a ground plane and the copper cladding on the other side being etched to form the leads. The Kapton is fan folded with the leads located at the bottom of the fan-folds. Electrical connections are made by partially opening the folds of the fan and soldering, for example, the connections directly to the ground plane and/or the lead. The fan folded arrangement produces a number of advantages, such as electrically shielding the leads from the environment, is totally non-magnetic, and has a very low thermal conductivity, while being easy to fabricate. 3 figs.

Rohatgi, R.R.; Cowan, T.E.

1996-06-11T23:59:59.000Z

306

Fan-fold shielded electrical leads  

DOE Patents (OSTI)

Fan-folded electrical leads made from copper cladded Kapton, for example, with the copper cladding on one side serving as a ground plane and the copper cladding on the other side being etched to form the leads. The Kapton is fan folded with the leads located at the bottom of the fan-folds. Electrical connections are made by partially opening the folds of the fan and soldering, for example, the connections directly to the ground plane and/or the lead. The fan folded arrangement produces a number of advantages, such as electrically shielding the leads from the environment, is totally non-magnetic, and has a very low thermal conductivity, while being easy to fabricate.

Rohatgi, Rajeev R. (Mountain View, CA); Cowan, Thomas E. (Livermore, CA)

1996-01-01T23:59:59.000Z

307

Superconducting magnetic shielding apparatus and method  

DOE Patents (OSTI)

Disclosed is a method and apparatus for providing magnetic shielding around a working volume. The apparatus includes a hollow elongated superconducting shell or cylinder having an elongated low magnetic pinning central portion, and two high magnetic pinning end regions. Transition portions of varying magnetic pinning properties are interposed between the central and end portions. The apparatus further includes a solenoid substantially coextensive with and overlying the superconducting cylinder, so as to be magnetically coupled therewith. The method includes the steps passing a longitudinally directed current through the superconducting cylinder so as to depin magnetic reservoirs trapped in the cylinder. Next, a circumferentially directed current is passed through the cylinder, while a longitudinally directed current is maintained. Depinned magnetic reservoirs are moved to the end portions of the cylinder, where they are trapped.

Clem, J.R.

1982-07-09T23:59:59.000Z

308

Superconducting magnetic shielding apparatus and method  

DOE Patents (OSTI)

Disclosed are a method and apparatus for providing magnetic shielding around a working volume. The apparatus includes a hollow elongated superconducting shell or cylinder having an elongated low magnetic pinning central portion, and two high magnetic pinning end regions. Transition portions of varying magnetic pinning properties are interposed between the central and end portions. The apparatus further includes a solenoid substantially coextensive with and overlying the superconducting cylinder, so as to be magnetically coupled therewith. The method includes the steps passing a longitudinally directed current through the superconducting cylinder so as to depin magnetic reservoirs trapped in the cylinder. Next, a circumferentially directed current is passed through the cylinder, while a longitudinally directed current is maintained. Depinned magnetic reservoirs are moved to the end portions of the cylinder, where they are trapped. 5 figs.

Clem, J.R.; Clem, J.R.

1983-10-11T23:59:59.000Z

309

Superconducting magnetic shielding apparatus and method  

DOE Patents (OSTI)

Disclosed is a method and apparatus for providing magnetic shielding around a working volume. The apparatus includes a hollow elongated superconducting shell or cylinder having an elongated low magnetic pinning central portion, and two high magnetic pinning end regions. Transition portions of varying magnetic pinning properties are interposed between the central and end portions. The apparatus further includes a solenoid substantially coextensive with and overlying the superconducting cylinder, so as to be magnetically coupled therewith. The method includes the steps passing a longitudinally directed current through the superconducting cylinder so as to depin magnetic reservoirs trapped in the cylinder. Next, a circumferentially directed current is passed through the cylinder, while a longitudinally directed current is maintained. Depinned magnetic reservoirs are moved to the end portions of the cylinder, where they are trapped.

Clem, John R. (Ames, IA); Clem, John R. (Ames, IA)

1983-01-01T23:59:59.000Z

310

Shielding of the Target Area of the Fragment Separator Super-FRS  

Science Conference Proceedings (OSTI)

Shielding / Special Issue on the 11th International Conference on Radiation Shielding and the 15th Topical Meeting of the Radiation Protection and Shielding Division (Part 2) / Radiation Protection

T. Radon; E. Kozlova; G. Fehrenbacher; H. Geissel; K. Sümmerer; H. Weick; M. Winkler

311

Evaluation of Radionuclides in Concrete Shielding for Nuclear Power Plant Decommissioning  

Science Conference Proceedings (OSTI)

Shielding / Special Issue on the 11th International Conference on Radiation Shielding and the 15th Topical Meeting of the Radiation Protection and Shielding Division (Part 2) / Decontamination/Decommissioning

Yen-Fu Chen; Yen-Kung Lin; Rong-Jiun Sheu; Shiang-Huei Jiang

312

Experimental Studies of Shielding and Irradiation Effects at High-Energy Accelerator Facilities  

Science Conference Proceedings (OSTI)

Shielding / Special Issue on the 11th International Conference on Radiation Shielding and the 15th Topical Meeting of the Radiation Protection and Shielding Division (Part 2) / Radiation Protection

H. Nakashima; Y. Sakamoto; Y. Iwamoto; N. Matsuda; Y. Kasugai; Y. Nakane; F. Masukawa; N. V. Mokhov; A. F. Leveling; D. J. Boehnlein; K. Vaziri; T. Sanami; H. Matsumura; M. Hagiwara; H. Iwase; N. Kinoshita; H. Hirayama; K. Oishi; T. Nakamura; H. Arakawa; N. Shigyo; K. Ishibashi; H. Yashima; N. Nakao; K. Niita

313

Characteristics of Prompt Radiation Field and Shielding Design for Taiwan Photon Source  

Science Conference Proceedings (OSTI)

Shielding / Special Issue on the 11th International Conference on Radiation Shielding and the 15th Topical Meeting of the Radiation Protection and Shielding Division (Part 2) / Radiation Protection

R. J. Sheu; J. Liu; J. P. Wang; K. K. Lin; G. H. Luo

314

Beam Dump and Local Shielding Layout Around the ITEP Radiation Test Facility  

Science Conference Proceedings (OSTI)

Shielding / Special Issue on the 11th International Conference on Radiation Shielding and the 15th Topical Meeting of the Radiation Protection and Shielding Division (Part 2) / Radiation Protection

Yury Titarenko; Viacheslav Batyaev; Alexey Titarenko; Michael Butko; Kirill Pavlov; Sergey Florya; Roman Tikhonov; Pavel Boyko; Alexey Kovalenko; Nikolai Sobolevsky; Vasily Anashin; Stepan Mashnik; Waclaw Gudowski; Nikolai Mokhov; Igor Rakhno

315

Monte Carlo Application in Shielding Design for an X-Ray Container/Vehicle Inspection System  

Science Conference Proceedings (OSTI)

Shielding / Special Issue on the 11th International Conference on Radiation Shielding and the 15th Topical Meeting of the Radiation Protection and Shielding Division (Part 2) / Radiation Protection

Junli Li; Shenjin Ming; Yanfeng Cao; Yanli Deng

316

Shielding Analysis for a 40-MeV Electron Linac Facility  

Science Conference Proceedings (OSTI)

Shielding / Special Issue on the 11th International Conference on Radiation Shielding and the 15th Topical Meeting of the Radiation Protection and Shielding Division (Part 2) / Radiation Protection

Kai Masuda; Toshiteru Kii; Hideaki Ohgaki; Heishun Zen; Tetsuo Yamazaki

317

Testing of a High-Density Concrete as Neutron Shielding Material  

Science Conference Proceedings (OSTI)

Shielding / Special Issue on the 11th International Conference on Radiation Shielding and the 15th Topical Meeting of the Radiation Protection and Shielding Division (Part 2) / Radiation Protection

Eduardo Gallego; Alfredo Lorente; Héctor René Vega-Carrillo

318

Shielding Calculations for the New Experimental Storage Ring for the Facility for Antiproton and Ion Research  

Science Conference Proceedings (OSTI)

Shielding / Special Issue on the 11th International Conference on Radiation Shielding and the 15th Topical Meeting of the Radiation Protection and Shielding Division (Part 2) / Radiation Protection

K. Vogt; G. Fehrenbacher; A. Knapp; T. Radon

319

Biological Shield Design and Analysis of KIPT Accelerator-Driven Subcritical Facility  

Science Conference Proceedings (OSTI)

Shielding / Special Issue on the 11th International Conference on Radiation Shielding and the 15th Topical Meeting of the Radiation Protection and Shielding Division (PART 3) / Radiation Protection

Zhaopeng Zhong; Yousry Gohar

320

Impact of the Detailed Fission Source Distribution on IRIS Shielding Analyses  

Science Conference Proceedings (OSTI)

Shielding / Special Issue on the 11th International Conference on Radiation Shielding and the 15th Topical Meeting of the Radiation Protection and Shielding Division (Part 2) / Radiation Protection

Fausto Franceschini; Bojan Petrovic

Note: This page contains sample records for the topic "reactor components shield" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


321

HIGHLY SELF-SHIELDED BURNABLE POISONS  

SciTech Connect

A new and different type of burnable poison element which can be used essentially to match any decreasing reactivity curve in a reactor and still have a tolerable residue was studied. Specific application is made to Gd as the poison elements; however, the results would be similar for any other suitable high cross-section element such as Cd. (J.R.D.)

Holl, R.J.

1963-03-01T23:59:59.000Z

322

Decommissioning of the Austrian 10 MW Research Reactor, Results and Lessons learned Paper  

SciTech Connect

After the decision to shut down the 10 MW ASTRA-MTR Research Reactor was reached in May 1998, the possible options and required phases for decommissioning and removal of the radioactive components were evaluated in a decommissioning study. To support the decisions at each phase, an estimate of the activity inventory in the various parts of the reactor and the waste volume to be expected was performed. Of the possible options an immediate dismantling to phase 1 of IAEA Technical Guide Lines after the immediately following, continued dismantling to phase 2 of these guide lines was identified as the most reasonable and under the auspices optimum choice. The actual decommissioning work on the ASTRA-Reactor began in January 2000 after its final shutdown on July 31, 1999. Preliminary evaluations of the activity inventory gave an estimated amount of 320 kg of intermediate level waste, of about 60 metric tons of contaminated and another 100 metric tons of activated low level radioactive waste. The activities were roughly estimated to be at 200 TBq in the intermediate level and 6 GBq in the low level. The structure of the decommissioning process was decided against cost-, time- and risk-optimization following the basic layout of the main tasks, e.g. the removing of the fuel, the recovering and the treatment of the intermediate level activities in the vicinity of the core, the handling and conditioning of the neutron exposed graphite and the Beryllium-elements. As an example, the dismantling of approx. 1400 metric tons of the biological shield is described in more detail from the determination of the dismantling technique to the clearing procedures and the deposition. The process of dismantling of the biological shield is presented in fast motion. The dismantling of the pump-room installations of the primary loop, the processing of the contaminated or activated metals, the dismantling of the ventilation system and the radiological clearance of the reactor building was done under optimized conditions and is explained in the following. Spent fuel was generally delivered to the US Department of Energy - DOE in several shipments over the operational time of the ASTRA reactor. With the last shipment in May 2001 all the remaining spent fuel elements out of the ASTRA reactor consignment were transferred to DOE. To reduce waste from concrete shielding, German regulations Dt.StrSchV, annex IV, table 1, two clearance values referring to 'clearance restricted for permanent deposit' and to a clearance for unrestricted re-use were used. In order to reduce the amount of an estimated 60 tons of slightly contaminated metals, it was determined that introducing re-melting procedures were the most economical way. To obtain radiological clearance of the reactor building, compliance with the release limits according to Austrian Radiation Protection Ordinance had to be proved to the regulatory body. There, in general, the limits for unrestricted release were defined as a maximum dose rate of 10 {mu}Sv effective for an individual person per year. The results of the regular yearly medical examinations of the staff indicated no influence of the work related to decommissioning. The readings of the personal dosimeters over the entire project amounted to a total of 85.6 mSv, averaging to 1.07 mSv per year and person. After finishing the decommissioning process, the material balance showed 89.6 % for unrestricted reuse, 6.6 % for conventional mass-dumping and 3.8 % of ILW and LLW. The project was covered by an extensive documentation. All operations within NES followed ISO 9000 quality insurance standards. Experiences and knowledge were presented to and shared with the community, e.g. AFR and IAEA throughout the project. (authors)

Hillebrand, G.; Meyer, F. [Nuclear Engineering Seibersdorf GmbH (NES), Seibersdorf, Austria, Europe (Austria)

2008-07-01T23:59:59.000Z

323

MODELING HEAT TRANSFER IN SPENT FUEL TRANSFER CASK NEUTRON SHIELDS – A CHALLENGING PROBLEM IN NATURAL CONVECTION  

SciTech Connect

In the United States, commercial spent nuclear fuel is typically moved from spent fuel pools to outdoor dry storage pads within a transfer cask system that provides radiation shielding to protect personnel and the surrounding environment. The transfer casks are cylindrical steel enclosures with integral gamma and neutron radiation shields. Since the transfer cask system must be passively cooled, decay heat removal from spent nuclear fuel canister is limited by the rate of heat transfer through the cask components, and natural convection from the transfer cask surface. The primary mode of heat transfer within the transfer cask system is conduction, but some cask designs incorporate a liquid neutron shield tank surrounding the transfer cask structural shell. In these systems, accurate prediction of natural convection within the neutron shield tank is an important part of assessing the overall thermal performance of the transfer cask system. The large-scale geometry of the neutron shield tank, which is typically an annulus approximately 2 meters in diameter but only 10-15 cm in thickness, and the relatively small scale velocities (typically less than 5 cm/s) represent a wide range of spatial and temporal scales that contribute to making this a challenging problem for computational fluid dynamics (CFD) modeling. Relevant experimental data at these scales are not available in the literature, but some recent modeling studies offer insights into numerical issues and solutions; however, the geometries in these studies, and for the experimental data in the literature at smaller scales, all have large annular gaps that are not prototypic of the transfer cask neutron shield. This paper proposes that there may be reliable CFD approaches to the transfer cask problem, specifically coupled steady-state solvers or unsteady simulations; however, both of these solutions take significant computational effort. Segregated (uncoupled) steady state solvers that were tested did not accurately capture the flow field and heat transfer distribution in this application. Mesh resolution, turbulence modeling, and the tradeoff between steady state and transient solutions are addressed. Because of the critical nature of this application, the need for new experiments at representative scales is clearly demonstrated.

Fort, James A.; Cuta, Judith M.; Bajwa, C.; Baglietto, E.

2010-07-18T23:59:59.000Z

324

Passive heat transfer means for nuclear reactors  

DOE Patents (OSTI)

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

Burelbach, James P. (Glen Ellyn, IL)

1984-01-01T23:59:59.000Z

325

NUCLEAR REACTOR  

DOE Patents (OSTI)

A heterogeneous, natural uranium fueled, solid moderated, gas cooled reactor is described, in which the fuel elements are in the form of elongated rods and are dlsposed within vertical coolant channels ln the moderator symmetrically arranged as a regular lattice in groups. This reactor employs control rods which operate in vertical channels in the moderator so that each control rod is centered in one of the fuel element groups. The reactor is enclosed in a pressure vessel which ls provided with access holes at the top to facilitate loading and unloadlng of the fuel elements, control rods and control rod driving devices.

Moore, R.V.; Bowen, J.H.; Dent, K.H.

1958-12-01T23:59:59.000Z

326

Residual Activation of Accelerator Components  

Science Conference Proceedings (OSTI)

Accelerators / Special Issue on the 11th International Conference on Radiation Shielding and the 15th Topical Meeting of the Radiation Protection and Shielding Division (PART 3) / Radiation Measurements and Instrumentation

I. L. Rakhno; N. V. Mokhov; S. I. Striganov

327

DARHT : integration of shielding design and analysis with facility design /  

SciTech Connect

The design of the interior portions of the Dual Axis Radiographic Hydrodynamic Test (DARHT) Facility incorporated shielding and controls from the beginning of the installation of the Accelerators. The purpose of the design and analysis was to demonstrate the adequacy of shielding or to determine the need for additional shielding or controls. Two classes of events were considered: (1) routine operation defined as the annual production of 10,000 2000-ns pulses of electrons at a nominal energy of 20 MeV, some of which are converted to the x-ray imaging beam consisting of four nominal 60-ns pulses over the 2000-ns time frame, and (2) accident case defined as up to 100 2000-ns pulses of electrons accidentally impinging on some metallic surface, thereby producing x rays. Several locations for both classes of events were considered inside and outside of the accelerator hall buildings. The analysis method consisted of the definition of a source term for each case studied and the definition of a model of the shielding and equipment present between the source and the dose areas. A minimal model of the fixed existing or proposed shielding and equipment structures was used for a first approximation. If the resulting dose from the first approximation was below the design goal (1 rem/yr for routine operations, 5 rem for accident cases), then no further investigations were performed. If the result of the first approximation was above our design goals, the model was refined to include existing or proposed shielding and equipment. In some cases existing shielding and equipment were adequate to meet our goals and in some cases additional shielding was added or administrative controls were imposed to protect the workers. It is expected that the radiation shielding design, exclusion area designations, and access control features, will result in low doses to personnel at the DARHT Facility.

Boudrie, R. L. (Richard L.); Brown, T. H. (Thomas H.); Gilmore, W. E. (Walter E.); Downing, J. N. (James N.), Jr.; Hack, Alan; McClure, D. A. (Donald A.); Nelson, C. A. (Christine A.); Wadlinger, E. Alan; Zumbro, M. V. (Martha V.)

2002-01-01T23:59:59.000Z

328

CONCEPTUAL DESIGN OF THE PEBBLE BED REACTOR EXPERIMENT  

SciTech Connect

The Pebble-Bed Reactor Experiment (PBRE) was designed to advance the pebble-bed concept by providing a test of characteristic features and make contriliutions to the general development of all-ceramic gas-cooled reactors. The following objectives were established for the reactor experiment: to investigate key features of the pebble-bed concept, including on-stream fuel handling, movement of fuel through bed, and performance of core; to obtain operation and maintenance experience with a system contaminated with fission- product activity; and to investigate the behavior of graphite fuel elements. A fourth objective, study of the behavior of core materials at conditions occurring with exit gas temperatures in the range 2000 to 2500 deg F, was tentatively included. The preliminary design oE a 5-Mw(t) reactor for achieving these objectives was prepared. The core of the PBRE is a 2 1/2-ft-diam, 4-ft-tall cylinder containing approximately 12,000 spherical graphite fuel elements 1 1/2 in. in diameter. Fuel spheres are added to and removed from the core by gravity flow, and these operations are performed while the reactor is at power by using pairs of valves for passage of elements into and out of the high-pressure system. Exposed fuel can be recycled to the top of the core. Helium coolant at 500 psia enters the bottom of the core at 550 deg F and emerges from the top at 1250 deg F. Concentric ducting connects the reactor to a single heat exchanger, which is located sufficiently high above the core that natural circulation will suffice to remove afterheat in the event the blower ceases to function. The coolant flow path is such that the entire pressure envelope is swept with helium at the temperature at which it emerges from the heat exchanger. Provision for semi- remote maintenance of contaminated components is emphasized in the layout, and most of the equipment in the primary and auxiliary systems is accessible from above by the removal of modular shielding units. Thc design permits replacement of the entire core graphite structure, The reactor can be adapted for testing core materials at high temperature by attemperation of the hot helium emerging from the core wwiih cool gas in a plenum in the upper graphite structure. Location of the PBRE at the site of the HRE-2 facility is proposed to take advantage of available buildings and services, but the reactor and auxiliary equipment will be contained in a completely new vessel located adjacent to the existing building. The design and direct construction cost of the reactor plant is estimated to be 958,000, allowance for contingencies, overhead, and escalation brings the total to ,260,000. High-temperature operation can be achieved when desired for an additional expenditure of less than 0,000. (auth)

1962-05-17T23:59:59.000Z

329

Radiolysis Concerns for Water Shielding in Fission Surface Power Applications  

Science Conference Proceedings (OSTI)

This paper presents an overview of radiolysis concerns with regard to water shields for fission surface power. A review of the radiolysis process is presented and key parameters and trends are identified. From this understanding of the radiolytic decomposition of water, shield pressurization and corrosion are identified as the primary concerns. Existing experimental and modeling data addressing concerns are summarized. It was found that radiolysis of pure water in a closed volume results in minimal, if any net decomposition, and therefore reduces the potential for shield pressurization and corrosion.

Schoenfeld, Michael P. [NASA Marshall Space Flight Center, ER24, MSFC, AL 35812 (United States); Anghaie, Samim [Innovative Space Power and Propulsion Institute, 800 SW Archer Rd. Bldg.554, P.O. Box 116502, University of Florida, Gainesville, FL 32611-6502 (United States)

2008-01-21T23:59:59.000Z

330

Reactor Materials  

Energy.gov (U.S. Department of Energy (DOE))

The reactor materials crosscut effort will enable the development of innovative and revolutionary materials and provide broad-based, modern materials science that will benefit all four DOE-NE...

331

NUCLEAR REACTOR  

DOE Patents (OSTI)

A nuclear reactor incorporating seed and blanket assemblies is designed. Means are provided for obtaining samples of the coolant from the blanket assemblies and for varying the flow of coolant through the blanket assemblies. (AEC)

Sherman, J.; Sharbaugh, J.E.; Fauth, W.L. Jr.; Palladino, N.J.; DeHuff, P.G.

1962-10-23T23:59:59.000Z

332

NEUTRONIC REACTORS  

DOE Patents (OSTI)

A nuclear reactor is described wherein horizontal rods of thermal- neutron-fissionable material are disposed in a body of heavy water and extend through and are supported by spaced parallel walls of graphite.

Wigner, E.P.

1960-11-22T23:59:59.000Z

333

Elimination of Heat-Shielding for Geothermal Tools Operating Up To 300 Degress Celsius  

DOE Green Energy (OSTI)

This report focuses Sandia National Laboratories' effort to create high-temperature logging tools for geothermal applications not requiring heat-shielding. Tool electronics can operate up to 300 C with a few limiting components operating to 250 C. Second generation electronics are needed to increase measurement accuracy and extend the operating range to 300 and then 350 C are identified. Custom development of high-temperature batteries and assembling techniques are touched on. Outcomes of this work are discussed and new directions for developing high-temperature industry are suggested.

HENFLING,JOSEPH A.; NORMANN,RANDY A.

1999-10-07T23:59:59.000Z

334

NUCLEAR REACTOR  

DOE Patents (OSTI)

High temperature reactors which are uniquely adapted to serve as the heat source for nuclear pcwered rockets are described. The reactor is comprised essentially of an outer tubular heat resistant casing which provides the main coolant passageway to and away from the reactor core within the casing and in which the working fluid is preferably hydrogen or helium gas which is permitted to vaporize from a liquid storage tank. The reactor core has a generally spherical shape formed entirely of an active material comprised of fissile material and a moderator material which serves as a diluent. The active material is fabricated as a gas permeable porous material and is interlaced in a random manner with very small inter-connecting bores or capillary tubes through which the coolant gas may flow. The entire reactor is divided into successive sections along the direction of the temperature gradient or coolant flow, each section utilizing materials of construction which are most advantageous from a nuclear standpoint and which at the same time can withstand the operating temperature of that particular zone. This design results in a nuclear reactor characterized simultaneously by a minimum critiral size and mass and by the ability to heat a working fluid to an extremely high temperature.

Grebe, J.J.

1959-07-14T23:59:59.000Z

335

Electric field shielding in dielectric nanosolutions  

E-Print Network (OSTI)

To gain some insight into electrochemical activity of dielectric colloids of technical and biomedical interest we investigate a model of dielectric nanosolution whose micro-constitution is dominated by dipolarions -- positively and negatively charged spherically symmetric nano-structures composed of ionic charge surrounded by cloud of radially polarized dipoles of electrically neutral molecules of solvent. Combing the standard constitutive equations of an isotropic dielectric liquid with Maxwell equation of electrostatics and presuming the Boltzmann shape of the particle density of bound-charge we derive equation for the in-medium electrostatic field. Particular attention is given to numerical analysis of obtained analytic solutions of this equation describing the exterior fields of dipolarions with dipolar atmospheres of solvent molecules endowed with either permanent or field-induced dipole moments radially polarized by central symmetric field of counterions. The presented computations show that the electric field shielding of dipolarions in dielectric nanosolutions is quite different from that of counterionic nano-complexes of Debye-H\\"uckel theory of electrolytes.

Sergey Bastrukov; Pik-Yin Lai; Irina Molodtsova

2011-09-20T23:59:59.000Z

336

Ultrasonic Signatures of Degradation in Advanced Reactor Materials  

Science Conference Proceedings (OSTI)

The ability to monitor and predict component life augments the safety features of the new reactor concepts. Characterization of materials is key to the life ...

337

Cold worked ferritic alloys and components  

DOE Patents (OSTI)

This invention relates to liquid metal fast breeder reactor and steam generator precipitation hardening fully ferritic alloy components which have a microstructure substantially free of the primary precipitation hardening phase while having cells or arrays of dislocations of varying population densities. It also relates to the process by which these components are produced, which entails solution treating the alloy followed by a final cold working step. In this condition, the first significant precipitation hardening of the component occurs during high temperature use.

Korenko, Michael K. (Wexford, PA)

1984-01-01T23:59:59.000Z

338

Rancho Seco Reactor Vessel Segmentation Experience Report  

Science Conference Proceedings (OSTI)

This report documents the approach taken by Sacramento Municipal Utility District (SMUD) in the segmentation and disposal of the Reactor Vessel from the Rancho Seco Nuclear Generating Station (RSNGS). The location of the Rancho Seco plant placed major constraints on the shipping options available for large plant components (Steam Generators and Reactor Vessel). This report details the engineering evaluations leading to the segmentation and disposal of the Reactor Vessel (RV). It describes the key element...

2008-03-18T23:59:59.000Z

339

Characterization of Fast Reactor Irradiated Stainless Steels  

Science Conference Proceedings (OSTI)

As part of the overall effort to understand the role of different material and environmental variables on irradiation-assisted stress corrosion cracking (IASCC) in light water reactor (LWR) components, the Cooperative IASCC Research (CIR-II) Program has conducted irradiation experiments in the BOR-60 fast reactor near Dimitrovgrad, Russia. This project was a continuation of research on characterization of microstructure and microchemistry of stainless steel heats irradiated in the BOR-60 fast reactor, do...

2008-05-07T23:59:59.000Z

340

DESIGN OF THE ARGONNE LOW POWER REACTOR (ALPR)  

SciTech Connect

A description is given of the design of a prototype "packaged" nuclear power plant. The purpose of the plant is to alleviate fuel oil logistics and storage problems posed by remote auxiliary DEW Line radar statibns north of the Arctic Circle. The ALPR (redesignated SL-1) is a 3 Mwt, heterogeneous, highly enriched uranium- fueled, naturalcirculation boiling water reactor, ccoled and moderated with light water. Steam at 300 psig, dry and saturated (421 deg F) is passed directly from the reactor to a conventional turbine-generator to produce electric power (300 kw nominal) and space-heating (400 kw) requirements consistent with rigid mechanical and structural specifications prescribed by the military, and dictated by the extreme geophysics prevailing at the ultimate site. The over all design criteria emphasize: simplicity and reliability of operation and maintenance, with minimum supervision; minimum on-site construction; maximum use of standard components; limited water supply; utilization of local gravel for biological shielding; transportability by air lift; and nominal 3-year fuel operating lifetime per core loading. The "packaged" concept is incorporated for the initial erection. The plant is not designed for relocation. The design criteria for the prototype necessitate special features. The fuel plates are clad with an alurninurn-nickel alloy (X8001). Burnable-poison (BIO) strips are mechancally attached to the fuel assemblies to compensate the excess reactivity required for a nominal 3-year core operating lifetime. The control rods are actuated by rackand-pinion drive extensions which incorporate rotary seals. Fuel exchange is accomplished without the removal of the pressure vessel head. The electrical power generated is used to operate plant auxiliaries; the "net electric power" is dissipated by resistors. The hot water for space heating is heated in a heat exchanger by 20-psig steam, use being made of the latent heat of vaporization, and all the heat is dissipated by a finned-tube, air-cooled heat exchanger. (auth)

Grant, N.R.; Hamer, E.E.; Hooker, H.H.; Jorgensen, G.L.; Kann, W.J.; Lipinski, W.C.; Milak, G.C.; Rossin, A.D.; Shaftman, D.H.; Smaardyk, A.; Treshow, M.

1961-05-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor components shield" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


341

Optimal Magnetic Shield Design with Second–Order Cone ...  

E-Print Network (OSTI)

... 106-8569 Japan. (tsuchiya@sun312.ism.ac.jp). 1 ... In this paper, a real-world optimal magnetic shielding design problem which arises in the development of ...

342

Experimental Test of Self-Shielding in VUV Photodissociation...  

NLE Websites -- All DOE Office Websites (Extended Search)

Experimental Test of Self-Shielding in VUV Photodissociation of CO Print One way to test models of the solar system's formation is to compare the isotopic abundances of the...

343

Innovative Heat Shield Structure Completed for Orion - Materials ...  

Science Conference Proceedings (OSTI)

Posted on: 3/1/2010 12:00:00 AM... Lockheed Martin announced on March 1 that it has completed fabrication of the world's largest heat shield structure.

344

Technology Development Program for an Advanced Potassium Rankine Power Conversion System Compatible with Several Space Reactor Designs  

SciTech Connect

This report documents the work performed during the first phase of the National Aeronautics and Space Administration (NASA), National Research Announcement (NRA) Technology Development Program for an Advanced Potassium Rankine Power Conversion System Compatible with Several Space Reactor Designs. The document includes an optimization of both 100-kW{sub e} and 250-kW{sub e} (at the propulsion unit) Rankine cycle power conversion systems. In order to perform the mass optimization of these systems, several parametric evaluations of different design options were investigated. These options included feed and reheat, vapor superheat levels entering the turbine, three different material types, and multiple heat rejection system designs. The overall masses of these Nb-1%Zr systems are approximately 3100 kg and 6300 kg for the 100- kW{sub e} and 250-kW{sub e} systems, respectively, each with two totally redundant power conversion units, including the mass of the single reactor and shield. Initial conceptual designs for each of the components were developed in order to estimate component masses. In addition, an overall system concept was presented that was designed to fit within the launch envelope of a heavy lift vehicle. A technology development plan is presented in the report that describes the major efforts that are required to reach a technology readiness level of 6. A 10-year development plan was proposed.

Yoder, G.L.

2005-10-03T23:59:59.000Z

345

A 2-D Test Problem for CFD Modeling Heat Transfer in Spent Fuel Transfer Cask Neutron Shields  

SciTech Connect

In the United States, commercial spent nuclear fuel is typically moved from spent fuel pools to outdoor dry storage pads within a transfer cask system that provides radiation shielding to protect personnel and the surrounding environment. The transfer casks are cylindrical steel enclosures with integral gamma and neutron radiation shields. Since the transfer cask system must be passively cooled, decay heat removal from spent nuclear fuel canister is limited by the rate of heat transfer through the cask components, and natural convection from the transfer cask surface. The primary mode of heat transfer within the transfer cask system is conduction, but some cask designs incorporate a liquid neutron shield tank surrounding the transfer cask structural shell. In these systems, accurate prediction of natural convection within the neutron shield tank is an important part of assessing the overall thermal performance of the transfer cask system. The large-scale geometry of the neutron shield tank, which is typically an annulus approximately 2 meters in diameter but only 5-10 cm in thickness, and the relatively small scale velocities (typically less than 5 cm/s) represent a wide range of spatial and temporal scales that contribute to making this a challenging problem for computational fluid dynamics (CFD) modeling. Relevant experimental data at these scales are not available in the literature, but some recent modeling studies offer insights into numerical issues and solutions; however, the geometries in these studies, and for the experimental data in the literature at smaller scales, all have large annular gaps that are not prototypic of the transfer cask neutron shield. This paper presents results for a simple 2-D problem that is an effective numerical analog for the neutron shield application. Because it is 2-D, solutions can be obtained relatively quickly allowing a comparison and assessment of sensitivity to model parameter changes. Turbulence models are considered as well as the tradeoff between steady state and transient solutions. Solutions are compared for two commercial CFD codes, FLUENT and STAR-CCM+. The results can be used to provide input to the CFD Best Practices for this application. Following study results for the 2-D test problem, a comparison of simulation results is provided for a high Rayleigh number experiment with large annular gap. Because the geometry of this validation is significantly different from the neutron shield, and due to the critical nature of this application, the argument is made for new experiments at representative scales

Zigh, Ghani; Solis, Jorge; Fort, James A.

2011-01-14T23:59:59.000Z

346

Nuclear Reactors and Technology  

SciTech Connect

This publication 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 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 the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, 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.

1992-01-01T23:59:59.000Z

347

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.

1997-03-01T23:59:59.000Z

348

REACTOR DEVELOPMENT PROGRAM PROGRESS REPORT, FEBRUARY 1962  

SciTech Connect

Progress is reported on EBWR, BORAX-V, and development of liquid metal cooled reactors including EBR-I and -II. Developments in general reactor technology are reported in sections on physics, fuels, components, materials, engineering, and chemical separations. Other research and development is reported in advanced systems and nuclear ssfety. (J.R.D.)

1962-02-01T23:59:59.000Z

349

Noise modeling from high-permeability shields using Kirchhoff equations  

SciTech Connect

Progress in the development of high-sensitivity magnetic-field measurements has stimulated interest in understanding magnetic noise of conductive materials, especially of magnetic shields (DC or rf) based on high-permeability materials and/or high-conductivity materials. For example, SQUIDs and atomic magnetometers have been used in many experiments with mu-metal shields, and additionally SQUID systems frequently have rf shielding based on thin conductive materials. Typical existing approaches to modeling noise only work with simple shield and sensor geometries while common experimental setups today consist of multiple sensor systems arbitrary shapes and complex shield geometries. With complex sensor arrays used in, for example, MEG and Ultra Low Field MRI studies the knowledge of the noise correlation between sensors is as important as the knowledge of the noise itself. This is crucial for incorporating efficient noise cancelation schemes for the system. We developed an approach that allows us to calculate the Johnson noise for any geometrically shaped shield and multiple sensor systems. The approach uses a fraction of the processing power of other approaches and with a multiple sensor system our approach not only calculates the noise for each sensor but it also calculates the noise correlation matrix between sensors. Here we will show the algorithm and examples where it can be implemented.

Sandin, Henrik J [Los Alamos National Laboratory; Volegov, Petr L [Los Alamos National Laboratory; Espy, Michelle A [Los Alamos National Laboratory; Matlashov, Andrei N [Los Alamos National Laboratory; Savukov, Igor M [Los Alamos National Laboratory; Schultz, Larry J [Los Alamos National Laboratory

2010-01-01T23:59:59.000Z

350

Shielding from instantaneously and adiabatically applied potential wells in collisionless plasmas  

E-Print Network (OSTI)

Shielding from instantaneously and adiabatically applied potential wells in collisionless plasmas A shielding'' . Experiments demonstrate that when a test potential well is applied to a one-dimensional pure of shielding implicitly or explicitly assume such collisions. Yet it is well known that shielding in plasmas

California at Berkeley, University of

351

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Initial Assessment of Thermal Annealing Needs and Challenges Initial Assessment of Thermal Annealing Needs and Challenges Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Initial Assessment of Thermal Annealing Needs and Challenges The most life-limiting structural component in light-water reactors (LWR) is the reactor pressure vessel (RPV) because replacement of the RPV is not considered a viable option at this time. LWR licenses are now being extended from 40y to 60y by the U.S. Nuclear Regulatory Commission (NRC) with intentions to extend licenses to 80y and beyond. The RPV materials exhibit varying degrees of sensitivity to irradiation-induced embrittlement (decreased toughness) , as shown in Fig. 1.1, and extending operation from 40y to 80y implies a doubling of the neutron exposure for the RPV. Thus,

352

Brookhaven Medical Research Reactor  

NLE Websites -- All DOE Office Websites (Extended Search)

Medical Research Reactor BMRR The last of the Lab's reactors, the Brookhaven Medical Research Reactor (BMRR), was shut down in December 2000. The BMRR was a three megawatt...

353

NEUTRONIC REACTOR  

DOE Patents (OSTI)

This patent relates to neutronic reactors of the heterogeneous water cooled type, and in particular to a fuel element charging and discharging means therefor. In the embodiment illustrated the reactor contains horizontal, parallel coolant tubes in which the fuel elements are disposed. A loading cart containing a magnzine for holding a plurality of fuel elements operates along the face of the reactor at the inlet ends of the coolant tubes. The loading cart is equipped with a ram device for feeding fuel elements from the magazine through the inlot ends of the coolant tubes. Operating along the face adjacent the discharge ends of the tubes there is provided another cart means adapted to receive irradiated fuel elements as they are forced out of the discharge ends of the coolant tubes by the incoming new fuel elements. This cart is equipped with a tank coataining a coolant, such as water, into which the fuel elements fall, and a hydraulically operated plunger to hold the end of the fuel element being discharged. This inveation provides an apparatus whereby the fuel elements may be loaded into the reactor, irradiated therein, and unloaded from the reactor without stopping the fiow of the coolant and without danger to the operating personnel.

Ohlinger, L.A.; Wigner, E.P.; Weinberg, A.M.; Young, G.J.

1958-09-01T23:59:59.000Z

354

Structural Calculations of Drip Shield Exposed to Vibratory Ground Motion  

Science Conference Proceedings (OSTI)

The objective of this calculation is twofold. First, to determine whether or not separation of interlocking drip shield (DS) segments occurs during vibratory ground motion. Second, if DS separation does not occur, to estimate the area of the DS for which the residual 1st principal stress exceeds a certain limit. (The area of DS plate-1 and DS plate-2 [see Attachment I] where the residual 1st principal stress exceeds a certain limit will be, for brevity, referred to as ''the damaged area'' throughout this document; also, DS plate-1 and DS plate-2 will be referred to, for brevity, as ''DS plates'' henceforth.) The stress limit used throughout this document is defined as 50 percent of yield strength of the DS plate material, Titanium Grade 7 (Ti-7) (SB-265 R52400), at temperature of 150 C. A set of 15 calculations is performed at two different annual frequencies of occurrence (annual exceedance frequency): 10{sup -6} per year (1/yr) and 10{sup -7} 1/yr . (Note: Due to computational problems only five realizations at 10{sup -7} 1/yr are presented in this document.) Additionally, one calculation is performed at the annual frequency of occurrence of 5 {center_dot} 10{sup -4} 1/yr. The scope of this document is limited to reporting whether or not the DS separation occurs. If the DS separation does not occur the scope is limited to reporting the calculation results in terms of the damaged area. All these results are evaluated for the DS plates. This calculation is intended for use in support of the Total System Performance Assessment-License Application seismicity modeling. This calculation is associated with the DS design and was performed by the Waste Package Design group. AP-3.12Q, ''Design Calculations and Analyses'' (Ref. 1) is used to perform the calculation and develop the document. The DS is classified as Quality Level 1 (Ref. 5, p. 7). Therefore, this calculation is subject to the Quality Assurance Requirements and Description (Ref. 4). The information provided by Attachment I is that of the potential design of the type of DS considered in this calculation, and provides the potential dimensions and materials for the DS design. Designs of the 21-PWR (Pressurized Water Reactor) waste package (WP) and emplacement pallet (pallet, for brevity, throughout the document) used in this calculation are those defined in References 24 and 22, respectively. All obtained results are valid for these designs only.

S. Mastilovic

2003-06-16T23:59:59.000Z

355

REACTOR CONTROL  

DOE Patents (OSTI)

A control system employed with a high pressure gas cooled reactor in which a control rod is positioned for upward and downward movement into the neutron field from a position beneath the reactor is described. The control rod is positioned by a coupled piston cylinder releasably coupled to a power drive means and the pressurized coolant is directed against the lower side of the piston. The coolant pressure is offset by a higher fiuid pressure applied to the upper surface of the piston and means are provided for releasing the higher pressure on the upper side of the piston so that the pressure of the coolant drives the piston upwardly, forcing the coupled control rod into the ncutron field of the reactor. (AEC)

Fortescue, P.; Nicoll, D.

1962-04-24T23:59:59.000Z

356

NUCLEAR REACTOR  

DOE Patents (OSTI)

A nuclear reactor of the homogeneous liquid fuel type is described wherein the fissionable isotope is suspended or dissolved in a liquid moderator such as water. The reactor core is comprised essentially of a spherical vessel for containing the reactive composition surrounded by a reflector, preferably of beryllium oxide. The reactive composition may be an ordinary water solution of a soluble salt of uranium, the quantity of fissionable isotope in solution being sufficient to provide a critical mass in the vessel. The liquid fuel is stored in a tank of non-crtttcal geometry below the reactor vessel and outside of the reflector and is passed from the tank to the vessel through a pipe connecting the two by air pressure means. Neutron absorbing control and safety rods are operated within slots in the reflector adjacent to the vessel.

Christy, R.F.

1958-07-15T23:59:59.000Z

357

Spherical torus fusion reactor  

DOE Patents (OSTI)

A fusion reactor is provided having a near spherical-shaped plasma with a modest central opening through which straight segments of toroidal field coils extend that carry electrical current for generating a toroidal magnet plasma confinement fields. By retaining only the indispensable components inboard of the plasma torus, principally the cooled toroidal field conductors and in some cases a vacuum containment vessel wall, the fusion reactor features an exceptionally small aspect ratio (typically about 1.5), a naturally elongated plasma cross section without extensive field shaping, requires low strength magnetic containment fields, small size and high beta. These features combine to produce a spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost.

Peng, Yueng-Kay M. (Oak Ridge, TN)

1989-01-01T23:59:59.000Z

358

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.)

1991-01-01T23:59:59.000Z

359

MATERIALS TESTING REACTOR PROJECT. QUARTERLY REPORT FOR PERIOD ENDING MARCH 1, 1950  

SciTech Connect

Progress is reported in finaiizing basic design data for the Materials Testing Reactor. The major emphasis at ANL was on issurance of design reports on practically all phases of the MTR project outside the reactor face and low the first fioor level. Operation of the mock-up reacr at ORNL at 10 watts resulted in no major design changes. Topics discussed include the reactor building, wing, and reactor service building; canal and canal facilities; water systems; air exhaust systems; electrical power systems; effluent control; and shielding requirements. 11 drawings. (C.H.)

Huffman, J.R.

1958-10-31T23:59:59.000Z

360

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

SciTech Connect

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

1961-05-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor components shield" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


361

NUCLEAR REACTOR  

DOE Patents (OSTI)

This patent covers a power-producing nuclear reactor in which fuel rods of slightly enriched U are moderated by heavy water and cooled by liquid metal. The fuel rods arranged parallel to one another in a circle are contained in a large outer closed-end conduit that extends into a tank containing the heavy water. Liquid metal is introduced into the large conduit by a small inner conduit that extends within the circle of fuel rods to a point near the lower closed end of the outer conduit. (AEC) Production Reactors

Young, G.

1963-01-01T23:59:59.000Z

362

An analysis of radionuclide behavior in water pools during accidents at the Annular Core Research Reactor  

SciTech Connect

Physical and chemical phenomena that will affect the behavior of radionuclides released from fuel in the Annular Core Research Reactor during a hypothetical, core disruptive accident are described. The phenomena include boiling of water on heated clad, metal-water reactions, vapor nucleation to form aerosol particles, coagulation of aerosol particles, aerosol deposition within bubbles rising through the shield pool, vapor dissolution in the shield pool, and revaporization of radionuclides from the shield pool. A model of these phenomena is developed and applied to predict the release of radionuclides to the confinement building of the Annular Core Research Reactor. It is found that the shield pool provides overall decontamination factors for particulate of about 2.8 {times} 10{sup 5} and decontamination factors for noble gases of about 2.5--3.7. These results are found to be sensitive to the predicted clad temperature and bubble behavior in the shield pool. Slow revalorization of krypton, xenon and iodine from the shield pool is shown to create a prolonged, low-intensity source term of radioactive material to the confinement atmosphere.

Powers, D.A.

1992-08-01T23:59:59.000Z

363

Conceptual Engineering Method for Attenuating He Ion Interactions on First Wall Components in the Fusion Test Facility (FTF) Employing a Low-Pressure Noble Gas  

SciTech Connect

It has been shown that post detonation energetic helium ions can drastically reduce the useful life of the (dry) first wall of an IFE reactor due to the accumulation of implanted helium. For the purpose of attenuating energetic helium ions from interacting with first wall components in the Fusion Test Facility (FTF) target chamber, several concepts have been advanced. These include magnetic intervention (MI), deployment of a dynamically moving first wall, use of a sacrificial shroud, designing the target chamber large enough to mitigate the damage caused by He ions on the target chamber wall, and the use of a low pressure noble gas resident in the target chamber during pulse power operations. It is proposed that employing a low-pressure (~ 1 torr equivalent) noble gas in the target chamber will thermalize energetic helium ions prior to interaction with the wall. The principle benefit of this concept is the simplicity of the design and the utilization of (modified) existing technologies for pumping and processing the noble ambient gas. Although the gas load in the system would be increased over other proposed methods, the use of a "gas shield" may provide a cost effective method of greatly extending the first wall of the target chamber. An engineering study has been initiated to investigate conceptual engineering metmethods for implementing a viable gas shield strategy in the FTF.

C.A.Gentile, W.R.Blanchard, T.Kozub, C.Priniski, I.Zatz, S.Obenschain

2009-09-21T23:59:59.000Z

364

SPECIAL TOPIC INTERNATIONAL TOKAMAK REACTOR -  

E-Print Network (OSTI)

(beyond the present generation of large tokamaks) in the world tokamak programme; (2) to assess experiment in the tokamak programme between the present generation of large tokamaks (TFTR, JET, JT-60, T-15 of the development and integration of full-scale components which can be extra- polated to a commercial reactor; (d

Abdou, Mohamed

365

Light Water Reactor Sustainability Technical Documents | Department of  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Initiatives » Nuclear Reactor Technologies » Light Water Reactor Initiatives » Nuclear Reactor Technologies » Light Water Reactor Sustainability Program » Light Water Reactor Sustainability Technical Documents Light Water Reactor Sustainability Technical Documents September 30, 2011 Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Initial Assessment of Thermal Annealing Needs and Challenges The most life-limiting structural component in light-water reactors (LWR) is the reactor pressure vessel (RPV) because replacement of the RPV is not considered a viable option at this time. LWR licenses are now being extended from 40y to 60y by the U.S. Nuclear Regulatory Commission (NRC) with intentions to extend licenses to 80y and beyond. The RPV materials exhibit varying degrees of sensitivity to irradiation-induced embrittlement

366

Power Burst Facility (PBF) Reactor Reactor Decommissioning  

NLE Websites -- All DOE Office Websites (Extended Search)

Reactor Decommissioning Click here to view Click here to view Reactor Decommissioning Click on an image to enlarge A crane removes the reactor vessel from the Power Burst Facility...

367

Hydrogen-Induced Cracking of the Drip Shield  

DOE Green Energy (OSTI)

Hydrogen-induced cracking is characterized by the decreased ductility and fracture toughness of a material due to the absorption of atomic hydrogen in the metal crystal lattice. Corrosion is the source of hydrogen generation. For the current design of the engineered barrier without backfill, hydrogen-induced cracking may be a concern because the titanium drip shield can be galvanically coupled to rock bolts (or wire mesh), which may fall onto the drip shield, thereby creating conditions for hydrogen production by electrochemical reaction. The purpose of this report is to analyze whether the drip shield will fail by hydrogen-induced cracking under repository conditions within 10,000 years after emplacement. Hydrogen-induced cracking is a scenario of premature failure of the drip shield. This report develops a realistic model to assess the form of hydrogen-induced cracking degradation of the drip shield under the hydrogen-induced cracking. The scope of this work covers the evaluation of hydrogen absorbed due to general corrosion and galvanic coupling to less noble metals (e.g., Stainless Steel Type 316 and carbon steels) under the repository conditions during the 10,000-year regulatory period after emplacement and whether the absorbed hydrogen content will exceed the critical hydrogen concentration value, above which the hydrogen-induced cracking is assumed to occur. This report also provides the basis for excluding the features, events, and processes (FEPs) related to hydrogen-induced cracking of the drip shield with particular emphasis on FEP 2.1.03.04.OB, hydride cracking of drip shields (DTN: M00407SEPFEPLA.000 [DIRS 170760]). This report is prepared according to ''Technical Work Plan (TWP) for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 169944]).

F. Hua

2004-09-07T23:59:59.000Z

368

Materials and Components Technology Division research summary, 1991  

Science Conference Proceedings (OSTI)

This division has the purpose of providing a R and D capability for design, fabrication, and testing of high-reliability materials, components, and instrumentation. Current divisional programs are in support of the Integral Fast Reactor, life extension for light water reactors, fuels development for the new production reactor and research and test reactors, fusion reactor first-wall and blanket technology, safe shipment of hazardous materials, fluid mechanics/materials/instrumentation for fossile energy systems, and energy conservation and renewables (including tribology, high- temperature superconductivity). Separate abstracts have been prepared for the data base.

Not Available

1991-04-01T23:59:59.000Z

369

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A reactor is described comprising a plurality of horizontal trays containing a solution of a fissionable material, the trays being sleeved on a vertical tube which contains a vertically-reciprocable control rod, a gas-tight chamber enclosing the trays, and means for conducting vaporized moderator from the chamber and for replacing vaporized moderator in the trays. (AEC)

Wigner, E.P.

1962-12-25T23:59:59.000Z

370

NUCLEAR REACTOR  

DOE Patents (OSTI)

A nuclear reactor is described that includes spaced vertical fuel elements centrally disposed in a pressure vessel, a mass of graphite particles in the pressure vessel, means for fluidizing the graphite particles, and coolant tubes in the pressure vessel laterally spaced from the fuel elements. (AEC)

Post, R.G.

1963-05-01T23:59:59.000Z

371

NUCLEAR REACTOR  

DOE Patents (OSTI)

This patent relates to a combination useful in a nuclear reactor and is comprised of a casing, a mass of graphite irapregnated with U compounds in the casing, and at least one coolant tube extending through the casing. The coolant tube is spaced from the mass, and He is irtroduced irto the space between the mass and the coolant tube. (AEC)

Starr, C.

1963-01-01T23:59:59.000Z

372

NEUTRONIC REACTOR  

DOE Patents (OSTI)

BS>A reactor cooled by water, biphenyl, helium, or other fluid with provision made for replacing the fuel rods with the highest plutonium and fission product content without disassembling the entire core and for promptly cooling the rods after their replacement in order to prevent build-up of heat from fission product activity is described.

Creutz, E.C.; Ohlinger, L.A.; Weinberg, A.M.; Wigner, E.P.; Young, G.J.

1959-10-27T23:59:59.000Z

373

NEUTRONIC REACTORS  

DOE Patents (OSTI)

The design of control rods for nuclear reactors are described. In this design the control rod consists essentially of an elongated member constructed in part of a neutron absorbing material and having tube means extending therethrough for conducting a liquid to cool the rod when in use.

Anderson, H.L.

1958-10-01T23:59:59.000Z

374

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

2000-11-22T23:59:59.000Z

375

REACTOR DEVELOPMENT PROGRAM PROGRESS REPORT, SEPTEMBER 1961  

SciTech Connect

BS>Data from examination of blade-type control rods which were used in BORAX are discussed. Operation and maintenance of EBWR is outlined. In work on Borax V, modifications for easier installation of reactor and components is outlined followed by discussion of superheat fuel element development, and fabrication of various reactor components. Borax reactor design is also reported along with information on development and testing. In research on sodiumcooled reactors, activities are summarized in the LPR III and LPR IV programs along with developmental work on EBR I and II. Studies on reactor safety are reported and activities in a program of nuclear technology and general support are outlined. (J.R.D.)

1961-10-15T23:59:59.000Z

376

Hanford Waste Treatment Plant Sets Massive Protective Shield door in  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Waste Treatment Plant Sets Massive Protective Shield door Waste Treatment Plant Sets Massive Protective Shield door in Pretreatment Facility Hanford Waste Treatment Plant Sets Massive Protective Shield door in Pretreatment Facility January 12, 2011 - 12:00pm Addthis The carbon steel doors come together to form an upside-down L-shape. The 102-ton door was set on top of the 85-ton door that was installed at the end of December. The carbon steel doors come together to form an upside-down L-shape. The 102-ton door was set on top of the 85-ton door that was installed at the end of December. The 102-ton shield door measures 52 feet wide and 15 feet tall The 102-ton shield door measures 52 feet wide and 15 feet tall The carbon steel doors come together to form an upside-down L-shape. The 102-ton door was set on top of the 85-ton door that was installed at the end of December.

377

A Historically Significant Shield for In Vivo Measurements  

SciTech Connect

Due to the ubiquitous nature of ionizing radiation, in vivo measurement systems designed to measure low levels of radionuclides in people are usually enclosed within a high density shield. Lead, steel, earth, and water are just some of the materials that have been and are being used to shield the detectors from radiations of cosmic, atmospheric, and terrestrial origin. At many Department of Energy sites, the counting room shields are constructed of pre-world War II steel to reduce the background levels to achieve measurements with low minimum detectable activities (MDA). This is one example of what is commonly called low background steel in the in vivo industry vernacular. The name arises from the fact the steel was manufactured prior to the beginning of atmospheric testing of nuclear weapons in the 1940s. Consequently, the steel is not likely to be contaminated with fission or activation products from fallout. For high energy photons (600 keV shielding significantly reduces the background levels. This is the story "swords-to-plowshare" of the unique steel that now forms a shielded room used at the In Vivo Radioassay and Research Facility (IVRRF) in Richland, Washington.

Lynch, Timothy P.

2007-08-01T23:59:59.000Z

378

Pulse-echo ultrasonic inspection system for in-situ nondestructive inspection of Space Shuttle RCC heat shields.  

SciTech Connect

The reinforced carbon-carbon (RCC) heat shield components on the Space Shuttle's wings must withstand harsh atmospheric reentry environments where the wing leading edge can reach temperatures of 3,000 F. Potential damage includes impact damage, micro cracks, oxidation in the silicon carbide-to-carbon-carbon layers, and interlaminar disbonds. Since accumulated damage in the thick, carbon-carbon and silicon-carbide layers of the heat shields can lead to catastrophic failure of the Shuttle's heat protection system, it was essential for NASA to institute an accurate health monitoring program. NASA's goal was to obtain turnkey inspection systems that could certify the integrity of the Shuttle heat shields prior to each mission. Because of the possibility of damaging the heat shields during removal, the NDI devices must be deployed without removing the leading edge panels from the wing. Recently, NASA selected a multi-method approach for inspecting the wing leading edge which includes eddy current, thermography, and ultrasonics. The complementary superposition of these three inspection techniques produces a rigorous Orbiter certification process that can reliably detect the array of flaws expected in the Shuttle's heat shields. Sandia Labs produced an in-situ ultrasonic inspection method while NASA Langley developed the eddy current and thermographic techniques. An extensive validation process, including blind inspections monitored by NASA officials, demonstrated the ability of these inspection systems to meet the accuracy, sensitivity, and reliability requirements. This report presents the ultrasonic NDI development process and the final hardware configuration. The work included the use of flight hardware and scrap heat shield panels to discover and overcome the obstacles associated with damage detection in the RCC material. Optimum combinations of custom ultrasonic probes and data analyses were merged with the inspection procedures needed to properly survey the heat shield panels. System features were introduced to minimize the potential for human factors errors in identifying and locating the flaws. The in-situ NDI team completed the transfer of this technology to NASA and USA employees so that they can complete 'Return-to-Flight' certification inspections on all Shuttle Orbiters prior to each launch.

Roach, Dennis Patrick; Walkington, Phillip D.; Rackow, Kirk A.

2005-06-01T23:59:59.000Z

379

TEST-HOLE CONSTRUCTION FOR A NEUTRONIC REACTOR  

DOE Patents (OSTI)

Test-hole construction is described for a reactor which provides safe and ready access to the neutron flux region for specimen materials which are to be irradiated therein. An elongated tubular thimble adapted to be inserted in the access hole through the wall of the reactor is constructed of aluminum and is provided with a plurality of holes parallel to the axis of the thimble for conveying the test specimens into position for irradiation, and a conduit for the circulation of coolant. A laminated shield formed of alternate layers of steel and pressed wood fiber is disposed lengthwise of the thimble near the outer end thereof.

Ohlinger, L.A.; Seitz, F.; Young, G.J.

1959-02-17T23:59:59.000Z

380

Space-reactor electric systems: subsystem technology assessment  

SciTech Connect

This report documents the subsystem technology assessment. For the purpose of this report, five subsystems were defined for a space reactor electric system, and the report is organized around these subsystems: reactor; shielding; primary heat transport; power conversion and processing; and heat rejection. The purpose of the assessment was to determine the current technology status and the technology potentials for different types of the five subsystems. The cost and schedule needed to develop these potentials were estimated, and sets of development-compatible subsystems were identified.

Anderson, R.V.; Bost, D.; Determan, W.R.

1983-03-29T23:59:59.000Z

Note: This page contains sample records for the topic "reactor components shield" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


381

Custom Components - Microsystems Science, Technology, and Components  

NLE Websites -- All DOE Office Websites (Extended Search)

Connectors area will help you optimize your choice of connectors for your requirements Passive RF Components Our Passive RF Components area will work with you to identify, specify...

382

THE RADIATION SAFETY INFORMATION COMPUTATIONAL CENTER: A RESOURCE FOR REACTOR DOSIMETRY SOFTWARE AND NUCLEAR DATA  

Science Conference Proceedings (OSTI)

The Radiation Safety Information Computational Center (RSICC) was established in 1963 to collect and disseminate computational nuclear technology in the form of radiation transport, shielding and safety software and corresponding nuclear cross sections. Approximately 1700 nuclear software and data packages are in the RSICC collection, and the majority are applicable to reactor dosimetry.

Kirk, Bernadette Lugue [ORNL

2009-01-01T23:59:59.000Z

383

Compact reversed-field pinch reactors (CRFPR)  

Science Conference Proceedings (OSTI)

The unique confinement properties of the Reversed-Field Pinch (RFP) are exploited to examine physics and technical issues related to a compact, high-power-density fusion reactor. This resistive-coil, steady-state, toroidal device would use a dual-media power cycle driven by a fusion power core (FPC, i.e., plasma chamber, first wall, blanket, shield, and coils) with a power density and mass approaching values characteristic of pressurized-water fission rectors. A 1000-MWe(net) base case is selected from a comprehensive trade-off study to examine technological issues related to operating a high-power-density FPC. After describing the main physics and technology issues for this base-case reactor, directions for future study are suggested.

Krakowski, R.A.; Miller, R.L.; Bathke, C.G.; Hagenson, R.L.; Copenhaver, C.; Werley, K.A.

1986-01-01T23:59:59.000Z

384

Exhaust gas reactor  

SciTech Connect

A reactor for the oxidation of unburned and partially burned components in the exhaust gas from an internal combustion engine comprising a chamber which is substantially circular in cross sections perpendicular to its axis, one or more inlet pipes which pass a mixture of exhaust gas and air substantially tangentially into the chamber near to one end thereof, and an outlet pipe near to the other end of the chamber and which is so arranged that exhaust gas leaves the chamber substantially tangentially. The tangential inlet and tangential outlet of gases minimizes energy losses in the gas passing through the reactor. The ratio of the cross-sectional areas of the inlet pipe(s) to reactor chamber is preferably from 1:9 to 25:36, and similar ranges of crosssectional area ratios are preferred for the outlet pipe and chamber. The ratio of the length of the reaction chamber to diameter is preferably from 1:1 to 4:1. The chamber may be cylindrical or divergent from inlet end to outlet end and may be thermally insulated.

Camarsa, M.; Cocchiara, F.; Garcea, G.P.

1981-11-24T23:59:59.000Z

385

REACTOR UNLOADING  

DOE Patents (OSTI)

This patent is related to gas cooled reactors wherein the fuel elements are disposed in vertical channels extending through the reactor core, the cooling gas passing through the channels from the bottom to the top of the core. The invention is a means for unloading the fuel elements from the core and comprises dump values in the form of flat cars mounted on wheels at the bottom of the core structure which support vertical stacks of fuel elements. When the flat cars are moved, either manually or automatically, for normal unloading purposes, or due to a rapid rise in the reproduction ratio within the core, the fuel elements are permtted to fall by gravity out of the core structure thereby reducing the reproduction ratio or stopping the reaction as desired.

Leverett, M.C.

1958-02-18T23:59:59.000Z

386

NUCLEAR REACTOR  

DOE Patents (OSTI)

A neuclear reactor is described of the heterogeneous type and employing replaceable tubular fuel elements and heavy water as a coolant and moderator. A pluraltty of fuel tubesa having their axes parallel, extend through a tank type pressure vessel which contatns the liquid moderator. The fuel elements are disposed within the fuel tubes in the reaetive portion of the pressure vessel during normal operation and the fuel tubes have removable plug members at each end to permit charging and discharging of the fuel elements. The fuel elements are cylindrical strands of jacketed fissionable material having helical exterior ribs. A bundle of fuel elements are held within each fuel tube with their longitudinal axes parallel, the ribs serving to space them apart along their lengths. Coolant liquid is circulated through the fuel tubes between the spaced fuel elements. Suitable control rod and monitoring means are provided for controlling the reactor.

Treshow, M.

1958-08-19T23:59:59.000Z

387

Decontamination of large components-test case  

SciTech Connect

The rising per-cubic-foot burial costs, together with the trend toward standardized above-ground burial sites, provides the basis for seeking an alternative to direct burial of large components. Large contaminated components such as steam generators can be safely dismantled and decontaminated for free release, metals recycle, and volume reduction. This grand-scale disposal technology will prove to be an economical and ecological alternative to direct burial or interim storage. Yankee Atomic Electric Company (YAEC) in Bolton, operators and decommissioners of the Yankee Nuclear Power Station in Rowe, Massachusetts, has teamed with Frank W Hake Associates in Memphis, TN, to decontaminate a large component as a test case. The large component is YAEC`s reactor pressure vessel head (RPVH). The 79 100 lb RPVH is surface contaminated with 0.7 Ci (1500 mR/h contact) resulting from 32 yr of operating in a 2000 psi, 530{degrees}F pressurized water reactor environment.

Mancini, A. [Yankee Atomic Electric Co., Bolton, MA (United States); Bosco, B. [Frank W. Hake Associates, Memphis, TN (United States)

1996-12-31T23:59:59.000Z

388

Neutronic reactor  

DOE Patents (OSTI)

A graphite-moderated, water-cooled nuclear reactor including a plurality of rectangular graphite blocks stacked in abutting relationship in layers, alternate layers having axes which are normal to one another, alternate rows of blocks in alternate layers being provided with a channel extending through the blocks, said channeled blocks being provided with concave sides and having smaller vertical dimensions than adjacent blocks in the same layer, there being nuclear fuel in the channels.

Lewis, Warren R. (Richland, WA)

1978-05-30T23:59:59.000Z

389

NUCLEAR REACTORS  

DOE Patents (OSTI)

An active portion assembly for a fast neutron reactor is described wherein physical distortions resulting in adverse changes in the volume-to-mass ratio are minimized. A radially expandable locking device is disposed within a cylindrical tube within each fuel subassembly within the active portion assembly, and clamping devices expandable toward the center of the active portion assembly are disposed around the periphery thereof. (AEC)

Koch, L.J.; Rice, R.E. Jr.; Denst, A.A.; Rogers, A.J.; Novick, M.

1961-12-01T23:59:59.000Z

390

Neutron Radiography and Fission Mapping Measurements of Nuclear Materials with Varying Composition and Shielding  

Science Conference Proceedings (OSTI)

Neutron radiography and fission mapping measurements were performed on four measurement objects with varying composition and shielding arrangements at the Idaho National Laboratory's Zero Power Physics Reactor (ZPPR) facility. The measurement objects were assembled with ZPPR reactor plate materials comprising plutonium, natural uranium, or highly enriched uranium and were presented as unknowns for characterization. As a part of the characterization, neutron radiography was performed using a deuterium-tritium (D-T) neutron generator as a source of time and directionally tagged 14 MeV neutrons. The neutrons were detected by plastic scintillators placed on the opposite side of the object, using the time-correlation-based data acquisition of the Nuclear Materials Identification System developed at Oak Ridge National Laboratory. Each object was measured at several rotations with respect to the neutron source to obtain a tomographic reconstruction of the object and a limited identification of materials via measurement of the neutron attenuation. Large area liquid scintillators with pulse shape discrimination were used to detect the induced fission neutrons. A fission site map reconstruction was produced by time correlating the induced fission neutrons with each tagged neutron from the D-T neutron generator. This paper describes the experimental configuration, the ZPPR measurement objects used, and the neutron imaging and fission mapping results.

Mullens, James Allen [ORNL; McConchie, Seth M [ORNL; Hausladen, Paul [ORNL; Mihalczo, John T [ORNL; Grogan, Brandon R [ORNL; Sword, Eric D [ORNL

2011-01-01T23:59:59.000Z

391

FUNDAMENTALS IN THE OPERATION OF NUCLEAR TEST REACTORS. VOLUME 2. MATERIALS TESTING REACTOR DESIGN AND OPERATION  

SciTech Connect

The reactor components, building, control system and circuitry, and experimental and handling facilities are described and discussed, together with operation, shutdown, tank work and supplemental facilities. Training questions and answers are included. (D.C.W.)

1963-10-01T23:59:59.000Z

392

REACTOR CONTROL  

DOE Patents (OSTI)

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

Ruano, W.J.

1957-12-10T23:59:59.000Z

393

Evaluation of Torsatrons as reactors  

SciTech Connect

Stellarators have significant operational advantages over tokamaks as ignited steady-state reactors. This scoping study, which uses an integrated cost-minimization code that incorporates costing and reactor component models self-consistently with a 1-D energy transport calculation, shows that a torsatron reactor could also be economically competitive with a tokamak reactor. The projected cost of electricity (COE) estimated using the Advanced Reactor Innovation and Evaluation Studies (ARIES) costing algorithms is 65.6 mill/kW(e)h in constant 1992 dollars for a reference 1-GW(e) Compact Torsatron reactor case. The COE is relatively insensitive (<10% variation) over a wide range of assumptions, including variations in the maximum field allowed on the coils, the coil elongation, the shape of the density profile, the beta limit, the confinement multiplier, and the presence of a large loss region for alpha particles. The largest variations in the COE occur for variations in the electrical power output demanded and the plasma-coil separation ratio.

Lyon, J.F. [Oak Ridge National Lab., TN (United States); Gulec, K. [Univ. of Tennessee, Knoxville, TN (United States); Miller, R.L. [Los Alamos National Lab., NM (United States); El-Guebaly, L. [Univ. of Wisconsin, Madison, WI (United States)

1994-03-01T23:59:59.000Z

394

Shielding of a hadron in a finite e-beam  

SciTech Connect

The thorough study of coherent electron cooling, the modern cooling technique capable to deal with accelerators operating in the range of few TeVs, rises many interesting questions. One of them is a shielding dynamics of a hadron in an electron beam. Now this effect is computed analytically in the infinite beam approximation. Many effects are drastically different in finite and infinite plasmas. Here we propose a method to compute the dynamical shielding effect in a finite cylindrical plasma - the realistic model of an electron beam in accelerators.

Elizarov A.; Litvinenko, V.; Wang, G.

2012-05-20T23:59:59.000Z

395

Light Water Reactors Technology Development - Nuclear Reactors  

NLE Websites -- All DOE Office Websites (Extended Search)

Light Water Reactors Light Water Reactors About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy Argonne's Nuclear Science and Technology Legacy

396

Lunar Nuclear Power Plant With Solid Core Reactor, Heatpipes and Thermoelectric Conversion  

Science Conference Proceedings (OSTI)

This is a lunar nuclear power plant with the advantages of minimum mass, with no moving parts, no pumped liquid coolant, a solid metal rugged core, with no single point of failure. The electrical output is 100 kilowatts with a 500 kilowatt thermal reactor. The thermoelectric converters surround the potassium heatpipes from the core and water heatpipes surround the converter and connect to the radiator. The solid core reactor is made from HT9 alloy. The fuel is uranium oxide with 90% enrichment. The thermoelectric converter is bonded to the outside of the 1.10 inch ID heat pipe and is 30 inches long. The thermoelectric couple is Si/SiGe-Si/SiC Quantum Well with over 20% efficiency with an 890 K hot side and a 490 K cold side and produces 625 Watts. 176 converters produce 110 kWe. With less than 10% loss in controls this yields 100 kWe for use. The cylindrical thermoelectric converter is designed and fabricated by HIPing to keep brittle materials in compression and to ensure conductivity. The solid core is fabricated by machining the heatpipe tubes with 6 grooves that are diffusion bonded together by HIPing to form the fuel tubes. The maximum temperature of the heat pipes is 940 K and the return flow temperature is 890 K. The reactor core is hexagonal shaped, 61 cm. wide and 76.2 cm high with 12 rotating control drums surrounding it. There is shielding to protect components and human habitation. The radiator is daisy shaped at 45 degrees with each petal 5.5 meters long. The design life is ten years.

Sayre, Edwin D. [Engineering Consultant, 218 Brooke Acres Drive, Los Gatos, CA 95032 (United States); Ring, Peter J. [Advanced Methods and Materials, 1190 Mountain View-Alviso Rd. Suite P, Sunnyvale, CA 94089 (United States); Brown, Neil [Engineering Consultant, 5134 Cordoy Lane, San Jose, CA 95124 (United States); Elsner, Norbert B.; Bass, John C. [Hi-Z Technology, Inc., 7606 Miramar Rd. Suite 7400, San Diego, CA 92126 (United States)

2008-01-21T23:59:59.000Z

397

Monte Carlo Calculations for the Shielding Design of Beam Injection and Extraction Areas at the 3-GeV Synchrotron in J-PARC  

Science Conference Proceedings (OSTI)

Shielding / Special Issue on the 11th International Conference on Radiation Shielding and the 15th Topical Meeting of the Radiation Protection and Shielding Division (Part 2) / Accelerators

Yoshihiro Nakane; Teruo Abe; Hiroshi Nakashima

398

Materials Reliability Program: Reevaluation of Stainless Steel Components in NUREG/CR-6674 (MRP-172)  

Science Conference Proceedings (OSTI)

NUREG/CR-6674, published June 2000, described a probabilistic fracture mechanics evaluation of light water reactor components subject to environmental fatigue effects.

2005-11-15T23:59:59.000Z

399

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

Science Conference Proceedings (OSTI)

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

John Darrell Bess

2008-06-01T23:59:59.000Z

400

Stress Corrosion Cracking of the Drip Shield, the Waste Package Outer Barrier, and the Stainless Steel Structural Material  

Science Conference Proceedings (OSTI)

Stress corrosion cracking is one of the most common corrosion-related causes for premature breach of metal structural components. Stress corrosion cracking is the initiation and propagation of cracks in structural components due to three factors that must be present simultaneously: metallurgical susceptibility, critical environment, and static (or sustained) tensile stresses. This report was prepared according to ''Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]). The purpose of this report is to provide an evaluation of the potential for stress corrosion cracking of the engineered barrier system components (i.e., the drip shield, waste package outer barrier, and waste package stainless steel inner structural cylinder) under exposure conditions consistent with the repository during the regulatory period of 10,000 years after permanent closure. For the drip shield and waste package outer barrier, the critical environment is conservatively taken as any aqueous environment contacting the metal surfaces. Appendix B of this report describes the development of the SCC-relevant seismic crack density model (SCDM). The consequence of a stress corrosion cracking breach of the drip shield, the waste package outer barrier, or the stainless steel inner structural cylinder material is the initiation and propagation of tight, sometimes branching, cracks that might be induced by the combination of an aggressive environment and various tensile stresses that can develop in the drip shields or the waste packages. The Stainless Steel Type 316 inner structural cylinder of the waste package is excluded from the stress corrosion cracking evaluation because the Total System Performance Assessment for License Application (TSPA-LA) does not take credit for the inner cylinder. This document provides a detailed description of the process-level models that can be applied to assess the performance of Alloy 22 (used for the waste package outer barrier) and Titanium Grade 7 (used for the drip shield) that are subjected to the effects of stress corrosion cracking. The use of laser peening or other residual stress mitigation techniques is considered as a means of mitigating stress corrosion cracking in the waste package final closure lid weld.

G. Gordon

2004-10-13T23:59:59.000Z

Note: This page contains sample records for the topic "reactor components shield" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


401

Shielding and Build-up Considerations for Radiation Detection  

SciTech Connect

Abstract -- Shielding for gamma radiation has traditionally focused on the reduction of dose effects. For these applications, reducing the energy of the radiation is important along with reducing the actual number of photons, and therefore large masses of high Z material are typically used. However, for measurements requiring low backgrounds or for detecting low activity signals, such as in homeland security applications, the primary use of shielding is to decrease the total number of background photons (perhaps in a region of interest), and therefore the processes of buildup and down scattering become important. In these applications, where the important measure is count rate instead of dose and low background are important, improved reduction in counts from background radiation may be achieved with specially designed configurations of thin layers of different materials instead of a single thick layer. This paper briefly describes recent modeling and experimental investigations in layered-shielding methodology and provides results with comparison to single shielding material such as Pb. Application of these techniques to some real world problems, such as detector systems for homeland security, is discussed.

Ely, James H.; Kernan, Warnick J.; Kouzes, Richard T.; Siciliano, Edward R.

2008-10-19T23:59:59.000Z

402

RZ calculations for self shielded multigroup cross sections  

Science Conference Proceedings (OSTI)

A collision probability method has been implemented for RZ geometries. The method accounts for white albedo, specular and translation boundary condition on the top and bottom surfaces of the geometry and for a white albedo condition on the outer radial surface. We have applied the RZ CP method to the calculation of multigroup self shielded cross sections for Gadolinia absorbers in BWRs. (authors)

Li, M.; Sanchez, R.; Zmijarevic, I.; Stankovski, Z. [Commissariat a l'Energie Atomique CEA, Direction de l'Energie Nucleaire, DEN/DM2S/SERMA/LENR, 91191 Gif-sur-Yvette Cedex (France)

2006-07-01T23:59:59.000Z

403

Multi electron species and shielding potentials in plasmas  

SciTech Connect

The phenomenon of Debye shielding is investigated in electron ion plasmas using the approach of two temperature electrons. We get different profiles of potential for different parameters and observe that the potentials fall very slowly than the standard Coulomb and Debye potentials. The importance of work is pointed out in the introduction.

Khan, Arroj A.; Murtaza, G. [Salam Chair in Physics, Government College University, Lahore 54000 (Pakistan); Rasheed, A.; Jamil, M. [Department of Physics, Government College University, Faisalabad 38000 (Pakistan)

2012-11-15T23:59:59.000Z

404

Electromechanical coupling model for cutterhead driving system of shield machines  

Science Conference Proceedings (OSTI)

The synchronization of the cutterhead driving system of shield machines is affected by not only the nonlinearity of mechanical transmission mechanism and the characteristics of driving motors, but also their interaction. In this paper, dynamics of the ... Keywords: cutterhead, electromechanical coupling, synchronization

Jianzhong Sun; Ran Liu; Yaqin Luo; Wei Sun

2010-11-01T23:59:59.000Z

405

Monte Carlo Simulations for the Shielding of the Collector Ring and the Recycled Experimental Storage Ring at FAIR  

Science Conference Proceedings (OSTI)

Shielding / Special Issue on the 11th International Conference on Radiation Shielding and the 15th Topical Meeting of the Radiation Protection and Shielding Division (Part 2) / Radiation Protection

Alexandra Knapp; Torsten Radon; Karsten Vogt; Georg Fehrenbacher

406

First-wall, blanket, and shield engineering test program for magnetically confined fusion power reactors  

Science Conference Proceedings (OSTI)

The key engineering areas identified for early study relate to FW/B/S system thermal-hydraulics, thermomechnics, nucleonics, electromagnetics, assembly, maintenance, and repair. Programmatic guidance derived frm planning exercises involving over thirty organizations (laboratories, industries, and universities) has indicated (1) that meaningful near term engineering testing should be feasible within the bounds of a modest funding base, (2) that there are existing facilities and expertise which can be profitably utilized in this testing, and (3) that near term efforts should focus on the measurement of engineering data and the verification/calibration of predictive methods for anticipated normal operational and transient FW/B/S conditions. The remainder of this paper discusses in more detail the planning strategies, proposed approach to near term testing, and longer range needs for integrated FW/B/S test facilities.

Maroni, V.A.

1980-01-01T23:59:59.000Z

407

ELECTRONUCLEAR REACTOR  

DOE Patents (OSTI)

An electronuclear reactor is described in which a very high-energy particle accelerator is employed with appropriate target structure to produce an artificially produced material in commercial quantities by nuclear transformations. The principal novelty resides in the combination of an accelerator with a target for converting the accelerator beam to copious quantities of low-energy neutrons for absorption in a lattice of fertile material and moderator. The fertile material of the lattice is converted by neutron absorption reactions to an artificially produced material, e.g., plutonium, where depleted uranium is utilized as the fertile material.

Lawrence, E.O.; McMillan, E.M.; Alvarez, L.W.

1960-04-19T23:59:59.000Z

408

Photocatalytic reactor  

DOE Patents (OSTI)

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)

1999-01-01T23:59:59.000Z

409

Concrete Shield Performance of the VSC-17 Spent Nuclear Fuel Cask  

Science Conference Proceedings (OSTI)

The VSC-17 Spent Nuclear Fuel Storage Cask was surveyed for degradation of the concrete shield by radiation measurement, temperature measurement, and ultrasonic testing. No general loss of shielding function was identified.

Koji Shirai

2006-04-01T23:59:59.000Z

410

12/16/03 1/21/04 Cable Management and Shielding  

Science Conference Proceedings (OSTI)

... 12/16/03 1/21/04 Cable Management and Shielding. ... The aperture plug installed in the shield wall and the 1 inch insert on the outer optical bench. ...

411

CIVILIAN POWER REACTOR PROGRAM. PART III. BOOK I. STATUS REPORT ON FAST REACTORS AS OF 1959  

SciTech Connect

A description of fast reactor types is given and the objectives of the AEC program on fast reactors are outlined. General research and development programs completed and in process on physics, fuel and materials, heat transfer, fluid flow, coolant chemistry, reactor safety, and components and systems are discussed. Reactors completed, under design and construction, and under study are described. (W.D.M.) A general description of ship and power plant, tabulation of principle characteristics, economics, potential advances, and necessary research and development are discussed in detail. (W.D.M.)

1960-01-01T23:59:59.000Z

412

Materials Degradation in Light Water Reactors: Life After 60 | Department  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Materials Degradation in Light Water Reactors: Life After 60 Materials Degradation in Light Water Reactors: Life After 60 Materials Degradation in Light Water Reactors: Life After 60 Nuclear reactors present a very harsh environment for components service. Components within a reactor core must tolerate high temperature water, stress, vibration, and an intense neutron field. Degradation of materials in this environment can lead to reduced performance, and in some cases, sudden failure. A recent EPRI-led study interviewed 47 US nuclear utility executives to gauge perspectives on long-term operation of nuclear reactors. Nearly 90% indicated that extensions of reactor lifetimes to beyond 60 years were likely. When polled on the most challenging issues facing further life extension, two-thirds cited plant reliability as the

413

Materials Degradation in Light Water Reactors: Life After 60 | Department  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Materials Degradation in Light Water Reactors: Life After 60 Materials Degradation in Light Water Reactors: Life After 60 Materials Degradation in Light Water Reactors: Life After 60 Nuclear reactors present a very harsh environment for components service. Components within a reactor core must tolerate high temperature water, stress, vibration, and an intense neutron field. Degradation of materials in this environment can lead to reduced performance, and in some cases, sudden failure. A recent EPRI-led study interviewed 47 US nuclear utility executives to gauge perspectives on long-term operation of nuclear reactors. Nearly 90% indicated that extensions of reactor lifetimes to beyond 60 years were likely. When polled on the most challenging issues facing further life extension, two-thirds cited plant reliability as the

414

IMRT treatment of anal cancer with a scrotal shield  

SciTech Connect

The risk of sterility in males undergoing radiotherapy in the pelvic region indicates the use of a shielding device, which offers protection to the testes for patients wishing to maintain fertility. The use of such devices in the realm of intensity-modulated radiotherapy (IMRT) in the pelvic region can pose many obstacles during simulation, treatment planning, and delivery of radiotherapy. This work focuses on the development and execution of an IMRT plan for the treatment of anal cancer using a scrotal shielding device on a clinical patient. An IMRT plan was developed using Eclipse treatment planning system (Varian Medical Systems, Palo Alto, CA), using a wide array of gantry angles as well as fixed jaw and fluence editing techniques. When possible, the entire target volume was encompassed by the treatment field. When the beam was incident on the scrotal shield, the jaw was fixed to avoid the device and the collimator rotation optimized to irradiate as much of the target as possible. This technique maximizes genital sparing and allows minimal irradiation of the gonads. When this fixed-jaw technique was found to compromise adequate coverage of the target, manual fluence editing techniques were used to avoid the shielding device. Special procedures for simulation, imaging, and treatment verification were also developed. In vivo dosimetry was used to verify and ensure acceptable dose to the gonads. The combination of these techniques resulted in a highly conformal plan that spares organs and risk and avoids the genitals as well as entrance of primary radiation onto the shielding device.

Hood, Rodney C., E-mail: Rodney.Hood@duke.edu [Department of Radiation Oncology, Duke University Medical Center, Durham, NC (United States); Wu, Q. Jackie; McMahon, Ryan; Czito, Brian; Willett, Christopher [Department of Radiation Oncology, Duke University Medical Center, Durham, NC (United States)

2012-01-01T23:59:59.000Z

415

Flood and Shield Basalts from Ethiopia: Magmas from the African Superswell  

E-Print Network (OSTI)

-known, but less well-preserved, flood basalt provinces such as the Deccan and Karoo. The shield volcanoes have

Demouchy, Sylvie

416

BSA 02-23: Delivering Shielded Electron Beams to a Target in a ...  

The apparatus includes a plasma generator aligned with a beam generator for producing plasma to shield an energized beam.

417

Reliability Assessment of SMART Reactor Protection System  

Science Conference Proceedings (OSTI)

Component failure rates and integrated system reliability of the SMART reactor protection system were analyzed. The analysis tool of the study was the RELEX 7 computer program developed by Relex Software Corporation. The RELEX software is a PC based computer program which includes the part stress analysis models and the RBD analysis model to calculate component and system reliability. The component failure rate data for the study was selected from the MIL-HDBK-217F. (authors)

Won Young, Yun; Choong Heui, Jeong [Korea Institute of Nuclear Safety, P. O. Box 114, Yusong Post Office, Daejeon (Korea, Republic of); Seong Hun, Kim; Sang Yong, Lee [Sam Chang Enterprise Co. LTD, 974-1, Goyen-ri, Woongchon-myon, Ulju-gun, Ulsan (Korea, Republic of)

2006-07-01T23:59:59.000Z

418

HOMOGENEOUS REACTOR PROJECT QUARTERLY PROGRESS REPORT FOR PERIOD ENDING JULY 31, 1955  

SciTech Connect

8 5 6 2 0 5 8 5 5 ? 5 : tion of the HRT is summarized. The design of some reactor components is described, including fuel system line coolers, the thermal shield, instrument cubicles, the reactor cell air monitor, and sampling facilities. The important design specifications for the HRT fissionproduct adsorption system are outlined. The effects of fluid circulation on HRT stability, the results of the rupture of the fuel inlet nozzle on the pressure vessel, core tank temperature, and heat generation and temperature distribution in the control room area shield were examined. Operation of the HRT mockup loop is summarized. High generalized corrosion rates and U precipitation were observed when 50 ppm oxygen was used. Results from tests of HRT pumps, heat exchanger, sampler mockup, and reflex condenser and recombiner are presented. Design of the instrumentation in the control room area and fabrication of differential pressure and liquid level transmitters are reported. Thorium Breeder Reactor. Stability criteria, the effect of slurry settling on reactivity, and neutron losses due to coating the core tank with platinum bave been analyzed for the TBR. An economic study of partial fuel costs for one-region Pu-U reactors was completed. The design concept of a 300- electrical-Mw thorium breeder power station utilizing low-pressure gas recombination was completed, and construction and operating costs were estimated. Design problems which must be solved before the design can be considered practical are discussed. Medification of TBR design from low- to high-pressure recombination was started. Corrosion. Operation of dynamic-corroslon test loops with uranyl sulfate solutions and thorium oxide slurries proceeded in a routine manner, and results are presented tabularly. Sedimentation characteristics of thorium dioxide slurries were determined at 100 to 300 deg C. Development, fabrication, and operation of in-side loops for radiation corrosion studies are described. Data from in-side loop and bomb experiments to study the corrosion of Zircaloy-2 and stainless steel under irradiation are reported. Results are presented from laboratory corrosion studies of stressed and unstressed specimens of stainless steel, titanium alloy and Zlrcaloy-2, and unstressed specimens of aluminum oxide in simulated HRT solutions. The effects of sulfate additions, calcination temperature, and particle size on the corrosive effects of thorium oxide slurries were studied. Engneering Development. The status of the heat exchanger development program is summarized. Thorium oxide slurry circulation studies and the development of blanket-system components such as dump tanks and pressurizer are continuing. The power removal capabilities of boiling homogeneous reactors are being studied. The results of stress-corrosion tests on austenitic stainless steels and dynamiccorrosion tests of stainless steel welds are reported. The mechanical properties of irradiated Zircaloy-2, stainless steel, and carbon steel specimens are being studied. Chemical Engineering Development. Studies to determine the effect of irradiation on the properties of thorium oxide slurries and to demonstrate slurry radiation stability were continued. Silver was shown to be a good internal catalyst for the recombination of radiolytic gases in thorium oxide slurries. Progress was made in the characterization of thorium oxide preducts and slurries by x-ray, nitrogen adsorption surface area, thermogravimetric, and gaseous adsorption techniques, sedimentation particle size analyses, and bulk density measurements. The HRT fuel processing plant is described. The precipitation of needymium sulfate from

McDuffle, H.F.; Kelly, D.C. comps.

1955-08-01T23:59:59.000Z

419

Progressive Application Decommissioning Models for U.S. Power and Research Reactors  

SciTech Connect

This paper presents progressive engineering techniques and experiences in decommissioning projects performed by Bums and Roe Enterprises within the last fifteen years. Specifically, engineering decommissioning technical methods and lessons learned are discussed related to the Trojan Large Component Removal Project, San Onofre Nuclear Generating Station (SONGS) Decommissioning Project and the Brookhaven Graphite Research Reactor (BGRR) Decommissioning Project Study. The 25 years since the 1979 TMI accident and the events following 9/11 have driven the nuclear industry away from excessive, closed/elitist conservative methods towards more pragmatic results-oriented and open processes. This includes the essential recognition that codes, standards and regulatory procedures must be efficient, effective and fit for purpose. Financial and open-interactive stakeholder pressures also force adherence to aggressive risk reduction posture in the area of a safety, security and operations. The engineering methods and techniques applied to each project presented unique technical solutions. The decommissioning design for each project had to adopt existing design rules applicable to construction of new nuclear power plants and systems. It was found that the existing ASME, NRC, and DOE codes and regulations for deconstruction were, at best, limited or extremely conservative in their applicability to decommissioning. This paper also suggests some practical modification to design code rules in application for decommissioning and deconstruction. The representative decommissioning projects, Trojan, SONGS and Brookhaven, are discussed separately and the uniqueness of each project, in terms of engineering processes and individual deconstruction steps, is discussed. Trojan Decommissioning. The project included removal of entire NSSS system. The engineering complexity was mainly related to the 1200 MW Reactor. The approach, process of removal, engineering method related to protect the worker against excessive radiation exposure, transportation, and satisfying applicable rules and regulations, were the major problems to overcome. The project's successful completed earned a patent award. SONGS Decommissioning. The reactor's spherical containment and weakened integrity was the scope of this decommissioning effort. The aspects of structure stability and method of deconstruction is the major part of the presentation. The economical process of deconstruction, aspects of structural stability, worker safety, and the protection of the surrounding environment from contamination is highlighted in this section. BGRR Decommissioning Study. BREI was commissioned by Brookhaven National Laboratory (BNL) to evaluate and analyze the stability, and progressive decommissioning, and removal of BGRR components. This analysis took the form of several detailed decommissioning studies that range from disassembly and removal of the unit's graphite pile to the complete environmental restoration of the reactor site. While most of the facility's decommissioning effort is conventional, the graphite pile and its biological shield present the greatest challenge. The studies develop a unique method of removing high-activity waste trapped in the graphite joints. (authors)

Studnicka, Z.; Lacy, N.H.; Nicholas, R.G.; Campagna, M.; Morgan, R.D. [Bums and Roe Enterprises, Inc., 800 Kinderkamack Road, Oradell, NJ 07649 (United States); Sawruk, W. [ABS Consulting, Inc., 5 Birdsong Court, Shillington, PA 19607 (United States)

2006-07-01T23:59:59.000Z

420

Reduced Loss in Precipitation Measurements Using a New Wind Shield for Raingages  

Science Conference Proceedings (OSTI)

A problem of great concern in precipitation measurements is the wind loss. This paper presents a new wind shield that reduces this loss. Tests of the new shield were made in a wind tunnel and in the field. The wind shield consisted of a flange ...

Anders Lindroth

1991-06-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor components shield" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


421

NHI Component Technical Readiness Evaluation System  

DOE Green Energy (OSTI)

A decision process for evaluating the technical readiness or maturity of components (i.e., heat exchangers, chemical reactors, valves, etc.) for use by the U.S. DOE Nuclear Hydrogen Initiative is described. This system is used by the DOE NHI to assess individual components in relation to their readiness for pilot-scale and larger-scale deployment and to drive the research and development work needed to attain technical maturity. A description of the evaluation system is provided, and examples are given to illustrate how it is used to assist in component R&D decisions.

Steven R. Sherman; Dane F. Wilson; Steven J. Pawel

2007-09-01T23:59:59.000Z

422

CONTROL MEANS FOR REACTOR  

DOE Patents (OSTI)

An apparatus for controlling a nuclear reactor includes a tank just below the reactor, tubes extending from the tank into the reactor, and a thermally expansible liquid neutron absorbent material in the tank. The liquid in the tank is exposed to a beam of neutrons from the reactor which heats the liquid causing it to expand into the reactor when the neutron flux in the reactor rises above a predetermincd danger point. Boron triamine may be used for this purpose.

Manley, J.H.

1961-06-27T23:59:59.000Z

423

Economic analysis of nuclear reactors  

SciTech Connect

The report presents several methods for estimating the power costs of nuclear reactors. When based on a consistent set of economic assumptions, total power costs may be useful in comparing reactor alternatives. The principal items contributing to the total power costs of a nuclear power plant are: (1) capital costs, (2) fuel cycle costs, (3) operation and maintenance costs, and (4) income taxes and fixed charges. There is a large variation in capital costs and fuel expenses among different reactor types. For example, the standard once-through LWR has relatively low capital costs; however, the fuel costs may be very high if U/sub 3/O/sub 8/ is expensive. In contrast, the FBR has relatively high capital costs but low fuel expenses. Thus, the distribution of expenses varies significantly between these two reactors. In order to compare power costs, expenses and revenues associated with each reactor may be spread over the lifetime of the plant. A single annual cost, often called a levelized cost, may be obtained by the methods described. Levelized power costs may then be used as a basis for economic comparisons. The paper discusses each of the power cost components. An exact expression for total levelized power costs is derived. Approximate techniques of estimating power costs will be presented.

Owen, P.S.; Parker, M.B.; Omberg, R.P.

1979-05-01T23:59:59.000Z

424

ALPR PRELIMINARY DESIGN STUDY (ARGONNE LOW POWER REACTOR). PHASE I  

SciTech Connect

A preliminary design study, Phase I of the ALPR . project, has been made in accordance with the Army Reactors Branch specifications for a nuclear ''package'' power plant with a 200 to 260 kw electric and 400 kw heating capacity..The plant is to be installed at the Idaho Reactor Testing Station as a prototype for remote arctic installations. The ''conventiornl'' power plant as well as the exterior reactor components are described, and cost estimates are given. ''Nuclear'' components of the reactor are described. (auth)

Treshow, M.; Hamer, E.; Pearlman, H.; Rossin, D.; Shaftman, D.

1956-04-20T23:59:59.000Z

425

Nuclear Reactor Accidents  

NLE Websites -- All DOE Office Websites (Extended Search)

Reactor Accidents The accidents at the Three Mile Island (TMI) and Chernobyl nuclear reactors have triggered particularly intense concern about radiation hazards. The TMI accident,...

426

Principles of Reactor Physics  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Reactor Physics M A Smith Argonne National Laboratory Nuclear Engineering Division Phone: 630-252-9747, Email: masmith@anl.gov Abstract: Nuclear reactor physics deals with...

427

Reactor safeguards against insider sabotage  

SciTech Connect

A conceptual safeguards system is structured to show how both reactor operations and physical protection resources could be integrated to prevent release of radioactive material caused by insider sabotage. Operational recovery capabilities are addressed from the viewpoint of both detection of and response to disabled components. Physical protection capabilities for preventing insider sabotage through the application of work rules are analyzed. Recommendations for further development of safeguards system structures, operational recovery, and sabotage prevention are suggested.

Bennett, H.A.

1982-03-01T23:59:59.000Z

428

NUCLEAR REACTOR COMPENENT CLADDING MATERIAL  

DOE Patents (OSTI)

Fuel elements and coolant tubes used in nuclear reactors of the heterogeneous, water-cooled type are described, wherein the coolant tubes extend through the moderator and are adapted to contain the fuel elements. The invention comprises forming the coolant tubes and the fuel element cladding material from an alloy of aluminum and nickel, or an alloy of aluminum, nickel, alloys are selected to prevent intergranular corrosion of these components by water at temperatures up to 35O deg C.

Draley, J.E.; Ruther, W.E.

1959-01-27T23:59:59.000Z

429

Conceptual design study of spheromak reactors  

SciTech Connect

Preliminary design studies are carried out for a spheromak fusion reactor. Simplified circuit theory is applied to obtain characteristic relations among various parameters of the spheromak configuration for an aspect ratio A greater than or equal to 1.6. These relations are used to calculate the parameters for the conceptual designs of three types of fusion reactor: (1) DT two-component, (2) DT ignited, and, (3) catalyzed DD ignited reactors. With a total wall loading of approx. 4 MWm/sup -2/, it is found that edge magnetic fields of only approx. 4T (DT) and approx. 9T (cat. DD) are required for ignited reactors of one-meter plasma (minor) radius with output powers in the gigawatt range. Assessment of various methods of generating reactor-grade spheromak plasmas is discussed briefly.

Katsurai, M.; Yamada, M.

1980-07-01T23:59:59.000Z

430

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A power plant is described comprising a turbine and employing round cylindrical fuel rods formed of BeO and UO/sub 2/ and stacks of hexagonal moderator blocks of BeO provided with passages that loosely receive the fuel rods so that coolant may flow through the passages over the fuels to remove heat. The coolant may be helium or steam and fiows through at least one more heat exchanger for producing vapor from a body of fluid separate from the coolant, which fluid is to drive the turbine for generating electricity. By this arrangement the turbine and directly associated parts are free of particles and radiations emanating from the reactor. (AEC)

Daniels, F.

1962-12-18T23:59:59.000Z

431

MARITIME GAS-COOLED REACTOR PROGRAM QUARTERLY PROGRESS REPORT FOR PERIOD ENDING, SEPTEMBER 30, 1960  

SciTech Connect

The feasibility was studied and a cost estimate prepared of an experimental reactor to determine the operating characteristics of beryllia- moderated, gas-cooled systems wtthin a power limit of 10 Mw(t). The heat energy produced by the experimental reactor is to be dissipated in a heat dump. No machinery for production of power was to be provided. Other requirements were that the reactor should be capable of testing core types different from the current MGCR design, and the system should permit use of gases other than helium. It was further directed that the reactor should be designated BORE for Beryllium- Oxide Reactor Experiment. Reactor development work was mainly in connection with the BORE preliminary design. It was established that the most important information which could be provided by a 10 Mw(t) reactor experiment would be on performance of fuel elements and moderator bodies. This required that the experirment duplicate the power density in the fuel and moderator that would exist in the full size reactor and made it advisable to use full length fuel elements. This resulted in an unconventionally shaped core which is roughly cylindrical with the length more than twice its mean diameter. Studies continued on performance of fuel elements, and methods were developed for calculation of thermal stress in BeO-moderated modules. These studies are equally applicable to the MGCR prototype and BORE. A thermal analysis of the MGCR pressure vessel and thermal shields was performed and means of externally cooling the vessel were studied. Some of the components of the experimental control rod drive mechanism were received. Endurance tests of a ball-nut lead screw in hot helium continued. Heat exchanger tests were resumed after an interruption due to leaks in the tube to tubesheet joints in the test unit. Plant control studies were continued with analyses of the system dynamics with a turbomachinery configuration in which the high pressure turbine provides the output power. This arrangement was found to be more sensitive than the low-pressure drive system. Turbomachinery component tests are providing stage performance data. The seal and bearing test rig was completed and tests were begun. Physics calculations were made for the BORE design. Basic physics information on Fermi age and neutron thermalization in BeO was provided during the quarter by experlments in the linear accelerator. Materials work continued with further investigations of the effect of additives on properties of UC/sub 2/--BeO diluted fuel bodies. Hot cell examination of the MTR-31-3 fuel capsule indicated no significant dimensional changes after burnups from 30,000 to 45,000 Mwd/T. An experiment to determine irradiation effects on BeO was inserted in GETR during September. Work was started on development of high density BeO bodies. Structural materials research continued with completion of a second series of self-welding tests of different metal pairs in high temperature helium. A series of galling tests and a series of creep rupture tests of weld specimens in SA302-B steel were completed. Work on site development included establishing requirements of the BORE for buildings, auxiliary facilities, and utilities. (auth)

1960-09-30T23:59:59.000Z

432

Reactor and method of operation  

DOE Patents (OSTI)

A nuclear reactor having a flattened reactor activity curve across the reactor includes fuel extending over a lesser portion of the fuel channels in the central portion of the reactor than in the remainder of the reactor.

Wheeler, John A. (Princeton, NJ)

1976-08-10T23:59:59.000Z

433

SUMMARY OF REACTOR DESIGN INFORMATION FROM THREE YEARS' OPERATION OF A SMALL PWR  

SciTech Connect

Reactor design information obtained from 3 years' operation of a small pressurized-water reactor, the SM-1 (formerly APPR-l), is presented and discussed. The SM-1 reactor, designed to produce 10 Mw(t) power, employs fully enriched uranium fuel in the form of UO/sub 2/ dispersed in stainless-steel fuel plates. The reactor is cooled by water at 1200 psia and mean temperature of 44) deg F. Core-physics measurements were performed of temperature coefficient, pressure coefficient, rod calibration, stuck rod position, and transient xenon as a function of core burn-out. Core burn-out characteristics were compared with few- group calculations, and reasonable agreement was obtained. Thermal-heat-balance data were obtained on the reactor core. The temperature pattern in the nominal and hot channels under operating conditions was calculated. These calculations indicated that certain of the fuel channels operated in the nucleate boiling regime. Examination of one of the fuel channels suspected of nucleate boiling indicated no adverse effects. The system response to load perturbations and during pump coast-down was measured utilizing plant instrumentation. This response was compared with analytical predictions using a lumped kinetic model, and reasonable agreement was found. Both neutron and gamma traverses were made through the primary shield during reactor operation. Gamma traverses were also made through the primary shield as a function of time after reactor shutdown. Conventional shielding calculational methods are found to give agreement with experiment sufficient for design purposes. An absolute ionization chamber was employed to measure N/sup 16/ activity in the reactor coolant. These measurements were compared with N/sup 16/ calculated from the (n,p) reaction on O/ sup 16/. (auth)

Gallagher, J.G.

1960-09-01T23:59:59.000Z

434

Shielding of mirror FERF plasma by arc discharges  

SciTech Connect

The feasibility of shielding a mirror-confined fusion plasma against erosion by incident neutrals with a plasma blanket generated by an array of hollow-cathode arc discharges was studied. Such a plasma blanket could also be used for linetying stabilization of a single mirror confined plasma as well as to provide a warm plasma stream for stabilization of microinstabilities. The requirements for the plasma blanket are dependent on the parameter ..gamma.., the ratio of the actual cross-field diffusion coefficient to the classical value. The power requirement compares favorably with power loss due to change exchange without shielding. More importantly, the blanket permits a relaxation of vacuum requirements to prevent erosion of the hot plasma by background neutrals.

Woo, J.T.

1976-12-08T23:59:59.000Z

435

SEU43 fuel bundle shielding analysis during spent fuel transport  

Science Conference Proceedings (OSTI)

The basic task accomplished by the shielding calculations in a nuclear safety analysis consist in radiation doses calculation, in order to prevent any risks both for personnel protection and impact on the environment during the spent fuel manipulation, transport and storage. The paper investigates the effects induced by fuel bundle geometry modifications on the CANDU SEU spent fuel shielding analysis during transport. For this study, different CANDU-SEU43 fuel bundle projects, developed in INR Pitesti, have been considered. The spent fuel characteristics will be obtained by means of ORIGEN-S code. In order to estimate the corresponding radiation doses for different measuring points the Monte Carlo MORSE-SGC code will be used. Both codes are included in ORNL's SCALE 5 programs package. A comparison between the considered SEU43 fuel bundle projects will be also provided, with CANDU standard fuel bundle taken as reference. (authors)

Margeanu, C. A.; Ilie, P.; Olteanu, G. [Inst. for Nuclear Research Pitesti, No. 1 Campului Street, Mioveni 115400, Arges County (Romania)

2006-07-01T23:59:59.000Z

436

Accordian-folded boot shield for flexible swivel connection  

DOE Patents (OSTI)

A flexible swivel boot connector for connecting a first boot shield section to a second boot shield section, both first and second boot sections having openings therethrough, the second boot section having at least two adjacent accordian folds at the end having the opening, the second boot section being positioned through the opening of the first boot section such that a first of the accordian folds is within the first boot section and a second of the accordian folds is outside of the first boot, includes first and second annular discs, the first disc being positioned within and across the first accordian fold, the second disc being positioned within and across the second accordian fold, such that the first boot section is moveably and rigidly connected between the first and second accordian folds of the second boot section.

Hoh, Joseph C. (Naperville, IL)

1986-01-01T23:59:59.000Z

437

Secret key distillation from shielded two-qubit states  

Science Conference Proceedings (OSTI)

The quantum states corresponding to a secret key are characterized using the so-called private states, where the key part consisting of a secret key is shielded by the additional systems. Based on the construction, it was shown that a secret key can be distilled from bound entangled states. In this work, I consider the shielded two-qubit states in a key-distillation scenario and derive the conditions under which a secret key can be distilled using the recurrence protocol or the two-way classical distillation, advantage distillation together with one-way postprocessing. From the security conditions, it is shown that a secret key can be distilled from bound entangled states in a much wider range. In addition, I consider the case that in which white noise is added to quantum states and show that the classical distillation protocol still works despite a certain amount of noise although the recurrence protocol does not.

Bae, Joonwoo [School of Computational Sciences, Korea Institute for Advanced Study, Seoul 130-722 (Korea, Republic of)

2010-05-15T23:59:59.000Z

438

Secret key distillation from shielded two-qubit states  

E-Print Network (OSTI)

The quantum states corresponding to a secret key are characterized using the so-called private states, where the key part consisting of a secret key is shielded by the additional systems. Based on the construction, it was shown that a secret key can be distilled from bound entangled states. In this work, I consider the shielded two-qubit states in a key-distillation scenario and derive the conditions under which a secret key can be distilled using the recurrence protocol or the two-way classical distillation, advantage distillation together with one-way postprocessing. From the security conditions, it is shown that a secret key can be distilled from bound entangled states in a much wider range. In addition, I consider the case that in which white noise is added to quantum states and show that the classical distillation protocol still works despite a certain amount of noise although the recurrence protocol does not.

Joonwoo Bae

2008-03-03T23:59:59.000Z

439

Dynamic magnetic shield for the CLAS12 central TOF detector photomultiplier tubes  

SciTech Connect

The Central Time-of-Flight detector for the Jefferson Laboratory 12-GeV upgrade is being designed with linear-focused photomultiplier tubes that require a robust magnetic shield against the CLAS12 main 5-T solenoid fringe fields of 100 mT (1 kG). Theoretical consideration of a ferromagnetic cylinder in an axial field has demonstrated that its shielding capability decreases with increasing length. This observation has been confirmed with finite element analysis using Poisson model software. Several shields composed of coaxial ferromagnetic cylinders have been studied. All difficulties caused by saturation effects were overcome with a novel dynamical shield, which utilizes a demagnetizing solenoid between the shielding cylinders. Basic dynamical shields for ordinary linear-focused 2-in. photomultiplier tubes were designed and tested both with models and experimental prototypes at different external field and demagnetizing current values. Our shield design reduces the 1 kG external axial field by a factor of 5000.

V. Baturin, V. Burkert, D.S. Carman, L. Elouadrhyri, D. Grilli, D. Kashy, E. Pasyuk, L. Quettier, B. Wieland

2012-02-01T23:59:59.000Z

440

Thermal Analysis of a SHIELD Electromigration Test Structure  

Science Conference Proceedings (OSTI)

The steady state and transient thermal behavior of an electromigration test structure was analyzed. The test structure was a Sandia SHIELD (Self-stressing HIgh fregquency rELiability Device) electromigration test device manufactured by an outside vendor. This device has a high frequency oscillator circuit, a buffer circuit to isolate and drive the metal line to the tested (DUT), the DUT to be electromigrated itself, a metal resistance thermometry monitor, and a heater elment to temperature accelerate the electromigration effect.

Benson, David A.; Bowman, Duane J.; Mitchell, Robert T.

1999-05-01T23:59:59.000Z

Note: This page contains sample records for the topic "reactor components shield" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.