Reactor vessel support system. [LMFBR

DOEpatents

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

1980-05-09

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

ZPR-6 assembly 7 high {sup 240}Pu core experiments : a fast reactor core with mixed (Pu,U)-oxide fuel and a centeral high{sup 240}Pu zone.

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

Lell, R. M.; Morman, J. A.; Schaefer, R.W.

ZPR-6 Assembly 7 (ZPR-6/7) encompasses a series of experiments performed at the ZPR-6 facility at Argonne National Laboratory in 1970 and 1971 as part of the Demonstration Reactor Benchmark Program (Reference 1). Assembly 7 simulated a large sodium-cooled LMFBR with mixed oxide fuel, depleted uranium radial and axial blankets, and a core H/D near unity. ZPR-6/7 was designed to test fast reactor physics data and methods, so configurations in the Assembly 7 program were as simple as possible in terms of geometry and composition. ZPR-6/7 had a very uniform core assembled from small plates of depleted uranium, sodium, iron oxide,more » U{sub 3}O{sub 8} and Pu-U-Mo alloy loaded into stainless steel drawers. The steel drawers were placed in square stainless steel tubes in the two halves of a split table machine. ZPR-6/7 had a simple, symmetric core unit cell whose neutronic characteristics were dominated by plutonium and {sup 238}U. The core was surrounded by thick radial and axial regions of depleted uranium to simulate radial and axial blankets and to isolate the core from the surrounding room. The ZPR-6/7 program encompassed 139 separate core loadings which include the initial approach to critical and all subsequent core loading changes required to perform specific experiments and measurements. In this context a loading refers to a particular configuration of fueled drawers, radial blanket drawers and experimental equipment (if present) in the matrix of steel tubes. Two principal core configurations were established. The uniform core (Loadings 1-84) had a relatively uniform core composition. The high {sup 240}Pu core (Loadings 85-139) was a variant on the uniform core. The plutonium in the Pu-U-Mo fuel plates in the uniform core contains 11% {sup 240}Pu. In the high {sup 240}Pu core, all Pu-U-Mo plates in the inner core region (central 61 matrix locations per half of the split table machine) were replaced by Pu-U-Mo plates containing 27% {sup 240}Pu in the plutonium component to construct a central core zone with a composition closer to that in an LMFBR core with high burnup. The high {sup 240}Pu configuration was constructed for two reasons. First, the composition of the high {sup 240}Pu zone more closely matched the composition of LMFBR cores anticipated in design work in 1970. Second, comparison of measurements in the ZPR-6/7 uniform core with corresponding measurements in the high {sup 240}Pu zone provided an assessment of some of the effects of long-term {sup 240}Pu buildup in LMFBR cores. The uniform core version of ZPR-6/7 is evaluated in ZPR-LMFR-EXP-001. This document only addresses measurements in the high {sup 240}Pu core version of ZPR-6/7. Many types of measurements were performed as part of the ZPR-6/7 program. Measurements of criticality, sodium void worth, control rod worth and reaction rate distributions in the high {sup 240}Pu core configuration are evaluated here. For each category of measurements, the uncertainties are evaluated, and benchmark model data are provided.« less

Multirecycling of Plutonium from LMFBR Blanket in Standard PWRs Loaded with MOX Fuel

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

Sonat Sen; Gilles Youinou

2013-02-01

It is now well-known that, from a physics standpoint, Pu, or even TRU (i.e. Pu+M.A.), originating from LEU fuel irradiated in PWRs can be multirecycled also in PWRs using MOX fuel. However, the degradation of the isotopic composition during irradiation necessitates using enriched U in conjunction with the MOX fuel either homogeneously or heterogeneously to maintain the Pu (or TRU) content at a level allowing safe operation of the reactor, i.e. below about 10%. The study is related to another possible utilization of the excess Pu produced in the blanket of a LMFBR, namely in a PWR(MOX). In this casemore » the more Pu is bred in the LMFBR, the more PWR(MOX) it can sustain. The important difference between the Pu coming from the blanket of a LMFBR and that coming from a PWR(LEU) is its isotopic composition. The first one contains about 95% of fissile isotopes whereas the second one contains only about 65% of fissile isotopes. As it will be shown later, this difference allows the PWR fed by Pu from the LMFBR blanket to operate with natural U instead of enriched U when it is fed by Pu from PWR(LEU)« less

Transient Effects in Turbulence Modelling.

DTIC Science & Technology

1979-12-01

plenum region of a liquid-metal- cooled fast breeder reactor (LMFBR). The efficient heat transfer characteristics of liquid metal coolant, combined...Transients in Generalized Liquid-Metal Fast Breeder Reactor Outlet Plenums," Nuclear Technology, Vol. 44, July 1979, p. 210. 135 15. Lorenz, J. J., "MIX... Sodium Coolant in the Outlet Plenum of a Fast Nuclear Reactor ," Int. J. Heat Mass Transfer, Vol. 21, 1978, pp. 1565-1579. 19. Chen, Y. B., Golay, M. W

In-situ method for treating residual sodium

DOEpatents

Sherman, Steven R.; Henslee, S. Paul

2005-07-19

A unique process for deactivating residual sodium in Liquid Metal Fast Breeder Reactor (LMFBR) systems which uses humidified (but not saturated) carbon dioxide at ambient temperature and pressure to convert residual sodium into solid sodium bicarbonate.

In-Situ Method for Treating Residual Sodium

DOEpatents

Sherman, Steven R.; Henslee, S. Paul

2005-07-19

A unique process for deactivating residual sodium in Liquid Metal Fast Breeder Reactor (LMFBR) systems which uses humidified (but not saturated) carbon dioxide at ambient temperature and pressure to convert residual sodium into solid sodium bicarbonate.

Nuclear Engineering Computer Modules, Thermal-Hydraulics, TH-2: Liquid Metal Fast Breeder Reactors.

ERIC Educational Resources Information Center

Reihman, Thomas C.

This learning module is concerned with the temperature field, the heat transfer rates, and the coolant pressure drop in typical liquid metal fast breeder reactor (LMFBR) fuel assemblies. As in all of the modules of this series, emphasis is placed on developing the theory and demonstrating the use with a simplified model. The heart of the module is…

Automatic safety rod for reactors. [LMFBR

DOEpatents

Germer, J.H.

1982-03-23

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

Carter's breeder policy has failed, claims Westinghouse manager

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

Not Available

1979-07-01

Nuclear nations developing liquid metal fast breeder reactor (LMFBR) technology have not been dissuaded by President Carter's efforts to stop the breeder program as a way to control the proliferation of nuclear weapons. There is no evidence that Carter's policy of moral persuasion has had any impact on their efforts. A review of the eight leading countries cites their extensive progress in the areas of breeder technology and fuel reprocessing, while the US has made only slight gains. The Fast Flux Test Facility at Hanford is near completion, but the Clinch River project has been slowed to a minimum.

Lessons Learned about Liquid Metal Reactors from FFTF Experience

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

Wootan, David W.; Casella, Andrew M.; Omberg, Ronald P.

2016-09-20

The Fast Flux Test Facility (FFTF) is the most recent liquid-metal reactor (LMR) to operate in the United States, from 1982 to 1992. FFTF is located on the DOE Hanford Site near Richland, Washington. The 400-MWt sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission test reactor was designed specifically to irradiate Liquid Metal Fast Breeder Reactor (LMFBR) fuel and components in prototypical temperature and flux conditions. FFTF played a key role in LMFBR development and testing activities. The reactor provided extensive capability for in-core irradiation testing, including eight core positions that could be used with independent instrumentation for the test specimens.more » In addition to irradiation testing capabilities, FFTF provided long-term testing and evaluation of plant components and systems for LMFBRs. The FFTF was highly successful and demonstrated outstanding performance during its nearly 10 years of operation. The technology employed in designing and constructing this reactor, as well as information obtained from tests conducted during its operation, can significantly influence the development of new advanced reactor designs in the areas of plant system and component design, component fabrication, fuel design and performance, prototype testing, site construction, and reactor operations. The FFTF complex included the reactor, as well as equipment and structures for heat removal, containment, core component handling and examination, instrumentation and control, and for supplying utilities and other essential services. The FFTF Plant was designed using a “system” concept. All drawings, specifications and other engineering documentation were organized by these systems. Efforts have been made to preserve important lessons learned during the nearly 10 years of reactor operation. A brief summary of Lessons Learned in the following areas will be discussed: Acceptance and Startup Testing of FFTF FFTF Cycle Reports« less

Atom probe tomography analysis of high dose MA957 at selected irradiation temperatures

NASA Astrophysics Data System (ADS)

Bailey, Nathan A.; Stergar, Erich; Toloczko, Mychailo; Hosemann, Peter

2015-04-01

Oxide dispersion strengthened (ODS) alloys are meritable structural materials for nuclear reactor systems due to the exemplary resistance to radiation damage and high temperature creep. Summarized in this work are atom probe tomography (APT) investigations on a heat of MA957 that underwent irradiation in the form of in-reactor creep specimens in the Fast Flux Test Facility-Materials Open Test Assembly (FFTF-MOTA) for the Liquid Metal Fast Breeder Reactor (LMFBR) program. The oxide precipitates appear stable under irradiation at elevated temperature over extended periods of time. Nominally, the precipitate chemistry is unchanged by the accumulated dose; although, evidence suggests that ballistic dissolution and reformation processes are occurring at all irradiation temperatures. At 412 °C-109 dpa, chromium enrichments - consistent with the α‧ phase - appear between the oxide precipitates, indicating radiation induced segregation. Grain boundaries, enriched with several elements including nickel and titanium, are observed at all irradiation conditions. At 412 °C-109 dpa, the grain boundaries are also enriched in molecular titanium oxide (TiO).

Combination pipe-rupture mitigator and in-vessel core catcher. [LMFBR

DOEpatents

Tilbrook, R.W.; Markowski, F.J.

1982-03-09

A device is described which mitigates against the effects of a failed coolant loop in a nuclear reactor by restricting the outflow of coolant from the reactor through the failed loop and by retaining any particulated debris from a molten core which may result from coolant loss or other cause. The device reduces the reverse pressure drop through the failed loop by limiting the access of coolant in the reactor to the inlet of the failed loop. The device also spreads any particulated core debris over a large area to promote cooling.

Comparison of computational results of the SABRE LMFBR pin bundle blockage code with data from well-instrumented out-of-pile test bundles (THORS bundles 3A and 5A)

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

Dearing, J.F.

The Subchannel Analysis of Blockages in Reactor Elements (SABRE) computer code, developed by the United Kingdom Atomic Energy Authority, is currently the only practical tool available for performing detailed analyses of velocity and temperature fields in the recirculating flow regions downstream of blockages in liquid-metal fast breeder reactor (LMFBR) pin bundles. SABRE is a subchannel analysis code; that is, it accurately represents the complex geometry of nuclear fuel pins arranged on a triangular lattice. The results of SABRE computational models are compared here with temperature data from two out-of-pile 19-pin test bundles from the Thermal-Hydraulic Out-of-Reactor Safety (THORS) Facility atmore » Oak Ridge National Laboratory. One of these bundles has a small central flow blockage (bundle 3A), while the other has a large edge blockage (bundle 5A). Values that give best agreement with experiment for the empirical thermal mixing correlation factor, FMIX, in SABRE are suggested. These values of FMIX are Reynolds-number dependent, however, indicating that the coded turbulent mixing correlation is not appropriate for wire-wrap pin bundles.« less

Development of martensitic steels for high neutron damage applications

NASA Astrophysics Data System (ADS)

Gelles, D. S.

1996-12-01

Martensitic stainless steels have been developed for both in-core applications in advanced liquid metal fast breeder reactors (LMFBR) and for first wall and structural materials applications for commercial fusion reactors. It can now be shown that these steels can be expected to maintain properties to levels as high as 175 or 200 dpa, respectively. The 12Cr1Mo0.5W0.2C alloy HT-9 has been extensively tested for LMFBR applications and shown to resist radiation damage, providing a creep and swelling resistant alternative to austenitic steels. Degradation of fracture toughness and Charpy impact properties have been observed, but properties are sufficient to provide reliable service. In comparison, alloys with lower chromium contents are found to decarburize in contact with liquid sodium and are therefore not recommended. Tungsten stabilized martensitic stainless steels have appropriate properties for fusion applications. Radioactivity levels are benign less than 500 years after service, radiation damage resistance is excellent, including impact properties, and swelling is modest. This report describes the history of the development effort.

Safety approach to the selection of design criteria for the CRBRP reactor refueling system

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

Meisl, C J; Berg, G E; Sharkey, N F

1979-01-01

The selection of safety design criteria for Liquid Metal Fast Breeder Reactor (LMFBR) refueling systems required the extrapolation of regulations and guidelines intended for Light Water Reactor refueling systems and was encumbered by the lack of benefit from a commercially licensed predecessor other than Fermi. The overall approach and underlying logic are described for developing safety design criteria for the reactor refueling system (RRS) of the Clinch River Breeder Reactor Plant (CRBRP). The complete selection process used to establish the criteria is presented, from the definition of safety functions to the finalization of safety design criteria in the appropriate documents.more » The process steps are illustrated by examples.« less

Vibrating fuel grapple. [LMFBR

DOEpatents

Chertock, A.J.; Fox, J.N.; Weissinger, R.B.

A reactor refueling method is described which utilizes a vibrating fuel grapple for removing spent fuel assemblies from a reactor core. It incorporates a pneumatic vibrator in the grapple head which allows additional withdrawal capability without exceeding the allowable axial force limit. The only moving part in the vibrator is a steel ball, pneumatically driven by a gas, such as argon, around a track, with centrifugal force created by the ball being transmitted through the grapple to the assembly handling socket.

Reactivity control assembly for nuclear reactor. [LMFBR

DOEpatents

Bollinger, L.R.

1982-03-17

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

Design criteria for a self-actuated shutdown system to ensure limitation of core damage. [LMFBR

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

Deane, N.A.; Atcheson, D.B.

1981-09-01

Safety-based functional requirements and design criteria for a self-actuated shutdown system (SASS) are derived in accordance with LOA-2 success criteria and reliability goals. The design basis transients have been defined and evaluated for the CDS Phase II design, which is a 2550 MWt mixed oxide heterogeneous core reactor. A partial set of reactor responses for selected transients is provided as a function of SASS characteristics such as reactivity worth, trip points, and insertion times.

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

Buchanan, J.R.; Keilholtz, G.W.

This report discusses the current status of liquid-metal fast breeder (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA). Bibliographic information on worldwide LMFBRs relative to the development of the breeder reactor as a safe source of nuclear power is presented for the period 1960 through 1969. The bibliography consists of 1560 abstracts covering early research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Key-word, author, and permuted-title indexes are included for completeness.

Microprocessor tester for the treat upgrade reactor trip system

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

Lenkszus, F.R.; Bucher, R.G.

1984-01-01

The upgrading of the Transient Reactor Test (TREAT) Facility at ANL-Idaho has been designed to provide additional experimental capabilities for the study of core disruptive accident (CDA) phenomena. In addition, a programmable Automated Reactor Control System (ARCS) will permit high-power transients up to 11,000 MW having a controlled reactor period of from 15 to 0.1 sec. These modifications to the core neutronics will improve simulation of LMFBR accident conditions. Finally, a sophisticated, multiply-redundant safety system, the Reactor Trip System (RTS), will provide safe operation for both steady state and transient production operating modes. To insure that this complex safety systemmore » is functioning properly, a Dedicated Microprocessor Tester (DMT) has been implemented to perform a thorough checkout of the RTS prior to all TREAT operations.« less

SASSYS pretest analysis of the THORS-SHRS experiments. [LMFBR

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

Bordner, G.L.; Dunn, F.E.

The THORS Facility at ORNL was recently modified to allow the testing of two parallel 19-pin simulated fueled subassemblies under natural circulation conditions similar to those that might occur during a partial failure of the shutdown heat removal system (SHRS) of a liquid-metal fast breeder reactor. The planned experimental program included a series of tests at various inlet plenum temperatures to determine boiling threshold power levels and the power range for stable boiling during natural circulation operation. Pretest calculations were performed at ANL, which supplement those carried out at ORNL for the purposes of validating the SASSYS model in themore » natural circulation regime and of providing data which would be useful in planning the experiments.« less

Reactor control rod timing system

DOEpatents

Wu, Peter T. K.

1982-01-01

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

Passive heat-transfer means for nuclear reactors. [LMFBR

DOEpatents

Burelbach, J.P.

1982-06-10

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

Magnetic switch for reactor control rod. [LMFBR

DOEpatents

Germer, J.H.

1982-09-30

A magnetic reed switch assembly is described for activating an electromagnetic grapple utilized to hold a control rod in position above a reactor core. In normal operation the magnetic field of a permanent magnet is short-circuited by a magnetic shunt, diverting the magnetic field away from the reed switch. The magnetic shunt is made of a material having a Curie-point at the desired release temperature. Above that temperature the material loses its ferromagnetic properties, and the magnetic path is diverted to the reed switch which closes and short-circuits the control circuit for the control rod electro-magnetic grapple which allows the control rod to drop into the reactor core for controlling the reactivity of the core.

Hanging core support system for a nuclear reactor. [LMFBR

DOEpatents

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

1984-04-26

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

Preliminary topical report on comparison reactor disassembly calculations

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

McLaughlin, T.P.

1975-11-01

Preliminary results of comparison disassembly calculations for a representative LMFBR model (2100-l voided core) and arbitrary accident conditions are described. The analytical methods employed were the computer programs: FX2- POOL, PAD, and VENUS-II. The calculated fission energy depositions are in good agreement, as are measures of the destructive potential of the excursions, kinetic energy, and work. However, in some cases the resulting fuel temperatures are substantially divergent. Differences in the fission energy deposition appear to be attributable to residual inconsistencies in specifying the comparison cases. In contrast, temperature discrepancies probably stem from basic differences in the energy partition models inherentmore » in the codes. Although explanations of the discrepancies are being pursued, the preliminary results indicate that all three computational methods provide a consistent, global characterization of the contrived disassembly accident. (auth)« less

Thermal-hydraulic posttest analysis for the ANL/MCTF 360/sup 0/ model heat-exchanger water test under mixed convection. [LMFBR

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

Yang, C.I.; Sha, W.T.; Kasza, K.E.

As a result of the uncertainties in the understanding of the influence of thermal-buoyancy effects on the flow and heat transfer in Liquid Metal Fast Breeder Reactor heat exchangers and steam generators under off-normal operating conditions, an extensive experimental program is being conducted at Argonne National Laboratory to eliminate these uncertainties. Concurrently, a parallel analytical effort is also being pursued to develop a three-dimensional transient computer code (COMMIX-IHX) to study and predict heat exchanger performance under mixed, forced, and free convection conditions. This paper presents computational results from a heat exchanger simulation and compares them with the results from amore » test case exhibiting strong thermal buoyancy effects. Favorable agreement between experiment and code prediction is obtained.« less

Input data requirements for special processors in the computation system containing the VENTURE neutronics code. [LMFBR

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

Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.

1979-07-01

User input data requirements are presented for certain special processors in a nuclear reactor computation system. These processors generally read data in formatted form and generate binary interface data files. Some data processing is done to convert from the user oriented form to the interface file forms. The VENTURE diffusion theory neutronics code and other computation modules in this system use the interface data files which are generated.

Cesium vapor cycle for an advanced LMFBR

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

Fraas, A.P.

1975-01-01

A review indicates that a cesium vapor topping cycle appears attractive for use in the intermediate fluid circuit of an advanced LMFBR designed for a reactor outlet temperature of 1250$sup 0$F or more and would have the following advantages: (1) it would increase the thermal efficiency by about 5 to 10 points (from approximately 40 percent to approximately 45 to 50 percent) thus reducing the amount of waste heat rejected to the environment by 15 to 30 percent. (2) the higher thermal efficiency should reduce the overall capital cost of the reactor plant in dollars per kilowatt. (3) the cesiummore » can be distilled out of the intermediate fluid circuit to leave it bone-dry, thus greatly reducing the time and cost of maintenance work (particularly for the steam generator). (4) the large volume and low pressure of the cesium vapor region in the cesium condenser-steam generator greatly reduces the magnitude of pressure fluctuations that might occur in the event of a leak in a steam generator tube, and the characteristics inherent in a condenser make it easy to design for rapid concentration of any noncondensibles that may form as a consequence of a steam leak into the cesium region so that a steam leak can be detected easily in the very early stages of its development. (auth)« less

Intermediate-Size Inducer Pump design report. [LMFBR

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

Boardman, T.J.

1979-06-15

This report summarizes the mechanical, structural, and hydrodynamic design of the Intermediate-Size Inducer Pump (ISIP). The design was performed under Atomics International's DOE Base Technology Program by the Atomics International and Rocketdyne Divisions of Rockwell International. The pump was designed to utilize the FFTF prototype pump frame as a test vehicle to test the inducer, impeller, and diffuser plus necessary adapter hardware under simulated Large Scale Liquid Metal Fast Breeder Reactor service conditions. The report describes the design requirements including the purpose and objectives, and discusses those design efforts and considerations made to meet the requirements. Included in the reportmore » are appendices showing calculative methods and results. Also included are overall assembly and layout drawings plus some details used as illustrations for discussion of the design results and the results of water tests performed on a model of the inducer.« less

Experiment Needs and Facilities Study Appendix A Transient Reactor Test Facility (TREAT) Upgrade

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

None

The TREAT Upgrade effort is designed to provide significant new capabilities to satisfy experiment requirements associated with key LMFBR Safety Issues. The upgrade consists of reactor-core modifications to supply the physics performance needed for the new experiments, an Advanced TREAT loop with size and thermal-hydraulics capabilities needed for the experiments, associated interface equipment for loop operations and handling, and facility modifications necessary to accommodate operations with the Loop. The costs and schedules of the tasks to be accomplished under the TREAT Upgrade project are summarized. Cost, including contingency, is about 10 million dollars (1976 dollars). A schedule for execution ofmore » 36 months has been established to provide the new capabilities in order to provide timely support of the LMFBR national effort. A key requirement for the facility modifications is that the reactor availability will not be interrupted for more than 12 weeks during the upgrade. The Advanced TREAT loop is the prototype for the STF small-bundle package loop. Modified TREAT fuel elements contain segments of graphite-matrix fuel with graded uranium loadings similar to those of STF. In addition, the TREAT upgrade provides for use of STF-like stainless steel-UO{sub 2} TREAT fuel for tests of fully enriched fuel bundles. This report will introduce the Upgrade study by presenting a brief description of the scope, performance capability, safety considerations, cost schedule, and development requirements. This work is followed by a "Design Description". Because greatly upgraded loop performance is central to the upgrade, a description is given of Advanced TREAT loop requirements prior to description of the loop concept. Performance requirements of the upgraded reactor system are given. An extensive discussion of the reactor physics calculations performed for the Upgrade concept study is provided. Adequate physics performance is essential for performance of experiments with the Advanced TREAT loop, and the stress placed on these calculations reflects this. Additional material on performance and safety is provided. Backup calculations on calculations of plutonium-release limits are described. Cost and schedule information for the Upgrade are presented.« less

Self-actuated shutdown system for a commercial size LMFBR. Final report

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

Dupen, C.F.G.

1978-08-01

A Self-Actuated Shutdown System (SASS) is defined as a reactor shutdown system in which sensors, release mechanisms and neutron absorbers are contained entirely within the reactor core structure, where they respond inherently to abnormal local process conditions, by shutting down the reactor, independently of the plant protection system (PPS). It is argued that a SASS, having a response time similar to that of the PPS, would so reduce the already very low probability of a failure-to-scram event that costly design features, derived from core disruptive accident analysis, could be eliminated. However, the thrust of the report is the feasibility andmore » reliability of the in-core SASS hardware to achieve sufficiently rapid shutdown. A number of transient overpower and transient undercooling-responsive systems were investigated leading to the selection of a primary candidate and a backup concept. During a transient undercooling event, the recommended device is triggered by the associated rate of change of pressure, whereas the alternate concept responds to the reduction in core pressure drop and requires calibration and adjustment by the operators to accommodate changes in reactor power.« less

Posttest examination of Sodium Loop Safety Facility experiments. [LMFBR

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

Holland, J.W.

In-reactor, safety experiments performed in the Sodium Loop Safety Facility (SLSF) rely on comprehensive posttest examinations (PTE) to characterize the postirradiation condition of the cladding, fuel, and other test-subassembly components. PTE information and on-line instrumentation data, are analyzed to identify the sequence of events and the severity of the accident for each experiment. Following in-reactor experimentation, the SLSF loop and test assembly are transported to the Hot Fuel Examination Facility (HFEF) for initial disassembly. Goals of the HFEF-phase of the PTE are to retrieve the fuel bundle by dismantling the loop and withdrawing the test assembly, to assess the macro-conditionmore » of the fuel bundle by nondestructive examination techniques, and to prepare the fuel bundle for shipment to the Alpha-Gamma Hot Cell Facility (AGHCF) at Argonne National Laboratory.« less

Liquid Metal Fast Breeder Reactors: a bibliography

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

Raleigh, H.D.

1980-11-01

This bibliography includes 5465 selected citations on LMFBR development. The citations were compiled from the DOE Energy Data Base covering the period January 1978 (EDB File No. 78R1087) through August 1980 (EDB File No. 80C79142). The references are to reports from the Department of Energy and its contractors, reports from other government or private organizations, and journal articles, books, conference papers, and monographs from US originators. Report citations are arranged alphanumerically by report number; nonreport literature citations are arranged chronologically. Corporate, Personal Author, Subject, and Report Number Indexes are provided in Volume 2.

Analysis of the OPERA 15-pin experiment with SABRE-2P. [LMFBR

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

Rose, S.D.; Carbajo, J.J.

The OPERA (Out-of-Pile Expulsion and Reentry Apparatus) experiment simulates the initial phase of a pump coastdown without scram of a liquid-metal fast breeder reactor, specifically the Fast Flux Test Facility. The test section is a 15-pin 60/sup 0/ triangular sector designed to simulate a full-size 61-pin hexagonal bundle. A previous study indicates this to be an adequate simulation. In this paper, experimental results from the OPERA 15-pin experiment performed at ANL in 1982 are compared to analytical calculations obtained with the SABRE-2P code at ORNL.

Liquid Metal Fast Breeder Reactors: a bibliography

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

Raleigh, H.D.

1980-11-01

This bibliogralphy includes 5465 selected citations on LMFBR development. The citations were compiled from the DOE Energy Data Base covering the period January 1978 (EDB File No. 78R1087) through August 1980 (EDB File No. 80C79142). The references are to reports from the Department of Energy and its contractors, reports from other government or private organizations, and journal articles, books, conference papers, and monographs from US originators. Report citations are arranged alphanumerically by report number; nonreport literature citations are arranged chronologically. Corporate, Personal Author, Subject, and Report Number Indexes are provided in Volume 2.

Atmospheric Dispersion of Sodium Aerosol due to a Sodium Leak in a Fast Breeder Reactor Complex

NASA Astrophysics Data System (ADS)

Punitha, G.; Sudha, A. Jasmin; Kasinathan, N.; Rajan, M.

Liquid sodium at high temperatures (470 K to 825 K) is used as the primary and secondary coolant in Liquid Metal cooled Fast Breeder Reactors (LMFBR). In the event of a postulated sodium leak in the Steam Generator Building (SGB) of a LMFBR, sodium readily combusts in the ambient air, especially at temperatures above 523 K. Intense sodium fire results and sodium oxide fumes are released as sodium aerosols. Sodium oxides are readily converted to sodium hydroxide in air due to the presence of moisture in it. Hence, sodium aerosols are invariably in the form of particulate sodium hydroxide. These aerosols damage not only the equipment and instruments due to their corrosive nature but also pose health hazard to humans. Hence, it is essential to estimate the concentration of sodium aerosols within the plant boundary for a sodium leak event. The Gaussian Plume Dispersion Model can obtain the atmospheric dispersion of sodium aerosols in an open terrain. However, this model does not give accurate results for dispersion in spaces close to the point of release and with buildings in between. The velocity field due to the wind is altered to a large extent by the intervening buildings and structures. Therefore, a detailed 3-D estimation of the velocity field and concentration has to be obtained through rigorous computational fluid dynamics (CFD) approach. PHOENICS code has been employed to determine concentration of sodium aerosols at various distances from the point of release. The dispersion studies have been carried out for the release of sodium aerosols at different elevations from the ground and for different wind directions.

Preliminary comparative assessment of land use for the Satellite Power System (SPS) and alternative electric energy technologies

NASA Technical Reports Server (NTRS)

Newsom, D. E.; Wolsko, T.

1980-01-01

A preliminary comparative assessment of land use for the satellite power system (SPS), other solar technologies, and alternative electric energy technologies was conducted. The alternative technologies are coal gasification/combined-cycle, coal fluidized-bed combustion (FBC), light water reactor (LWR), liquid metal fast breeder reactor (LMFBR), terrestrial photovoltaics (TPV), solar thermal electric (STE), and ocean thermal energy conversion (OTEC). The major issues of a land use assessment are the quantity, purpose, duration, location, and costs of the required land use. The phased methodology described treats the first four issues, but not the costs. Several past efforts are comparative or single technology assessment are reviewed briefly. The current state of knowledge about land use is described for each technology. Conclusions are drawn regarding deficiencies in the data on comparative land use and needs for further research.

LMFBR fuel assembly design for HCDA fuel dispersal

DOEpatents

Lacko, Robert E.; Tilbrook, Roger W.

1984-01-01

A fuel assembly for a liquid metal fast breeder reactor having an upper axial blanket region disposed in a plurality of zones within the fuel assembly. The characterization of a zone is dependent on the height of the axial blanket region with respect to the active fuel region. The net effect of having a plurality of zones is to establish a dispersal flow path for the molten materials resulting during a core meltdown accident. Upward flowing molten material can escape from the core region and/or fuel assembly without solidifying on the surface of fuel rods due to the heat sink represented by blanket region pellets.

Apparatus for controlling molten core debris. [LMFBR

DOEpatents

Golden, M.P.; Tilbrook, R.W.; Heylmun, N.F.

1977-07-19

Disclosed is an apparatus for containing, cooling, diluting, dispersing and maintaining subcritical the molten core debris assumed to melt through the bottom of a nuclear reactor pressure vessel in the unlikely event of a core meltdown. The apparatus is basically a sacrificial bed system which includes an inverted conical funnel, a core debris receptacle including a spherical dome, a spherically layered bed of primarily magnesia bricks, a cooling system of zig-zag piping in graphite blocks about and below the bed and a cylindrical liner surrounding the graphite blocks including a steel shell surrounded by firebrick. Tantalum absorber rods are used in the receptacle and bed. 9 claims, 22 figures.

Gravitational Agglomeration of Post-HCDA LMFBR Nonspherical Aerosols.

DTIC Science & Technology

1980-12-01

equations for two particle motions are developed . A computer program NGCEFF is constructed., the Navier-Stokes equation is solved by the finite difference...dynamic equations for two particle motions are developed . A computer program NGCEFF I is constructed, the Navier-Stokes equation is solved by the...spatial inhomogeneities for the aerosol. Thus, following an HCDA, an aerosol mixture of sodium compounds, fuel and core structural materials will

Nuclear Electric Magnetohydrodynamic Propulsion for Submarine

DTIC Science & Technology

1989-05-01

develop - ment strategies for the future. The base program includes the development of the LMFBR, and HTGR to...events. Oxide fuel is -134- being retained as a backup, pending the outcome of the metal fuel development program . The design allows for a quick fuel ... HTGR plants can be developed with much higher source temperature and core power density. High efficiency and low power den- sity characteristics

A microprocessor tester for the treat upgrade reactor trip system

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

Lenkszus, F.R.; Bucher, R.G.

1985-02-01

The upgrading of the Transient Reactor Test (TREAT) Facility at ANL-Idaho has been designed to provide additional experimental capabilities for the study of core disruptive accident (CDA) phenomena. To improve the analytical extrapolation of test results to full-size assembly bundles, the facility upgrade will increase the maximum size of the test bundle from 7 to 37 fuel pins. By creating a core convertor zone around the test location, the neutron spectrum incident on the test assembly will be hardened and the maximum energy deposited in the sample will be increased. In addition, a programmable Automated Reactor Control System (ARCS) willmore » permit high-power transients up to 11,000 MW having a controlled reactor period of from 15 to 0.1 sec. These modifications to the core neutronics will improve simulation of LMFBR accident conditions. Finally, a sophisticated, multiply-redundant safety system, the Reactor Trip System (RTS), will provide safe operation for both steady state and transient production operating modes. To insure that this complex safety system is functioning properly, a Dedicated Microprocessor Tester (DMT) has been implemented to perform a thorough checkout of the RTS prior to all TREAT operations. A quantitative reliability analysis of the RTS shows that the unreliability, that is, the probability of failure, is acceptable for a 10 hour mission time or risk interval.« less

Safety consequences of local initiating events in an LMFBR

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

Crawford, R.M.; Marr, W.W.; Padilla, A. Jr.

1975-12-01

The potential for fuel-failure propagation in an LMFBR at or near normal conditions is examined. Results are presented to support the conclusion that although individual fuel-pin failure may occur, rapid failure-propagation spreading among a large number of fuel pins in a subassembly is unlikely in an operating LMFBR. This conclusion is supported by operating experience, mechanistic analyses of failure-propagation phenomena, and experiments. In addition, some of the consequences of continued operation with defected fuel are considered.

Post-test analysis of dryout test 7B' of the W-1 Sodium Loop Safety Facility Experiment with the SABRE-2P code. [LMFBR

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

Rose, S.D.; Dearing, J.F.

An understanding of conditions that may cause sodium boiling and boiling propagation that may lead to dryout and fuel failure is crucial in liquid-metal fast-breeder reactor safety. In this study, the SABRE-2P subchannel analysis code has been used to analyze the ultimate transient of the in-core W-1 Sodium Loop Safety Facility experiment. This code has a 3-D simple nondynamic boiling model which is able to predict the flow instability which caused dryout. In other analyses dryout has been predicted for out-of-core test bundles and so this study provides additional confirmation of the model.

Occurence and prediction of sigma phase in fuel cladding alloys for breeder reactors. [LMFBR

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

Anantatmula, R.P.

1982-01-01

In sodium-cooled fast reactor systems, fuel cladding materials will be exposed for several thousand hours to liquid sodium. Satisfactory performance of the materials depends in part on the sodium compatibility and phase stability of the materials. This paper mainly deals with the phase stability aspect, with particular emphasis on sigma phase formation of the cladding materials upon extended exposures to liquid sodium. A new method of predicting sigma phase formation is proposed for austenitic stainless steels and predictions are compared with the experimental results on fuel cladding materials. Excellent agreement is obtained between theory and experiment. The new method ismore » different from the empirical methods suggested for superalloys and does not suffer from the same drawbacks. The present method uses the Fe-Cr-Ni ternary phase diagram for predicting the sigma-forming tendencies and exhibits a wide range of applicability to austenitic stainless steels and heat-resistant Fe-Cr-Ni alloys.« less

Studies on sodium boiling phenomena in out of pile rod bundles for various accidental situations in Liquid Metal Fast Breeder Reactors (LMFBR) experiments and interpretations

NASA Astrophysics Data System (ADS)

Seiler, J. M.; Rameau, B.

Bundle sodium boiling in nominal geometry for different accident conditions is reviewed. Voiding of a subassembly is controlled by not only hydrodynamic effects but mainly by thermal effects. There is a strong influence of the thermal inertia of the bundle material compared to the sodium thermal inertia. Flow instability, during a slow transient, can be analyzed with numerical tools and estimated using simplified approximations. Stable boiling operational conditions under bundle mixed convection (natural convection in the reactor) can be predicted. Voiding during a fast transient can be approximated from single channel calculations. The phenomenology of boiling behavior for a subassembly with inlet completely blocked, submitted to decay heat and lateral cooling; two-phase sodium flow pressure drop in a tube of large hydraulic diameter under adiabatic conditions; critical flow phenomena and voiding rate under high power, slow transient conditions; and onset of dry out under local boiling remains problematical.

Hockey-stick steam generator for LMFBR

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

Hallinan, G.J.; Svedlund, P.E.

1981-01-01

This paper presents the criteria and evaluation leading to the selection of the Hockey Stick Steam Generator Concept and subsequent development of that concept for LMFBR application. The selection process and development of the Modular Steam Generator (MSG) is discussed, including the extensive test programs that culminated in the manufacture and test of a 35 MW(t) Steam Generator. The design of the CRBRP Steam Generator is described, emphasizing the current status and a review of the critical structural areas. CRBRP steam generator development tests are evaluated, with a discussion of test objectives and rating of the usefulness of test resultsmore » to the CRBRP prototype design. Manufacturing experience and status of the CRBRP prototype and plant units is covered. The scaleup of the Hockey Stick concept to large commercial plant application is presented, with an evaluation of scaleup limitations, transient effects, and system design implications.« less

Summary and evaluation: fuel dynamics loss-of-flow experiments (tests L2, L3, and L4)

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

Barts, E.W.; Deitrich, L.W.; Eberhart, J.G.

1975-09-01

Three similar experiments conducted to support the analyses of hypothetical LMFBR unprotected-loss-of-flow accidents are summarized and evaluated. The tests, designated L2, L3, and L4, provided experimental data against which accident-analysis codes could be compared, so as to guide further analysis and modeling of the initiating phases of the hypothetical accident. The tests were conducted using seven-pin bundles of mixed-oxide fuel pins in Mark-II flowing-sodium loops in the TREAT reactor. Test L2 used fresh fuel. Tests L3 and L4 used irradiated fuel pins having, respectively, ''intermediate-power'' (no central void) and ''high-power'' (fully developed central void) microstructure. 12 references. (auth)

The Use of Thorium within the Nuclear Power Industry - 13472

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

Miller, Keith

2013-07-01

Thorium is 3 to 4 times more abundant than uranium and is widely distributed in nature as an easily exploitable resource in many countries. Unlike natural uranium, which contains ∼0.7% fissile {sup 235}U isotope, natural thorium does not contain any fissile material and is made up of the fertile {sup 232}Th isotope only. Therefore thorium and thorium-based fuel as metal, oxide or carbide, has been utilized in combination with fissile {sup 235}U or {sup 239}Pu in nuclear research and power reactors for conversion to fissile {sup 233}U, thereby enlarging fissile material resources. During the pioneering years of nuclear energy, frommore » the mid 1950's to mid 1970's, there was considerable interest worldwide to develop thorium fuels and fuel cycles in order to supplement uranium reserves. Thorium fuels and fuel cycles are particularly relevant to countries having large thorium deposits but very limited uranium reserves for their long term nuclear power programme. The feasibility of thorium utilization in high temperature gas cooled reactors (HTGR), light water reactors (LWR), pressurized heavy water reactors (PHWRs), liquid metal cooled fast breeder reactors (LMFBR) and molten salt breeder reactors (MSBR) were demonstrated. The initial enthusiasm for thorium fuels and fuel cycles was not sustained among the developing countries later, due to new discovery of uranium deposits and their improved availability. However, in recent times, the need for proliferation-resistance, longer fuel cycles, higher burnup, and improved waste form characteristics, reduction of plutonium inventories and in situ use of bred-in fissile material has led to renewed interest in thorium-based fuels and fuel cycles. (authors)« less

Evolution of thermo-physical properties and annealing of fast neutron irradiated boron carbide

NASA Astrophysics Data System (ADS)

Gosset, Dominique; Kryger, Bernard; Bonal, Jean-Pierre; Verdeau, Caroline; Froment, Karine

2018-03-01

Boron carbide is widely used as a neutron absorber in most nuclear reactors, in particular in fast neutron ones. The irradiation leads to a large helium production (up to 1022/cm3) together with a strong decrease of the thermal conductivity. In this paper, we have performed thermal diffusivity measurements and X-ray diffraction analyses on boron carbide samples coming from control rods of the French Phenix LMFBR reactor. The burnups range from 1021 to 8.1021/cm3. We first confirm the strong decrease of the thermal conductivity at the low burnup, together with high microstructural modifications: swelling, large micro-strains, high defects density, and disordered-like material conductivity. We observe the microstructural parameters are highly anisotropic, with high micro-strains and flattened coherent diffracting domains along the (00l) direction of the hexagonal structure. Performing heat treatments up to high temperature (2200 °C) allows us to observe the material thermal conductivity and microstructure restoration. It then appears the thermal conductivity healing is correlated to the micro-strain relaxation. We then assume the defects responsible for most of the damage are the helium bubbles and the associated stress fields.

TEMperature Pressure ESTimation of a homogeneous boiling fuel-steel mixture in an LMFBR core. [TEMPEST code

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

Pyun, J.J.; Majumdar, D.

The paper describes TEMPEST, a simple computer program for the temperature and pressure estimation of a boiling fuel-steel pool in an LMFBR core. The time scale of interest of this program is large, of the order of ten seconds. Further, the vigorous boiling in the pool will generate a large contact, and hence a large heat transfer between fuel and steel. The pool is assumed to be a uniform mixture of fuel and steel, and consequently vapor production is also assumed to be uniform throughout the pool. The pool is allowed to expand in volume if there is steel meltingmore » at the walls. In this program, the total mass of liquid and vapor fuel is always kept constant, but the total steel mass in the pool may change by steel wall melting. Because of a lack of clear understanding of the physical phenomena associated with the progression of a fuel-steel mixture at high temperature, various input options have been built-in to enable one to perform parametric studies. For example, the heat transfer from the pool to the surrounding steel structure may be controlled by input values for the heat transfer coefficients, or, the heat transfer may be calculated by a correlation obtained from the literature. Similarly, condensation of vapor on the top wall can be specified by input values of the condensation coefficient; the program can otherwise calculate condensation according to the non-equilibrium model predictions. Meltthrough rates of the surrounding steel walls can be specified by a fixed melt-rate or can be determined by a fraction of the heat loss that goes to steel-melting. The melted steel is raised to the pool temperature before it is joined with the pool material. Several applications of this program to various fuel-steel pools in the FFTF and the CRBR cores are discussed.« less

Heat transfer with hockey-stick steam generator. [LMFBR

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

Moody, E; Gabler, M J

1977-11-01

The hockey-stick modular design concept is a good answer to future needs for reliable, economic LMFBR steam generators. The concept was successfully demonstrated in the 30 Mwt MSG test unit; scaled up versions are currently in fabrication for CRBRP usage, and further scaling has been accomplished for PLBR applications. Design and performance characteristics are presented for the three generations of hockey-stick steam generators. The key features of the design are presented based on extensive analytical effort backed up by extensive ancillary test data. The bases for and actual performance evaluations are presented with emphasis on the CRBRP design. The designmore » effort on these units has resulted in the development of analytical techniques that are directly applicable to steam generators for any LMFBR application. In conclusion, the hockey-stick steam generator concept has been proven to perform both thermally and hydraulically as predicted. The heat transfer characteristics are well defined, and proven analytical techniques are available as are personnel experienced in their use.« less

JUPITER-II Program: ANL analysis of ZPPR-13A and ZPPR-13B

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

Collins, P.J.; Brumbach, S.B.

1984-08-09

The ZPPR-13 experiments provide basic physics data for radial-heterogeneous LMFBR cores of approximately 700 MWe size. Assemblies ZPPR-13A, ZPPR-13B and ZPPR-13C comprised the JUPITER-II cooperative program between US-DOE and PNC of Japan. The measurements were made between August 1982 and April 1984. This report describes in detail the results of the ANL analyses of phases 13A and 13B/1 and includes preliminary results for the later assemblies of phase 13B. The data were compiled primarily for discussions at the Third Jupiter Analysis Meeting to be held at ANL-West between September 11th and 14th, 1984.

Comparison of COBRA III-C and SABRE-1 (wire-wrap version) computational results with steady-state data from a 19-pin internally guard heated sodium-cooled bundle with a six-channel central blockage (THORS Bundle 3C). [LMFBR

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

Dearing, J F; Nelson, W R; Rose, S D

Computational thermal-hydraulic models of a 19-pin, electrically heated, wire-wrap liquid-metal fast breeder reactor test bundle were developed using two well-known subchannel analysis codes, COBRA III-C and SABRE-1 (wire-wrap version). These two codes use similar subchannel control volumes for the finite difference conservation equations but vary markedly in solution strategy and modeling capability. In particular, the empirical wire-wrap-forced diversion crossflow models are different. Surprisingly, however, crossflow velocity predictions of the two codes are very similar. Both codes show generally good agreement with experimental temperature data from a test in which a large radial temperature gradient was imposed. Differences between data andmore » code results are probably caused by experimental pin bowing, which is presently the limiting factor in validating coded empirical models.« less

Start-up control system and vessel for LMFBR

DOEpatents

Durrant, Oliver W.; Kakarala, Chandrasekhara R.; Mandel, Sheldon W.

1987-01-01

A reflux condensing start-up system includes a steam generator, a start-up vessel connected parallel to the steam generator, a main steam line connecting steam outlets of the steam generator and start-up vessel to a steam turbine, a condenser connected to an outlet of the turbine and a feedwater return line connected between the condenser and inlets of the steam generator and start-up vessel. The start-up vessel has one or more heaters at the bottom thereof for heating feedwater which is supplied over a start-up line to the start-up vessel. Steam is thus generated to pressurize the steam generator before the steam generator is supplied with a heat transfer medium, for example liquid sodium, in the case of a liquid metal fast breeder reactor. The start-up vessel includes upper and lower bulbs with a smaller diameter mid-section to act as water and steam reservoirs. The start-up vessel can thus be used not only in a start-up operation but as a mixing tank, a water storage tank and a level control at low loads for controlling feedwater flow.

Start-up control system and vessel for LMFBR

DOEpatents

Durrant, Oliver W.; Kakarala, Chandrasekhara R.; Mandel, Sheldon W.

1987-01-01

A reflux condensing start-up system comprises a steam generator, a start-up vessel connected parallel to the steam generator, a main steam line connecting steam outlets of the steam generator and start-up vessel to a steam turbine, a condenser connected to an outlet of the turbine and a feedwater return line connected between the condenser and inlets of the steam generator and start-up vessel. The start-up vessel has one or more heaters at the bottom thereof for heating feedwater which is supplied over a start-up line to the start-up vessel. Steam is thus generated to pressurize the steam generator before the steam generator is supplied with a heat transfer medium, for example liquid sodium, in the case of a liquid metal fast breeder reactor. The start-up vessel includes upper and lower bulbs with a smaller diameter mid-section to act as water and steam reservoirs. The start-up vessel can thus be used not only in a start-up operation but as a mixing tank, a water storage tank and a level control at low loads for controlling feedwater flow.

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

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

Not Available

1979-06-01

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

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

None

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

Seismic design verification of LMFBR structures

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

Not Available

1977-07-01

The report provides an assessment of the seismic design verification procedures currently used for nuclear power plant structures, a comparison of dynamic test methods available, and conclusions and recommendations for future LMFB structures.

MSG test report: removal of residual sodium. [LMFBR

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

Harty, R.B.

1974-03-08

This report presents the results of cleaning activities performed to remove residual sodium from the AI Modular Steam Generator. A description of the cleaning loop, cleaning procedure, results, and visual inspection are included.

Siegel[JMMM 7,312(`78)] FIRST EXPERIMENTAL DISCOVERY of Giant-Magnetoresistance Decade Pre ``Fert'' and ``Gruenberg'' ['88 - `78] = 10-Years = One-Decade Sounds, for Nuclear-Power Naïve ``Panacea'' for Global-Warming/Climate-Chan

NASA Astrophysics Data System (ADS)

Hoffmann, Masterace; Siegel, Edward

Siegel[JMMM 7,312(`78); Monju (12/'95) LMFBR PREDICTION!!!] following: Wigner[JAP 17,857(`46)]-(Alvin)Weinberg(ANL/ORNL/ANS)-(Sidney)Siegel(ANL/ORNL/ANS)-Seitz-Overhauser-Rollnick-Pollard-Lofaro-Markey-Pringle[Nuclear-PowerFrom Physics to Politics(`79)] FIRST EXPERIMENTAL DISCOVERY [Siegel<<<''Fert''-''Gruenberg'':2007-Physics-Nobel/2006:-Wolf/Japan-prizes:[`88 -`78] =10-years =1-decade precedence!!!] of granular giant-magnetoresistance(GMR) [Google: ``EDWARD SIEGEL GIANT-MAGNETORESISTANCE ICMAO 1977 FLICKER''] [Google: ``Ana Mayo If LEAKS`Could' KILL''] in austenitic/FCC Ni/Fe-based (so MIScalled)''super''alloy-182/82 transition-welds GENERIC ENDEMIC EXTANT detrimental (SYNONYMS): Wigner's-disease/Ostwald-ripening/spinodal-decompositio/OVERageing-EMBRITTLEMENT/THERMAL-leading-to-mechanical (TLTM)-INstability/``sensitization'' in: nuclear-reactors/spent-fuel dry-casks/refineries/jet/missile/rocket-engines/...SOUNDS A DIRE WARNING FOR NAIVE Hansen-Sommerville-Holdren-DOE-NRC-OSTP-WNA-NEI-AIP-APS-...calls/media-hype/P.R./spin-doctoring for carbon-``free'' nuclear-power as a SUPPOSED ``panacea'' for climate-change/global-warming: ``TRUST BUT VERIFY!!!'' ; a VERY LOUD CAVEAT EMPTOR!!!

SP-100 Program: space reactor system and subsystem investigations

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

Harty, R.B.

1983-09-30

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

SP-100 program: Space reactor system and subsystem investigations

NASA Astrophysics Data System (ADS)

Harty, R. B.

1983-09-01

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

Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Final report

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

Todreas, N.E.; Cheng, S.K.; Basehore, K.

1984-08-01

This project principally undertook the investigation of the thermal hydraulic performance of wire wrapped fuel bundles of LMFBR configuration. Results obtained included phenomenological models for friction factors, flow split and mixing characteristics; correlations for predicting these characteristics suitable for insertion in design codes; numerical codes for analyzing bundle behavior both of the lumped subchannel and distributed parameter categories and experimental techniques for pressure velocity, flow split, salt conductivity and temperature measurement in water cooled mockups of bundles and subchannels. Flow regimes investigated included laminar, transition and turbulent flow under forced convection and mixed convection conditions. Forced convections conditions were emphasized.more » Continuing efforts are underway at MIT to complete the investigation of the mixed convection regime initiated here. A number of investigations on outlet plenum behavior were also made. The reports of these investigations are identified.« less

Summary of NR Program Prometheus Efforts

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

J Ashcroft; C Eshelman

2006-02-08

The Naval Reactors Program led work on the development of a reactor plant system for the Prometheus space reactor program. The work centered on a 200 kWe electric reactor plant with a 15-20 year mission applicable to nuclear electric propulsion (NEP). After a review of all reactor and energy conversion alternatives, a direct gas Brayton reactor plant was selected for further development. The work performed subsequent to this selection included preliminary nuclear reactor and reactor plant design, development of instrumentation and control techniques, modeling reactor plant operational features, development and testing of core and plant material options, and development ofmore » an overall project plan. Prior to restructuring of the program, substantial progress had been made on defining reference plant operating conditions, defining reactor mechanical, thermal and nuclear performance, understanding the capabilities and uncertainties provided by material alternatives, and planning non-nuclear and nuclear system testing. The mission requirements for the envisioned NEP missions cannot be accommodated with existing reactor technologies. Therefore concurrent design, development and testing would be needed to deliver a functional reactor system. Fuel and material performance beyond the current state of the art is needed. There is very little national infrastructure available for fast reactor nuclear testing and associated materials development and testing. Surface mission requirements may be different enough to warrant different reactor design approaches and development of a generic multi-purpose reactor requires substantial sacrifice in performance capability for each mission.« less

78 FR 35056 - Effectiveness of the Reactor Oversight Process Baseline Inspection Program

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-11

... NUCLEAR REGULATORY COMMISSION [NRC-2013-0125] Effectiveness of the Reactor Oversight Process... the effectiveness of the reactor oversight process (ROP) baseline inspection program with members of... Nuclear Reactor Regulations, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone: 301...

A Review of Gas-Cooled Reactor Concepts for SDI Applications

DTIC Science & Technology

1989-08-01

710 program .) Wire- Core Reactor (proposed by Rockwell). The wire- core reactor utilizes thin fuel wires woven between spacer wires to form an open...reactor is based on results of developmental studies of nuclear rocket propulsion systems. The reactor core is made up of annular fuel assemblies of...XE Addendum to Volume II. NERVA Fuel Development , Westinghouse Astronuclear Laboratory, TNR-230, July 15’ 1972. J I8- Rover Program Reactor Tests

AI-MSG modification work plan. [LMFBR

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

Page, J.P.

1973-08-20

This document contains the Work Plan for the modification of the AI Steam Generator for tests in Large Leak Test Rig. This Work Plan describes the objectives, scope of work, schedule and manpower, end items, and meetings and reports required for the modification.

78 FR 63516 - Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-24

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0134] Initial Test Program of Emergency Core Cooling....79.1, ``Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors.'' This... emergency core cooling systems (ECCSs) for boiling- water reactors (BWRs) whose licenses are issued after...

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

Mulder, R. U.; Benneche, P. E.; Hosticka, B.

The objective of the DOE supported Reactor Sharing Program is to increase the availability of university nuclear reactor facilities to non-reactor-owning educational institutions. The educational and research programs of these users institutions is enhanced by the use of the nuclear facilities.

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

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

None, None

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

High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1982

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

Kasten, P.R.; Rittenhouse, P.L.; Bartine, D.E.

1983-06-01

During 1982 the High-Temperature Gas-Cooled Reactor (HTGR) Technology Program at Oak Ridge National Laboratory (ORNL) continued to develop experimental data required for the design and licensing of cogeneration HTGRs. The program involves fuels and materials development (including metals, graphite, ceramic, and concrete materials), HTGR chemistry studies, structural component development and testing, reactor physics and shielding studies, performance testing of the reactor core support structure, and HTGR application and evaluation studies.

NASA-EPA automotive thermal reactor technology program

NASA Technical Reports Server (NTRS)

Blankenship, C. P.; Hibbard, R. R.

1972-01-01

The status of the NASA-EPA automotive thermal reactor technology program is summarized. This program is concerned primarily with materials evaluation, reactor design, and combustion kinetics. From engine dynamometer tests of candidate metals and coatings, two ferritic iron alloys (GE 1541 and Armco 18-SR) and a nickel-base alloy (Inconel 601) offer promise for reactor use. None of the coatings evaluated warrant further consideration. Development studies on a ceramic thermal reactor appear promising based on initial vehicle road tests. A chemical kinetic study has shown that gas temperatures of at least 900 K to 1000 K are required for the effective cleanup of carbon monoxide and hydrocarbons, but that higher temperatures require shorter combustion times and thus may permit smaller reactors.

Pre-Licensing Evaluation of Legacy SFR Metallic Fuel Data

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

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

2016-09-16

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

Expanded scope of training and education programs at the UFTR

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

Vernetson, W.G.; Whaley, P.M.

1985-01-01

Historically, the University of Florida Training Reactor (UFTR) has been used to train both hot and cold license reactor operator candidates in intensive two- and three-week training programs consisting of a correlated set of classroom lectures, hands-on reactor operations, and laboratory exercises. These training programs provide nuclear plant operating staff with fundamental operational experience in understanding, controlling, and evaluating subcritical multiplication, reactivity effects, reactivity manipulations, and reactor operations; a sufficient number of startups and shutdowns is also assured. The UDTR is also used in a nuclear engineering course entitled ''Principles of Nuclear Reactor Operations.'' The purpose of this paper ismore » to report the results of efforts to redirect and refine tractor operations educational and training programs at the UFTR.« less

The RERTR Program status and progress

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

Travelli, A.

1995-12-01

The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. The major events, findings, and activities of 1995 are reviewed after a brief summary of the results which the RERTR Program had achieved by the end of 1994. The revelation that Iraq was on the verge of developing a nuclear weapon at the time of the Gulf War, and that it was planning to do so by extracting HEU from the fuel of its research reactors, has given new impetus and urgency to the RERTR commitment of eliminating HEU use in research and test reactors worldwide.more » Development of advanced LEU research reactor fuels is scheduled to begin in October 1995. The Russian RERTR program, which aims to develop and demonstrate within the next five years the technical means needed to convert Russian-supplied research reactors to LEU fuels, is now in operation. A Statement of Intent was signed by high US and Chinese officials, endorsing cooperative activities between the RERTR program and Chinese laboratories involved in similar activities. Joint studies of LEU technical feasibility were completed for the SAFARI-I reactor in South Africa and for the ANS reactor in the US. A new study has been initiated for the FRM-II reactor in Germany. Significant progress was made on several aspects of producing {sup 99}Mo from fission targets utilizing LEU instead of HEU. A cooperation agreements is in place with the Indonesian BATAN. The first prototypical irradiation of an LEU metal-foil target for {sup 99}Mo production was accomplished in Indonesia. The TR-2 reactor, in Turkey, began conversion. SAPHIR, in Switzerland, was shut down. LEU fuel fabrication has begun for the conversion of two more US reactors. Twelve foreign reactors and nine domestic reactors have been fully converted. Approximately 60 % of the work required to eliminate the use of HEU in US-supplied research reactors has been accomplished.« less

10 CFR Appendix D to Part 73 - Physical Protection of Irradiated Reactor Fuel in Transit, Training Program Subject Schedule

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Physical Protection of Irradiated Reactor Fuel in Transit... Irradiated Reactor Fuel in Transit, Training Program Subject Schedule Pursuant to the provision of § 73.37 of... reactor fuel is required to assure that individuals used as shipment escorts have completed a training...

10 CFR Appendix D to Part 73 - Physical Protection of Irradiated Reactor Fuel in Transit, Training Program Subject Schedule

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Physical Protection of Irradiated Reactor Fuel in Transit... Irradiated Reactor Fuel in Transit, Training Program Subject Schedule Pursuant to the provision of § 73.37 of... reactor fuel is required to assure that individuals used as shipment escorts have completed a training...

SNAP (Space Nuclear Auxiliary Power) reactor overview. Final report, June 1982-December 1983

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

Voss, S.S.

1984-08-01

The SNAP reactor programs are outlined in this report. A summary of the program is included along with a technical outline of the SER, S2DR, SNAP 10A/SNAPSHOT, S8ER, and S8DR reactor systems. Specifications of the designs, the design logic and a conclusion outlining some of the program weaknesses are given.

N Reactor Deactivation Program Plan. Revision 4

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

Walsh, J.L.

1993-12-01

This N Reactor Deactivation Program Plan is structured to provide the basic methodology required to place N Reactor and supporting facilities {center_dot} in a radiologically and environmentally safe condition such that they can be decommissioned at a later date. Deactivation will be in accordance with facility transfer criteria specified in Department of Energy (DOE) and Westinghouse Hanford Company (WHC) guidance. Transition activities primarily involve shutdown and isolation of operational systems and buildings, radiological/hazardous waste cleanup, N Fuel Basin stabilization and environmental stabilization of the facilities. The N Reactor Deactivation Program covers the period FY 1992 through FY 1997. The directivemore » to cease N Reactor preservation and prepare for decommissioning was issued by DOE to WHC on September 20, 1991. The work year and budget data supporting the Work Breakdown Structure in this document are found in the Activity Data Sheets (ADS) and the Environmental Restoration Program Baseline, that are prepared annually.« less

Experiences in utilization of research reactors in Yugoslavia

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

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

1971-06-15

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

The Shock and Vibration Digest. Volume 18, Number 6

DTIC Science & Technology

1986-06-01

linear, quadratic, or cubic. Bessel function Reed [124] reported a method for computing solutions were obtained for a truncated pyramid amplitudes of a...86-1198 A. Ragab, Chung C. Fu Seismic Analysis of a Large LMFBR with Flu- Cairo Univ., Giza , Egypt . . *. id-Structure Imteractions Computers Struc

A liquid-metal filling system for pumped primary loop space reactors

NASA Astrophysics Data System (ADS)

Crandall, D. L.; Reed, W. C.

Some concepts for the SP-100 space nuclear power reactor use liquid metal as the primary coolant in a pumped loop. Prior to filling ground engineering test articles or reactor systems, the liquid metal must be purified and circulated through the reactor primary system to remove contaminants. If not removed, these contaminants enhance corrosion and reduce reliability. A facility was designed and built to support Department of Energy Liquid Metal Fast Breeder Reactor tests conducted at the Idaho National Engineering Laboratory. This test program used liquid sodium to cool nuclear fuel in in-pile experiments; thus, a system was needed to store and purify sodium inventories and fill the experiment assemblies. This same system, with modifications and potential changeover to lithium or sodium-potassium (NaK), can be used in the Space Nuclear Power Reactor Program. This paper addresses the requirements, description, modifications, operation, and appropriateness of using this liquid-metal system to support the SP-100 space reactor program.

F Reactor Inspection

ScienceCinema

Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

2018-01-16

Workers from Mission Support Alliance, LLC., removed the welds around the steel door of the F Reactor before stepping inside the reactor to complete its periodic inspection. This is the first time the Department of Energy (DOE) has had the reactor open since 2008. The F Reactor is one of nine reactors along the Columbia River at the Department's Hanford Site in southeastern Washington State, where environmental cleanup has been ongoing since 1989. As part of the Tri-Party Agreement, the Department completes surveillance and maintenance activities of cocooned reactors periodically to evaluate the structural integrity of the safe storage enclosure and to ensure confinement of any remaining hazardous materials. "This entry marks a transition of sorts because the Hanford Long-Term Stewardship Program, for the first time, was responsible for conducting the entry and surveillance and maintenance activities," said Keith Grindstaff, Energy Department Long-Term Stewardship Program Manager. "As the River Corridor cleanup work is completed and transitioned to long-term stewardship, our program will manage any on-going requirements."

F Reactor Inspection

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

Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

2014-10-29

Workers from Mission Support Alliance, LLC., removed the welds around the steel door of the F Reactor before stepping inside the reactor to complete its periodic inspection. This is the first time the Department of Energy (DOE) has had the reactor open since 2008. The F Reactor is one of nine reactors along the Columbia River at the Department's Hanford Site in southeastern Washington State, where environmental cleanup has been ongoing since 1989. As part of the Tri-Party Agreement, the Department completes surveillance and maintenance activities of cocooned reactors periodically to evaluate the structural integrity of the safe storage enclosuremore » and to ensure confinement of any remaining hazardous materials. "This entry marks a transition of sorts because the Hanford Long-Term Stewardship Program, for the first time, was responsible for conducting the entry and surveillance and maintenance activities," said Keith Grindstaff, Energy Department Long-Term Stewardship Program Manager. "As the River Corridor cleanup work is completed and transitioned to long-term stewardship, our program will manage any on-going requirements."« less

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

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

Kelley, D.B.

1995-12-31

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

Automatic reactor model synthesis with genetic programming.

PubMed

Dürrenmatt, David J; Gujer, Willi

2012-01-01

Successful modeling of wastewater treatment plant (WWTP) processes requires an accurate description of the plant hydraulics. Common methods such as tracer experiments are difficult and costly and thus have limited applicability in practice; engineers are often forced to rely on their experience only. An implementation of grammar-based genetic programming with an encoding to represent hydraulic reactor models as program trees should fill this gap: The encoding enables the algorithm to construct arbitrary reactor models compatible with common software used for WWTP modeling by linking building blocks, such as continuous stirred-tank reactors. Discharge measurements and influent and effluent concentrations are the only required inputs. As shown in a synthetic example, the technique can be used to identify a set of reactor models that perform equally well. Instead of being guided by experience, the most suitable model can now be chosen by the engineer from the set. In a second example, temperature measurements at the influent and effluent of a primary clarifier are used to generate a reactor model. A virtual tracer experiment performed on the reactor model has good agreement with a tracer experiment performed on-site.

Production assurance program strategy for N Reactor balance of plant systems

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

House, R.D.; Bitten, E.J.; Keenan, J.P.

1986-03-18

A production assurance program has been established for N Reactor, a dual purpose reactor plant, operated to produce special nuclear materials and steam for electricity. N Reactor, which began operation in December 1963, is now approaching the end of its design life. This paper describes the two phase program for Balance of Plant (BOP) systems. The Phase I evaluation has been completed and indications are that the lifetime of systems and components could be extended by implementing appropriate surveillance, operations and maintenance strategies. In Phase II, a thorough evaluation of components and systems is underway and action items are beingmore » identified which will allow component and system extended operation.« less

Proceedings of the 1994 international meeting on reduced enrichment for research and test reactors

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

NONE

1997-08-01

This meeting brought together participants in the international effort to minimize and eventually eliminate the use of highly enriched uranium in civilian nuclear programs. Papers cover the following topics: National programs; fuel cycle; nuclear fuels; analyses; advanced reactors; and reactor conversions. Selected papers have been indexed separately for inclusion to the Energy Science and Technology Database.

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

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

Bess, John D.

2014-03-01

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

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

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

John D. Bess

2013-03-01

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

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

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

John D. Bess

2013-03-01

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

MONTE CARLO SIMULATIONS OF PERIODIC PULSED REACTOR WITH MOVING GEOMETRY PARTS

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

Cao, Yan; Gohar, Yousry

2015-11-01

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

Observed Changes in As-Fabricated U-10Mo Monolithic Fuel Microstructures After Irradiation in the Advanced Test Reactor

NASA Astrophysics Data System (ADS)

Keiser, Dennis; Jue, Jan-Fong; Miller, Brandon; Gan, Jian; Robinson, Adam; Madden, James

2017-12-01

A low-enriched uranium U-10Mo monolithic nuclear fuel is being developed by the Material Management and Minimization Program, earlier known as the Reduced Enrichment for Research and Test Reactors Program, for utilization in research and test reactors around the world that currently use high-enriched uranium fuels. As part of this program, reactor experiments are being performed in the Advanced Test Reactor. It must be demonstrated that this fuel type exhibits mechanical integrity, geometric stability, and predictable behavior to high powers and high fission densities in order for it to be a viable fuel for qualification. This paper provides an overview of the microstructures observed at different regions of interest in fuel plates before and after irradiation for fuel samples that have been tested. These fuel plates were fabricated using laboratory-scale fabrication methods. Observations regarding how microstructural changes during irradiation may impact fuel performance are discussed.

Progress in space nuclear reactor power systems technology development - The SP-100 program

NASA Technical Reports Server (NTRS)

Davis, H. S.

1984-01-01

Activities related to the development of high-temperature compact nuclear reactors for space applications had reached a comparatively high level in the U.S. during the mid-1950s and 1960s, although only one U.S. nuclear reactor-powered spacecraft was actually launched. After 1973, very little effort was devoted to space nuclear reactor and propulsion systems. In February 1983, significant activities toward the development of the technology for space nuclear reactor power systems were resumed with the SP-100 Program. Specific SP-100 Program objectives are partly related to the determination of the potential performance limits for space nuclear power systems in 100-kWe and 1- to 100-MW electrical classes. Attention is given to potential missions and applications, regimes of possible space power applicability, safety considerations, conceptual system designs, the establishment of technical feasibility, nuclear technology, materials technology, and prospects for the future.

The United Arab Emirates Nuclear Program and Proposed U.S. Nuclear Cooperation

DTIC Science & Technology

2009-10-28

global efforts to prevent nuclear proliferation” and, “the establishment of reliable sources of nuclear fuel for future civilian light water reactors ...nuclear reactor or on handling spent reactor fuel. (...continued) May 4, 2008; and, Chris...related to the UAE’s proposed nuclear program has already taken place. In August 2008, Virginia’s Thorium Power Ltd. signed two consulting and

The United Arab Emirates Nuclear Program and Proposed U.S. Nuclear Cooperation

DTIC Science & Technology

2009-07-17

global efforts to prevent nuclear proliferation” and, “the establishment of reliable sources of nuclear fuel for future civilian light water reactors ...planned nuclear reactor or on handling spent reactor fuel. (...continued) May 4, 2008...contracting between U.S. firms and the UAE related to the UAE’s proposed nuclear program has already taken place. In August 2008, Virginia’s Thorium Power

The United Arab Emirates Nuclear Program and Proposed U.S. Nuclear Cooperation

DTIC Science & Technology

2009-12-23

reactors deployed” in the UAE. Some Members of Congress had welcomed the UAE government’s stated commitments not to pursue proliferation-sensitive...for the planned nuclear reactor or on handling spent reactor fuel. (...continued) May...firms and the UAE related to the UAE’s proposed nuclear program has already taken place. In August 2008, Virginia’s Thorium Power Ltd. signed two

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

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

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

PATRAM '80. Proceedings. Volume 1

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

Huebner, H.W.

Volume 1 contains papers from the following sessions: Plenary Session; Regulations, Licensing and Standards; LMFBR Systems Concepts; Risk/Safety Assessment I; Systems and Package Design; US Institutional Issues; Risk/Safety Assessment II; Leakage, Leak Rate and Seals; Poster Session A; Operations and Systems Experience I; Manufacturing Processes and Materials; and Quality Assurance and Maintenance. Individual papers were processed. (LM)

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

Romander, C.M.

The potential hazard of HCDA-generated missiles is analyzed, and the current status of the potential hazards of sodium fires is summarized. Simple analyses are performed to determine lower bounds on the HCDA energetics required to generate missiles that could reach the secondary containment structure of a 1000-MWe LMFBR. The potential missiles considered include the vessel head, components mounted on the head, and conrol rods.

Testimony of Fred R. Mynatt before the Energy Research and Development Subcommittee of the Committee on Science, Space, and Technology, US House of Representatives. [Advanced fuel technology, gas-cooled reactor technology, and liquid metal-cooled reactor technology programs

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

Mynatt, F.R.

1987-03-18

This report provides a description of the statements submitted for the record to the committee on Science, Space, and Technology of the United States House of Representatives. These statements describe three principal areas of activity of the Advanced Reactor Technology Program of the Department of Energy (DOE). These areas are advanced fuel cycle technology, modular high-temperature gas-cooled reactor technology, and liquid metal-cooled reactor. The areas of automated reactor control systems, robotics, materials and structural design shielding and international cooperation were included in these statements describing the Oak Ridge National Laboratory's efforts in these areas. (FI)

Georgia Institute of Technology research on the Gas Core Actinide Transmutation Reactor (GCATR)

NASA Technical Reports Server (NTRS)

Clement, J. D.; Rust, J. H.; Schneider, A.; Hohl, F.

1976-01-01

The program reviewed is a study of the feasibility, design, and optimization of the GCATR. The program is designed to take advantage of initial results and to continue work carried out on the Gas Core Breeder Reactor. The program complements NASA's program of developing UF6 fueled cavity reactors for power, nuclear pumped lasers, and other advanced technology applications. The program comprises: (1) General Studies--Parametric survey calculations performed to examine the effects of reactor spectrum and flux level on the actinide transmutation for GCATR conditions. The sensitivity of the results to neutron cross sections are to be assessed. Specifically, the parametric calculations of the actinide transmutation are to include the mass, isotope composition, fission and capture rates, reactivity effects, and neutron activity of recycled actinides. (2) GCATR Design Studies--This task is a major thrust of the proposed research program. Several subtasks are considered: optimization criteria studies of the blanket and fuel reprocessing, the actinide insertion and recirculation system, and the system integration. A brief review of the background of the GCATR and ongoing research is presented.

The RERTR Program : a status report.

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

Travelli, A.

1998-10-19

This paper describes the progress achieved by the Reduced Enrichment for Research and Test Reactors (RERTR) Program in collaboration with its many international partners since its inception in 1978. A brief summary of the results that the program had attained by the end of 1997 is followed by a detailed review of the major events, findings, and activities that took place in 1998. The past year was characterized by exceptionally important accomplishments and events for the RERTR program. Four additional shipments of spent fuel from foreign research reactors were accepted by the U.S. Altogether, 2,231 spent fuel assemblies from foreignmore » research reactors have been received by the U.S. under the acceptance policy. Fuel development activities began to yield solid results. Irradiations of the first two batches of microplates were completed. Preliminary postirradiation examinations of these microplates indicate excellent irradiation behavior of some of the fuel materials that were tested. These materials hold the promise of achieving the pro am goal of developing LEU research reactor fuels with uranium density in the 8-9 g /cm{sup 3} range. Progress was made in the Russian RERTR program, which aims to develop and demonstrate the technical means needed to convert Russian-supplied research reactors to LEU fuels. Feasibility studies for converting to LEU fuel four Russian-designed research reactors (IR-8 in Russia, Budapest research reactor in Hungary, MARIA in Poland, and WWR-SM in Uzbekistan) were completed. A new program activity began to study the feasibility of converting three Russian plutonium production reactors to the use of low-enriched U0{sub 2}-Al dispersion fuel, so that they can continue to produce heat and electricity without producing significant amounts of plutonium. The study of an alternative LEU core for the FRM-II design has been extended to address, with favorable results, the transient performance of the core under hypothetical accident conditions. A major milestone was accomplished in the development of a process to produce molybdenum-99 from fission targets utilizing LEU instead of HEU. Targets containing LEU metal foils were irradiated in the RAS-GAS reactor at BATAN, Indonesia, and molybdenum-99 was successfully extracted through the ensuing process. These are exciting times for the program and for all those involved in it, and last year's successes augur well for the future. However, as in the past, the success of the RERTR program will depend on the international friendship and cooperation that have always been its trademark.« less

The U.S. RERTR program status and progress.

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

Travelli, A.

1998-01-21

The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program since its inception in 1978 is described. A brief summary of the results which the RERTR Program had achieved by the end of 1996 in collaboration with its many international partners is followed by a detailed review of the major events, findings, and activities of 1997. Significant progress has been made during the past year. In the area of U.S. acceptance of spent fuel from foreign research reactors, several shipments have taken place and additional are being planned. Intense fuel development activities are in progress, including procurement ofmore » equipment, screening of candidate materials, and production of microplates. Irradiation of the first series of microplates began in August 1997 in the Advanced Test Reactor, in Idaho. Progress has been made in the Russian RERTR program, which aims to develop and demonstrate within five years the technical means needed to convert Russian-supplied research reactors to LEU fuels. The study of an alternative LEU core for the FRM-II design has been extended to address, with favorable results, controversial performance issues which were raised at last year's meeting. Progress was also made on several aspects of producing molybdenum-99 from fission targets utilizing LEU instead of HEU. Various types of targets and processes are being pursued, with FDA approval of an LEU process projected to occur within two years. The feasibility of LEU Fuel conversion for three important DOE research reactors (BMRR, HFBR, and HFIR) has been evaluated by the RERTR program. In spite of the many momentous events which have occurred during the intervening years, and the excellent progress achieved, the most important challenges that the RERTR program faces today are not very different in type from those that were faced during the first RERTR meeting. Now, as then, the most important task is to develop new LEU fuels satisfying requirements which cannot be satisfied by any existing fuel. These new advanced fuels will enable conversion of the reactors which cannot be converted today, ensure better efficiency and performance for all research reactors, and allow the design of more powerful new advanced LEU reactors. As in the past, the success of the RERTR program will depend on free exchange of ideas and information, and on the international friendship and cooperation that have been a trademark of the RERTR program since its inception.« less

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

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

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

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

Technical support to the Nuclear Regulatory Commission for the boiling water reactor blowdown heat transfer program

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

Rice, R.E.

Results are presented of studies conducted by Aerojet Nuclear Company (ANC) in FY 1975 to support the Nuclear Regulatory Commission (NRC) on the boiling water reactor blowdown heat transfer (BWR-BDHT) program. The support provided by ANC is that of an independent assessor of the program to ensure that the data obtained are adequate for verification of analytical models used for predicting reactor response to a postulated loss-of-coolant accident. The support included reviews of program plans, objectives, measurements, and actual data. Additional activity included analysis of experimental system performance and evaluation of the RELAP4 computer code as applied to the experiments.

Nuclear Engine System Simulation (NESS) version 2.0

NASA Technical Reports Server (NTRS)

Pelaccio, Dennis G.; Scheil, Christine M.; Petrosky, Lyman J.

1993-01-01

The topics are presented in viewgraph form and include the following; nuclear thermal propulsion (NTP) engine system analysis program development; nuclear thermal propulsion engine analysis capability requirements; team resources used to support NESS development; expanded liquid engine simulations (ELES) computer model; ELES verification examples; NESS program development evolution; past NTP ELES analysis code modifications and verifications; general NTP engine system features modeled by NESS; representative NTP expander, gas generator, and bleed engine system cycles modeled by NESS; NESS program overview; NESS program flow logic; enabler (NERVA type) nuclear thermal rocket engine; prismatic fuel elements and supports; reactor fuel and support element parameters; reactor parameters as a function of thrust level; internal shield sizing; and reactor thermal model.

10 CFR 1.43 - Office of Nuclear Reactor Regulation.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 1 2014-01-01 2014-01-01 false Office of Nuclear Reactor Regulation. 1.43 Section 1.43 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION Headquarters Program Offices § 1.43 Office of Nuclear Reactor Regulation. The Office of Nuclear Reactor Regulation— (a...

10 CFR 1.43 - Office of Nuclear Reactor Regulation.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 1 2012-01-01 2012-01-01 false Office of Nuclear Reactor Regulation. 1.43 Section 1.43 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION Headquarters Program Offices § 1.43 Office of Nuclear Reactor Regulation. The Office of Nuclear Reactor Regulation— (a...

10 CFR 1.43 - Office of Nuclear Reactor Regulation.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 1 2013-01-01 2013-01-01 false Office of Nuclear Reactor Regulation. 1.43 Section 1.43 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION Headquarters Program Offices § 1.43 Office of Nuclear Reactor Regulation. The Office of Nuclear Reactor Regulation— (a...

10 CFR 1.43 - Office of Nuclear Reactor Regulation.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Office of Nuclear Reactor Regulation. 1.43 Section 1.43 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION Headquarters Program Offices § 1.43 Office of Nuclear Reactor Regulation. The Office of Nuclear Reactor Regulation— (a...

10 CFR 1.43 - Office of Nuclear Reactor Regulation.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 1 2011-01-01 2011-01-01 false Office of Nuclear Reactor Regulation. 1.43 Section 1.43 Energy NUCLEAR REGULATORY COMMISSION STATEMENT OF ORGANIZATION AND GENERAL INFORMATION Headquarters Program Offices § 1.43 Office of Nuclear Reactor Regulation. The Office of Nuclear Reactor Regulation— (a...

Function of university reactors in operator licensing training for nuclear utilities

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

Wicks, F.

1985-11-01

The director of the Division of the US Nuclear Regulatory Commission in generic letter 84-10, dated April 26, 1984, spoke the requirement that applicants for senior reactor operator licenses for power reactors shall have performed then reactor startups. Simulator startups were not acknowledged. Startups performed on a university reactor are acceptable. The content and results of a five-day program combining instruction and experiments with the Rensselaer reactor are summarized.

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

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

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

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

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

Stillman, J. A.; Feldman, E. E.; Wilson, E. H.

This report contains the results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL), the MURR Facility, and the Nuclear Engineering Program – College of Engineering, University of Missouri-Columbia. The core conversion to LEU is being performed with financial support from the U. S. government. This report contains themore » results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL), the MURR Facility, and the Nuclear Engineering Program – College of Engineering, University of Missouri-Columbia. The core conversion to LEU is being performed with financial support from the U. S. government. In the framework of non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context most research and test reactors, both domestic and international, have started a program of conversion to the use of LEU fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (U-Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MURR. This report presents the results of a study of core behavior under a set of accident conditions for MURR cores fueled with HEU U-Alx dispersion fuel or LEU monolithic U-Mo alloy fuel with 10 wt% Mo (U-10Mo).« less

REACTOR PHYSICS CONSTANTS

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

None

1963-07-01

This second edition is based on data available on March 15, 1961. Sections on constants necessary for the interpretation of experimental data and on digital computer programs for reactor design and reactor physics have been added. 1344 references. (D.C.W.)

Policies and practices pertaining to the selection, qualification requirements, and training programs for nuclear-reactor operating personnel at the Oak Ridge National Laboratory

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

Culbert, W.H.

1985-10-01

This document describes the policies and practices of the Oak Ridge National Laboratory (ORNL) regarding the selection of and training requirements for reactor operating personnel at the Laboratory's nuclear-reactor facilities. The training programs, both for initial certification and for requalification, are described and provide the guidelines for ensuring that ORNL's research reactors are operated in a safe and reliable manner by qualified personnel. This document gives an overview of the reactor facilities and addresses the various qualifications, training, testing, and requalification requirements stipulated in DOE Order 5480.1A, Chapter VI (Safety of DOE-Owned Reactors); it is intended to be in compliancemore » with this DOE Order, as applicable to ORNL facilities. Included also are examples of the documentation maintained amenable for audit.« less

Neutrino Physics with Nuclear Reactors: An Overview

NASA Astrophysics Data System (ADS)

Ochoa-Ricoux, J. P.

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

R and D program for French sodium fast reactor: On the description and detection of sodium boiling phenomena during sub-assembly blockages

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

Vanderhaegen, M.; Laboratory of Waves and Acoustic, Institut Langevin, ESPCI ParisTech, 10 rue Vauquelin, 75005 Paris; Paumel, K.

2011-07-01

In support of the French ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) reactor program, which aims to demonstrate the industrial applicability of sodium fast reactors with an increased level of safety demonstration and availability compared to the past French sodium fast reactors, emphasis is placed on reactor instrumentation. It is in this framework that CEA studies continuous core monitoring to detect as early as possible the onset of sodium boiling. Such a detection system is of particular interest due to the rapid progress and the consequences of a Total Instantaneous Blockage (TIB) at a subassembly inlet, where sodium boilingmore » intervenes in an early phase. In this paper, the authors describe all the particularities which intervene during the different boiling stages and explore possibilities for their detection. (authors)« less

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

Mulder, R.U.; Benneche, P.E.; Hosticka, B.

The objective of the DOE supported Reactor Sharing Program is to increase the availability of university nuclear reactor facilities to non-reactor-owning educational institutions. The educational and research programs of these user institutions is enhanced by the use of the nuclear facilities. Several methods have been used by the UVA Reactor Facility to achieve this objective. First, many college and secondary school groups toured the Reactor Facility and viewed the UVAR reactor and associated experimental facilities. Second, advanced undergraduate and graduate classes from area colleges and universities visited the facility to perform experiments in nuclear engineering and physics which would notmore » be possible at the user institution. Third, irradiation and analysis services at the Facility have been made available for research by faculty and students from user institutions. Fourth, some institutions have received activated material from UVA from use at their institutions. These areas are discussed in this report.« less

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

NONE

The objective of the DOE supported Reactor Sharing Program is to increase the availability of university nuclear reactor facilities to non-reactor-owning educational institutions. The educational and research programs of these user institutions is enhanced by the use of the nuclear facilities. Several methods have been used by the UVA Reactor Facility to achieve this objective. First, many college and secondary school groups toured the Reactor Facility and viewed the UVAR reactor and associated experimental facilities. Second, advanced undergraduate and graduate classes from area colleges and universities visited the facility to perform experiments in nuclear engineering and physics which would notmore » be possible at the user institution. Third, irradiation and analysis services at the Facility have been made available for research by faculty and students from user institutions. Fourth, some institutions have received activated material from UVA for use at their institutions. These areas are discussed further in the report.« less

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

Mulder, R.U.; Benneche, P.E.; Hosticka, B.

The objective of the DOE supported Reactor Sharing Program is to increase the availability of university nuclear reactor facilities to non-reactor-owning educational institutions. The educational and research programs of these user institutions is enhanced by the use of the nuclear facilities. Several methods have been used by the UVA Reactor Facility to achieve this objective. First, many college and secondary school groups toured the Reactor Facility and viewed the UVAR reactor and associated experimental facilities. Second, advanced undergraduate and graduate classes from area colleges and universities visited the facility to perform experiments in nuclear engineering and physics which would notmore » be possible at the user institution. Third, irradiation and analysis services at the Facility have been made available for research by faculty and students from user institutions. Fourth, some institutions have received activated material from UVA for use at their institutions. These areas are discussed here.« less

Materials technology for an advanced space power nuclear reactor concept: Program summary

NASA Technical Reports Server (NTRS)

Gluyas, R. E.; Watson, G. K.

1975-01-01

The results of a materials technology program for a long-life (50,000 hr), high-temperature (950 C coolant outlet), lithium-cooled, nuclear space power reactor concept are reviewed and discussed. Fabrication methods and compatibility and property data were developed for candidate materials for fuel pins and, to a lesser extent, for potential control systems, reflectors, reactor vessel and piping, and other reactor structural materials. The effects of selected materials variables on fuel pin irradiation performance were determined. The most promising materials for fuel pins were found to be 85 percent dense uranium mononitride (UN) fuel clad with tungsten-lined T-111 (Ta-8W-2Hf).

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

NASA Technical Reports Server (NTRS)

Weinstein, H.; Lavan, Z.

1975-01-01

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

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

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

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

2014-01-01

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

Development of advanced strain diagnostic techniques for reactor environments.

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

Fleming, Darryn D.; Holschuh, Thomas Vernon,; Miller, Timothy J.

2013-02-01

The following research is operated as a Laboratory Directed Research and Development (LDRD) initiative at Sandia National Laboratories. The long-term goals of the program include sophisticated diagnostics of advanced fuels testing for nuclear reactors for the Department of Energy (DOE) Gen IV program, with the future capability to provide real-time measurement of strain in fuel rod cladding during operation in situ at any research or power reactor in the United States. By quantifying the stress and strain in fuel rods, it is possible to significantly improve fuel rod design, and consequently, to improve the performance and lifetime of the cladding.more » During the past year of this program, two sets of experiments were performed: small-scale tests to ensure reliability of the gages, and reactor pulse experiments involving the most viable samples in the Annulated Core Research Reactor (ACRR), located onsite at Sandia. Strain measurement techniques that can provide useful data in the extreme environment of a nuclear reactor core are needed to characterize nuclear fuel rods. This report documents the progression of solutions to this issue that were explored for feasibility in FY12 at Sandia National Laboratories, Albuquerque, NM.« less

A Neutronic Program for Critical and Nonequilibrium Study of Mobile Fuel Reactors: The Cinsf1D Code

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

Lecarpentier, David; Carpentier, Vincent

2003-01-15

Molten salt reactors (MSRs) have the distinction of having a liquid fuel that is also the coolant. The transport of delayed-neutron precursors by the fuel modifies the precursors' equation. As a consequence, it is necessary to adapt the methods currently used for solid fuel reactors to achieve critical or kinetics calculations for an MSR. A program is presented for which this adaptation has been carried out within the framework of the two-energy-group diffusion theory with one dimension of space. This program has been called Cinsf1D (Cinetique pour reacteur a sels fondus 1D)

Tory II-A: a nuclear ramjet test reactor

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

Hadley, J.W.

Declassified 28 Nov 1973. The first test reactor in the Pluto program, leading to development of a nuclear ramjet engine, is called Tory II-A. While it is not an actual prototype engine, this reactor embodies a core design which is considered feasible for an engine, and operation of the reactor will provide a test of that core type as well as more generalized values in reactor design and testing. The design of Tory II-A and construction of the reactor and of its test facility are described. Operation of the Tory II-A core at a total power of 160 megawatts, withmore » 800 pounds of air per second passing through the core and emerging at a temperature of 2000 deg F, is the central objective of the test program. All other reactor and facility components exist to support operation of the core, and preliminary steps in the test program itself will be directed primarily toward ensuring attalnment of full-power operation and collection of meaningful data on core behavior during that operation. The core, 3 feet in diameter and 41/2 feet long, will be composed of bundled ceramic tubes whose central holes will provide continuous air passages from end to end of the reactor. These tubes are to be composed of a homogeneous mixture of UO/sub 2/ fuel and BeO moderator, compacted and sintered to achieve high strength and density. (30 references) (auth)« less

Determination of the Arrhenius Activation Energy Using a Temperature-Programmed Flow Reactor.

ERIC Educational Resources Information Center

Chan, Kit-ha C.; Tse, R. S.

1984-01-01

Describes a novel method for the determination of the Arrhenius activation energy, without prejudging the validity of the Arrhenius equation or the concept of activation energy. The method involves use of a temperature-programed flow reactor connected to a concentration detector. (JN)

Nuclear Technology Series. Nuclear Reactor (Plant) Operator Trainee. A Suggested Program Planning Guide. Revised June 80.

ERIC Educational Resources Information Center

Center for Occupational Research and Development, Inc., Waco, TX.

This program planning guide for a two-year postsecondary nuclear reactor (plant) operator trainee program is designed for use with courses 1-16 of thirty-five in the Nuclear Technology Series. The purpose of the guide is to describe the nuclear power field and its job categories for specialists, technicians and operators; and to assist planners,…

Review Of Piping And Pressure Vessel Code Design Criteria. Technical Report 217.

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

None, None

1969-04-18

This Technical Report summarizes a review of the design philosophies and criteria of the ASME Boiler and Pressure Vessel Code and the USASI Code for Pressure Piping. It traces the history of the Codes since their inception and critically reviews their present status. Recommendations are made concerning the applicability of the Codes to the special needs of LMFBR liquid sodium piping.

Three region steam drum model for a nuclear power plant simulator (BRENDA). Technical report 1 Oct 80-May 81. [LMFBR

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

Slovik, G.C.

1981-08-01

A new three region steam drum model has been developed. This model differs from previous works for it assumes the existence of three regions within the steam drum: a steam region, a mid region (assumed to be under saturation conditions at steady state), and a bottom region (having a mixed mean subcooled enthalpy).

KiloPower Program

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

McClure, Patrick Ray

2016-08-04

These are the slides for a phone interview with Aerospace America magazine of the AIAA. It goes over the KiloPower Program at Los Alamos National Laboratory (LANL), and covers the following: 1 kWe Kilopower, 10 kWe Kilopower, Kilopower Reactor Using Stirling Technology (KRUSTY) Integration Test (DAF), Reactor Configuration, and Platen Positions.

Development of a three-dimensional core dynamics analysis program for commercial boiling water reactors

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

Bessho, Yasunori; Yokomizo, Osamu; Yoshimoto, Yuichiro

1997-03-01

Development and qualification results are described for a three-dimensional, time-domain core dynamics analysis program for commercial boiling water reactors (BWRs). The program allows analysis of the reactor core with a detailed mesh division, which eliminates calculational ambiguity in the nuclear-thermal-hydraulic stability analysis caused by reactor core regional division. During development, emphasis was placed on high calculational speed and large memory size as attained by the latest supercomputer technology. The program consists of six major modules, namely a core neutronics module, a fuel heat conduction/transfer module, a fuel channel thermal-hydraulic module, an upper plenum/separator module, a feedwater/recirculation flow module, and amore » control system module. Its core neutronics module is based on the modified one-group neutron kinetics equation with the prompt jump approximation and with six delayed neutron precursor groups. The module is used to analyze one fuel bundle of the reactor core with one mesh (region). The fuel heat conduction/transfer module solves the one-dimensional heat conduction equation in the radial direction with ten nodes in the fuel pin. The fuel channel thermal-hydraulic module is based on separated three-equation, two-phase flow equations with the drift flux correlation, and it analyzes one fuel bundle of the reactor core with one channel to evaluate flow redistribution between channels precisely. Thermal margin is evaluated by using the GEXL correlation, for example, in the module.« less

Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program

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

Smith, Cyrus M; Nanstad, Randy K; Clayton, Dwight A

2012-09-01

The Department of Energy s (DOE) Light Water Reactor Sustainability (LWRS) Program is a five year effort which works to develop the fundamental scientific basis to understand, predict, and measure changes in materials and systems, structure, and components as they age in environments associated with continued long-term operations of existing commercial nuclear power reactors. This year, the Materials Aging and Degradation (MAaD) Pathway of this program has placed emphasis on emerging Non-Destructive Evaluation (NDE) methods which support these objectives. DOE funded Research and Development (R&D) on emerging NDE techniques to support commercial nuclear reactor sustainability is expected to begin nextmore » year. This summer, the MAaD Pathway invited subject matter experts to participate in a series of workshops which developed the basis for the research plan of these DOE R&D NDE activities. This document presents the results of one of these workshops which are the DOE LWRS NDE R&D Roadmap for Reactor Pressure Vessels (RPV). These workshops made a substantial effort to coordinate the DOE NDE R&D with that already underway or planned by the Electric Power Research Institute (EPRI) and the Nuclear Regulatory Commission (NRC) through their representation at these workshops.« less

Nuclear Thermal Propulsion: A Joint NASA/DOE/DOD Workshop

NASA Technical Reports Server (NTRS)

Clark, John S. (Editor)

1991-01-01

Papers presented at the joint NASA/DOE/DOD workshop on nuclear thermal propulsion are compiled. The following subject areas are covered: nuclear thermal propulsion programs; Rover/NERVA and NERVA systems; Low Pressure Nuclear Thermal Rocket (LPNTR); particle bed reactor nuclear rocket; hybrid propulsion systems; wire core reactor; pellet bed reactor; foil reactor; Droplet Core Nuclear Rocket (DCNR); open cycle gas core nuclear rockets; vapor core propulsion reactors; nuclear light bulb; Nuclear rocket using Indigenous Martian Fuel (NIMF); mission analysis; propulsion and reactor technology; development plans; and safety issues.

Aging management program of the reactor building concrete at Point Lepreau Generating Station

NASA Astrophysics Data System (ADS)

Aldea, C.-M.; Shenton, B.; Demerchant, M. M.; Gendron, T.

2011-04-01

In order for New Brunswick Power Nuclear (NBPN) to control the risks of degradation of the concrete reactor building at the Point Lepreau Generating Station (PLGS) the development of an aging management plan (AMP) was initiated. The intention of this plan was to determine the requirements for specific structural components of concrete of the reactor building that require regular inspection and maintenance to ensure the safe and reliable operation of the plant. The document is currently in draft form and presents an integrated methodology for the application of an AMP for the concrete of the reactor building. The current AMP addresses the reactor building structure and various components, such as joint sealant and liners that are integral to the structure. It does not include internal components housed within the structure. This paper provides background information regarding the document developed and the strategy developed to manage potential degradation of the concrete of the reactor building, as well as specific programs and preventive and corrective maintenance activities initiated.

Development of New Transportation/Storage Cask System for Use by DOE Russian Research Reactor Fuel Return Program

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

Michael Tyacke; Frantisek Svitak; Jiri Rychecky

2010-04-01

The United States, the Russian Federation, and the International Atomic Energy Agency (IAEA) have been working together on a program called the Russian Research Reactor Fuel Return (RRRFR) Program. The purpose of this program is to return Soviet or Russian supplied high-enriched uranium (HEU) fuel currently stored at Russian-designed research reactors throughout the world to Russia. To accommodate transport of the HEU spent nuclear fuel (SNF), a new large-capacity transport/storage cask system was specially designed for handling and operations under the unique conditions for these research reactor facilities. This new cask system is named the ŠKODA VPVR/M cask. The design,more » licensing, testing, and delivery of this new cask system are the results of a significant international cooperative effort by several countries and involved numerous private and governmental organizations. This paper contains the following sections: (1) Introduction/Background; (2) VPVR/M Cask Description; (3) Ancillary Equipment, (4) Cask Licensing; (5) Cask Demonstration and Operations; (6) IAEA Procurement, Quality Assurance Inspections, Fabrication, and Delivery; and, (7) Summary and Conclusions.« less

Development of a New Transportation/Storage Cask System for Use by the DOE Russian Research Reactor Fuel Return Program

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

Michael J. Tyacke; Frantisek Svitak; Jiri Rychecky

2007-10-01

The United States, the Russian Federation, and the International Atomic Energy Agency (IAEA) have been working together on a program called the Russian Research Reactor Fuel Return (RRRFR) Program. The purpose of this program is to return Soviet or Russian-supplied high-enriched uranium (HEU) fuel, currently stored at Russian-designed research reactors throughout the world, to Russia. To accommodate transport of the HEU spent nuclear fuel (SNF), a new large-capacity transport/storage cask system was specially designed for handling and operations under the unique conditions at these research reactor facilities. This new cask system is named the ŠKODA VPVR/M cask. The design, licensing,more » testing, and delivery of this new cask system result from a significant international cooperative effort by several countries and involved numerous private and governmental organizations. This paper contains the following sections: 1) Introduction; 2) VPVR/M Cask Description; 3) Ancillary Equipment, 4) Cask Licensing; 5) Cask Demonstration and Operations; 6) IAEA Procurement, Quality Assurance Inspections, Fabrication, and Delivery; and, 7) Conclusions.« less

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

NASA Astrophysics Data System (ADS)

Deswandri; Subekti, M.; Sunaryo, Geni Rina

2018-02-01

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

Improved Nuclear Reactor and Shield Mass Model for Space Applications

NASA Technical Reports Server (NTRS)

Robb, Kevin

2004-01-01

New technologies are being developed to explore the distant reaches of the solar system. Beyond Mars, solar energy is inadequate to power advanced scientific instruments. One technology that can meet the energy requirements is the space nuclear reactor. The nuclear reactor is used as a heat source for which a heat-to-electricity conversion system is needed. Examples of such conversion systems are the Brayton, Rankine, and Stirling cycles. Since launch cost is proportional to the amount of mass to lift, mass is always a concern in designing spacecraft. Estimations of system masses are an important part in determining the feasibility of a design. I worked under Michael Barrett in the Thermal Energy Conversion Branch of the Power & Electric Propulsion Division. An in-house Closed Cycle Engine Program (CCEP) is used for the design and performance analysis of closed-Brayton-cycle energy conversion systems for space applications. This program also calculates the system mass including the heat source. CCEP uses the subroutine RSMASS, which has been updated to RSMASS-D, to estimate the mass of the reactor. RSMASS was developed in 1986 at Sandia National Laboratories to quickly estimate the mass of multi-megawatt nuclear reactors for space applications. In response to an emphasis for lower power reactors, RSMASS-D was developed in 1997 and is based off of the SP-100 liquid metal cooled reactor. The subroutine calculates the mass of reactor components such as the safety systems, instrumentation and control, radiation shield, structure, reflector, and core. The major improvements in RSMASS-D are that it uses higher fidelity calculations, is easier to use, and automatically optimizes the systems mass. RSMASS-D is accurate within 15% of actual data while RSMASS is only accurate within 50%. My goal this summer was to learn FORTRAN 77 programming language and update the CCEP program with the RSMASS-D model.

Comparison of Analysis Results Between 2D/1D Synthesis and RAPTOR-M3G in the Korea Standard Nuclear Plant (KSNP)

NASA Astrophysics Data System (ADS)

Joung Lim, Mi; Maeng, Young Jae; Fero, Arnold H.; Anderson, Stanwood L.

2016-02-01

The 2D/1D synthesis methodology has been used to calculate the fast neutron (E > 1.0 MeV) exposure to the beltline region of the reactor pressure vessel. This method uses the DORT 3.1 discrete ordinates code and the BUGLE-96 cross-section library based on ENDF/B-VI. RAPTOR-M3G (RApid Parallel Transport Of Radiation-Multiple 3D Geometries) which performs full 3D calculations was developed and is based on domain decomposition algorithms, where the spatial and angular domains are allocated and processed on multi-processor computer architecture. As compared to traditional single-processor applications, this approach reduces the computational load as well as the memory requirement per processor. Both methods are applied to surveillance test results for the Korea Standard Nuclear Plant (KSNP)-OPR (Optimized Power Reactor) 1000 MW. The objective of this paper is to compare the results of the KSNP surveillance program between 2D/1D synthesis and RAPTOR-M3G. Each operating KSNP has a reactor vessel surveillance program consisting of six surveillance capsules located between the core and the reactor vessel in the downcomer region near the reactor vessel wall. In addition to the In-Vessel surveillance program, an Ex-Vessel Neutron Dosimetry (EVND) program has been implemented. In order to estimate surveillance test results, cycle-specific forward transport calculations were performed by 2D/1D synthesis and by RAPTOR-M3G. The ratio between measured and calculated (M/C) reaction rates will be discussed. The current plan is to install an EVND system in all of the Korea PWRs including the new reactor type, APR (Advanced Power Reactor) 1400 MW. This work will play an important role in establishing a KSNP-specific database of surveillance test results and will employ RAPTOR-M3G for surveillance dosimetry location as well as positions in the KSNP reactor vessel.

FALCON reactor-pumped laser description and program overview

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

None

1989-12-01

The FALCON (Fission Activated Laser CONcept) reactor-pumped laser program at Sandia National Laboratories is examining the feasibility of high-power systems pumped directly by the energy from a nuclear reactor. In this concept we use the highly energetic fission fragments from neutron induced fission to excite a large volume laser medium. This technology has the potential to scale to extremely large optical power outputs in a primarily self-powered device. A laser system of this type could also be relatively compact and capable of long run times without refueling.

A complete dosimetry experimental program in support to the core characterization and to the power calibration of the CABRI reactor. A complete dosimetry experimental program in support of the core characterization and of the power calibration of the CABRI reactor

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

Rodiac, F.; Hudelot, JP.; Lecerf, J.

CABRI is an experimental pulse reactor operated by CEA at the Cadarache research center. Since 1978 the experimental programs have aimed at studying the fuel behavior under Reactivity Initiated Accident (RIA) conditions. Since 2003, it has been refurbished in order to be able to provide RIA and LOCA (Loss Of Coolant Accident) experiments in prototypical PWR conditions (155 bar, 300 deg. C). This project is part of a broader scope including an overall facility refurbishment and a safety review. The global modification is conducted by the CEA project team. It is funded by IRSN, which is conducting the CIP experimentalmore » program, in the framework of the OECD/NEA project CIP. It is financed in the framework of an international collaboration. During the reactor restart, commissioning tests are realized for all equipment, systems and circuits of the reactor. In particular neutronics and power commissioning tests will be performed respectively in 2015 and 2016. This paper focuses on the design of a complete and original dosimetry program that was built in support to the CABRI core characterization and to the power calibration. Each one of the above experimental goals will be fully described, as well as the target uncertainties and the forecasted experimental techniques and data treatment. (authors)« less

A Special Topic From Nuclear Reactor Dynamics for the Undergraduate Physics Curriculum

ERIC Educational Resources Information Center

Sevenich, R. A.

1977-01-01

Presents an intuitive derivation of the point reactor equations followed by formulation of equations for inverse and direct kinetics which are readily programmed on a digital computer. Suggests several computer simulations involving the effect of control rod motion on reactor power. (MLH)

Status of the US RERTR Program

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

Travelli, A.

1995-02-01

The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. The major events, findings, and activities of 1994 are reviewed after a brief summary of the results which the RERTR Program had achieved by the end of 1993 in collaboration with its many international partners. The RERTR Program has moved aggressively to support President Clinton`s nonproliferation policy and his goal {open_quotes}to minimize the use of highly-enriched uranium in civil nuclear programs{close_quotes}. An Environmental Assessment which addresses the urgent-relief acceptance of 409 spent fuel elements was completed, and the first shipment of spent fuel elements is scheduledmore » for this month. An Environmental Impact Statement addressing the acceptance of spent research reactor fuel containing enriched uranium of U.S. origin is scheduled for completion by the end of June 1995. The U.S. administration has decided to resume development of high-density LEU research reactor fuels. DOE funding and guidance are expected to begin soon. A preliminary plan for the resumption of fuel development has been prepared and is ready for implementation. The scope and main technical activities of a plan to develop and demonstrate within the next five years the technical means needed to convert Russian-supplied research reactors to LEU fuels was agreed upon by the RERTR Program and four Russian institutes lead by RDIPE. Both Secretary O`Leary and Minister Michailov have expressed strong support for this initiative. Joint studies have made significant progress, especially in assessing the technical and economic feasibility of using reduced enrichment fuels in the SAFARI-I reactor in South Africa and in the Advanced Neutron Source reactor under design at ORNL. Significant progress was achieved on several aspects of producing {sup 99}Mo from fission targets utilizing LEU instead of HEU to the achievement of the common goal.« less

Nuclear Energy Enabling Technologies (NEET) Reactor Materials: News for the Reactor Materials Crosscut, May 2016

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

Maloy, Stuart Andrew

In this newsletter for Nuclear Energy Enabling Technologies (NEET) Reactor Materials, pages 1-3 cover highlights from the DOE-NE (Nuclear Energy) programs, pages 4-6 cover determining the stress-strain response of ion-irradiated metallic materials via spherical nanoindentation, and pages 7-8 cover theoretical approaches to understanding long-term materials behavior in light water reactors.

77 FR 55877 - Initial Test Program of Condensate and Feedwater Systems for Light-Water Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2012-09-11

...-492- 3668; email: [email protected] . NRC's Agencywide Documents Access and Management System... Systems for Light-Water Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Regulatory guide; issuance... Systems for Boiling Water Reactor Power Plants.'' This regulatory guide is being revised to: (1) Expand...

Status of JUPITER Program

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

Inoue, T.; Shirakata, K.; Kinjo, K.

To obtain the data necessary for evaluating the nuclear design method of a large-scale fast breeder reactor, criticality tests with a large- scale homogeneous reactor were conducted as part of a joint research program by Japan and the U.S. Analyses of the tests are underway in both countries. The purpose of this paper is to describe the status of this project.

Development and experimental qualification of a calculation scheme for the evaluation of gamma heating in experimental reactors. Application to MARIA and Jules Horowitz (JHR) MTR Reactors

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

Tarchalski, M.; Pytel, K.; Wroblewska, M.

2015-07-01

Precise computational determination of nuclear heating which consists predominantly of gamma heating (more than 80 %) is one of the challenges in material testing reactor exploitation. Due to sophisticated construction and conditions of experimental programs planned in JHR it became essential to use most accurate and precise gamma heating model. Before the JHR starts to operate, gamma heating evaluation methods need to be developed and qualified in other experimental reactor facilities. This is done inter alia using OSIRIS, MINERVE or EOLE research reactors in France. Furthermore, MARIA - Polish material testing reactor - has been chosen to contribute to themore » qualification of gamma heating calculation schemes/tools. This reactor has some characteristics close to those of JHR (beryllium usage, fuel element geometry). To evaluate gamma heating in JHR and MARIA reactors, both simulation tools and experimental program have been developed and performed. For gamma heating simulation, new calculation scheme and gamma heating model of MARIA have been carried out using TRIPOLI4 and APOLLO2 codes. Calculation outcome has been verified by comparison to experimental measurements in MARIA reactor. To have more precise calculation results, model of MARIA in TRIPOLI4 has been made using the whole geometry of the core. This has been done for the first time in the history of MARIA reactor and was complex due to cut cone shape of all its elements. Material composition of burnt fuel elements has been implemented from APOLLO2 calculations. An experiment for nuclear heating measurements and calculation verification has been done in September 2014. This involved neutron, photon and nuclear heating measurements at selected locations in MARIA reactor using in particular Rh SPND, Ag SPND, Ionization Chamber (all three from CEA), KAROLINA calorimeter (NCBJ) and Gamma Thermometer (CEA/SCK CEN). Measurements were done in forty points using four channels. Maximal nuclear heating evaluated from measurements is of the order of 2.5 W/g at half of the possible MARIA power - 15 MW. The approach and the detailed program for experimental verification of calculations will be presented. The following points will be discussed: - Development of a gamma heating model of MARIA reactor with TRIPOLI 4 (coupled neutron-photon mode) and APOLLO2 model taking into account the key parameters like: configuration of the core, experimental loading, control rod location, reactor power, fuel depletion); - Design of specific measurement tools for MARIA experiments including for instance a new single-cell calorimeter called KAROLINA calorimeter; - MARIA experimental program description and a preliminary analysis of results; - Comparison of calculations for JHR and MARIA cores with experimental verification analysis, calculation behavior and n-γ 'environments'. (authors)« less

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

Cappiello, M.; Hobbins, R.; Penny, K.

As part of the Department of Energy Advanced Fuel Cycle program, a series of fuels development irradiation tests have been performed in the Advanced Test Reactor (ATR) at the Idaho National Laboratory. These tests are providing excellent data for advanced fuels development. The program is focused on the transmutation of higher actinides which best can be accomplished in a sodium-cooled fast reactor. Because a fast test reactor is no longer available in the US, a special test vehicle is used to achieve near-prototypic fast reactor conditions (neutron spectra and temperature) for use in ATR (a water-cooled thermal reactor). As partmore » of the testing program, there were many successful tests of advanced fuels including metals and ceramics. Recently however, there have been three experimental campaigns using metal fuels that experienced failure during irradiation. At the request of the program, an independent review committee was convened to review the post-test analyses performed by the fuels development team, to assess the conclusions of the team for the cause of the failures, to assess the adequacy and completeness of the analyses, to identify issues that were missed, and to make recommendations for improvements in the design and operation of future tests. Although there is some difference of opinion, the review committee largely agreed with the conclusions of the fuel development team regarding the cause of the failures. For the most part, the analyses that support the conclusions are sufficient.« less

SP-100 program developments

NASA Technical Reports Server (NTRS)

Schnyer, A. D.; Sholtis, J. A., Jr.; Wahlquist, E. J.; Verga, R. L.; Wiley, R. L.

1985-01-01

An update is provided on the status of the Sp-100 Space Reactor Power Program. The historical background that led to the program is reviewed and the overall program objectives and development approach are discussed. The results of the mission studies identify applications for which space nuclear power is desirable and even essential. Results of a series of technology feasibility experiments are expected to significantly improve the earlier technology data base for engineering development. The conclusion is reached that a nuclear reactor space power system can be developed by the early 1990s to meet emerging mission performance requirements.

Quality Assurance Program Plan for SFR Metallic Fuel Data Qualification

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

Benoit, Timothy; Hlotke, John Daniel; Yacout, Abdellatif

2017-07-05

This document contains an evaluation of the applicability of the current Quality Assurance Standards from the American Society of Mechanical Engineers Standard NQA-1 (NQA-1) criteria and identifies and describes the quality assurance process(es) by which attributes of historical, analytical, and other data associated with sodium-cooled fast reactor [SFR] metallic fuel and/or related reactor fuel designs and constituency will be evaluated. This process is being instituted to facilitate validation of data to the extent that such data may be used to support future licensing efforts associated with advanced reactor designs. The initial data to be evaluated under this program were generatedmore » during the US Integral Fast Reactor program between 1984-1994, where the data includes, but is not limited to, research and development data and associated documents, test plans and associated protocols, operations and test data, technical reports, and information associated with past United States Nuclear Regulatory Commission reviews of SFR designs.« less

Interim MELCOR Simulation of the Fukushima Daiichi Unit 2 Accident Reactor Core Isolation Cooling Operation

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

Ross, Kyle W.; Gauntt, Randall O.; Cardoni, Jeffrey N.

2013-11-01

Data, a brief description of key boundary conditions, and results of Sandia National Laboratories’ ongoing MELCOR analysis of the Fukushima Unit 2 accident are given for the reactor core isolation cooling (RCIC) system. Important assumptions and related boundary conditions in the current analysis additional to or different than what was assumed/imposed in the work of SAND2012-6173 are identified. This work is for the U.S. Department of Energy’s Nuclear Energy University Programs fiscal year 2014 Reactor Safety Technologies Research and Development Program RC-7: RCIC Performance under Severe Accident Conditions.

Safe Affordable Fission Engine-(SAFE-) 100a Heat Exchanger Thermal and Structural Analysis

NASA Technical Reports Server (NTRS)

Steeve, B. E.

2005-01-01

A potential fission power system for in-space missions is a heat pipe-cooled reactor coupled to a Brayton cycle. In this system, a heat exchanger (HX) transfers the heat of the reactor core to the Brayton gas. The Safe Affordable Fission Engine- (SAFE-) 100a is a test program designed to thermally and hydraulically simulate a 95 Btu/s prototypic heat pipe-cooled reactor using electrical resistance heaters on the ground. This Technical Memorandum documents the thermal and structural assessment of the HX used in the SAFE-100a program.

Highly Selective Nuclide Removal from the R-Reactor Disassembly Basin at the SRS

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

Pickett, J. B.; Austin, W. E.; Dukes, H. H.

This paper describes the results of a deployment of highly selective ion-exchange resin technologies for the in-situ removal of Cs-137 and Sr-90 from the Savannah River Site (SRS) R-Reactor Disassembly Basin. The deployment was supported by the DOE Office of Science and Technology's (OST, EM-50) National Engineering Technology Laboratory (NETL), as a part of an Accelerated Site Technology Deployment (ASTD) project. The Facilities Decontamination and Decommissioning (FDD) Program at the SRS conducted this deployment as a part of an overall program to deactivate three of the site's five reactor disassembly basins.

Highly Selective Nuclide Removal from the R-Reactor Disassembly Basin at SRS

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

Pickett, J.B.

This paper describes the results of a deployment of highly selective ion-exchange resin technologies for the in-situ removal of Cs-137 and Sr-90 from the Savannah River Site (SRS) R-Reactor Disassembly Basin. The deployment was supported by the DOE Office of Science and Technology's (OST, EM-50) National Engineering Technology Laboratory (NETL), as a part of an Accelerated Site Technology Deployment (ASTD) project. The Facilities Decontamination and Decommissioning (FDD) Program at the SRS conducted this deployment as a part of an overall program to deactivate three of the site's five reactor disassembly basins

Status and progress of the RERTR program in the year 2002.

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

Travelli, A.; Technology Development

2003-01-01

Following the cancellation of the 2001 International RERTR Meeting, which had been planned to occur in Bali, Indonesia, this paper describes the progress achieved by the Reduced Enrichment for Research and Test Reactors (RERTR) Program in collaboration with its many international partners during the years 2001 and 2002, and discusses the main activities planned for the year 2003. The past two years have been characterized by very important achievements of the RERTR program, but these technical achievements have been overshadowed by the terrible events of September 11, 2001. Those events have caused the U.S. Government to reevaluate the importance andmore » urgency of the RERTR program goals. A recommendation made at the highest levels of the government calls for an immediate acceleration of the program activities, with the goal of converting all the world's research reactors to low-enriched fuel at the earliest possible time, and including both Soviet-designed and United States-designed research reactors.« less

Flow Induced Vibration Program at Argonne National Laboratory

NASA Astrophysics Data System (ADS)

1984-01-01

The Argonne National Laboratory's Flow Induced Vibration Program, currently residing in the Laboratory's Components Technology Division is discussed. Throughout its existence, the overall objective of the program was to develop and apply new and/or improved methods of analysis and testing for the design evaluation of nuclear reactor plant components and heat exchange equipment from the standpoint of flow induced vibration. Historically, the majority of the program activities were funded by the US Atomic Energy Commission, the Energy Research and Development Administration, and the Department of Energy. Current DOE funding is from the Breeder Mechanical Component Development Division, Office of Breeder Technology Projects; Energy Conversion and Utilization Technology Program, Office of Energy Systems Research; and Division of Engineering, Mathematical and Geosciences, office of Basic Energy Sciences. Testing of Clinch River Breeder Reactor upper plenum components was funded by the Clinch River Breeder Reactor Plant Project Office. Work was also performed under contract with Foster Wheeler, General Electric, Duke Power Company, US Nuclear Regulatory Commission, and Westinghouse.

On use of ZPR research reactors and associated instrumentation and measurement methods for reactor physics studies

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

Chauvin, J.P.; Blaise, P.; Lyoussi, A.

2015-07-01

The French atomic and alternative energies -CEA- is strongly involved in research and development programs concerning the use of nuclear energy as a clean and reliable source of energy and consequently is working on the present and future generation of reactors on various topics such as ageing plant management, optimization of the plutonium stockpile, waste management and innovative systems exploration. Core physics studies are an essential part of this comprehensive R and D effort. In particular, the Zero Power Reactor (ZPR) of CEA: EOLE, MINERVE and MASURCA play an important role in the validation of neutron (as well photon) physicsmore » calculation tools (codes and nuclear data). The experimental programs defined in the CEA's ZPR facilities aim at improving the calculation routes by reducing the uncertainties of the experimental databases. They also provide accurate data on innovative systems in terms of new materials (moderating and decoupling materials) and new concepts (ADS, ABWR, new MTR (e.g. JHR), GENIV) involving new fuels, absorbers and coolant materials. Conducting such interesting experimental R and D programs is based on determining and measuring main parameters of phenomena of interest to qualify calculation tools and nuclear data 'libraries'. Determining these parameters relies on the use of numerous and different experimental techniques using specific and appropriate instrumentation and detection tools. Main ZPR experimental programs at CEA, their objectives and challenges will be presented and discussed. Future development and perspectives regarding ZPR reactors and associated programs will be also presented. (authors)« less

10 CFR 72.218 - Termination of licenses.

Code of Federal Regulations, 2011 CFR

2011-01-01

... NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General License for Storage of Spent Fuel at Power Reactor Sites § 72.218 Termination of licenses. (a) The notification regarding the program for the management of spent fuel at the reactor required by § 50.54(bb) of...

10 CFR 72.218 - Termination of licenses.

Code of Federal Regulations, 2010 CFR

2010-01-01

... NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General License for Storage of Spent Fuel at Power Reactor Sites § 72.218 Termination of licenses. (a) The notification regarding the program for the management of spent fuel at the reactor required by § 50.54(bb) of...

Early Program Development

NASA Image and Video Library

1963-01-01

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

Zirconium Hydride Space Power Reactor design.

NASA Technical Reports Server (NTRS)

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

1972-01-01

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

FY16 Status Report for the Uranium-Molybdenum Fuel Concept

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

Bennett, Wendy D.; Doherty, Ann L.; Henager, Charles H.

2016-09-22

The Fuel Cycle Research and Development program of the Office of Nuclear Energy has implemented a program to develop a Uranium-Molybdenum metal fuel for light water reactors. Uranium-Molybdenum fuel has the potential to provide superior performance based on its thermo-physical properties. With sufficient development, it may be able to provide the Light Water Reactor industry with a melt-resistant, accident-tolerant fuel with improved safety response. The Pacific Northwest National Laboratory has been tasked with extrusion development and performing ex-reactor corrosion testing to characterize the performance of Uranium-Molybdenum fuel in both these areas. This report documents the results of the fiscal yearmore » 2016 effort to develop the Uranium-Molybdenum metal fuel concept for light water reactors.« less

MSG test report-steady-state heat transfer. [LMFBR

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

Harty, R.B.

This report documents the results of the Steady-State Heat Transfer Tests conducted on the AI Modular Steam Generator (MSG), at the Sodium Component Test Installation (SCTI) of the Liquid Metal Engineering Center. Heat transfer and pressure drop performance data are given along with current predictions of performance. Departure from nucleate boiling characteristics is given. A dispersed flow film boiling model, employing thermal nonequilibrium, was used to analyze data in the film boiling region.

Evaluation of Nuclear Facility Decommissioning Projects program: a reference research reactor. Project summary report

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

Baumann, B.L.; Miller, R.L.

1983-10-01

This document presents, in summary form, generic conceptual information relevant to the decommissioning of a reference research reactor (RRR). All of the data presented were extracted from NUREG/CR-1756 and arranged in a form that will provide a basis for future comparison studies for the Evaluation of Nuclear Facility Decommissioning Projects (ENFDP) program.

Status of Wrought FeCrAl-UO 2 Capsules Irradiated in the Advanced Test Reactor

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

Field, Kevin G.; Harp, J.; Core, G.

2017-07-01

Candidate cladding materials for accident tolerant fuel applications require extensive testing and validation prior to commercial deployment within the nuclear power industry. One class of cladding materials, FeCrAl alloys, is currently undergoing such effort. Within these activities is a series of irradiation programs within the Advanced Test Reactor. These programs are developed to aid in commercial maturation and understand the fundamental mechanisms controlling the cladding performance during normal operation of a typical light water reactor. Three different irradiation programs are on-going; one designed as a simple proof-of-principle concept, the other to evaluate the susceptibility of FeCrAl to fuel-cladding chemical interaction,more » and the last to fully simulate the conditions of a pressurized water reactor experimentally. To date, nondestructive post-irradiation examination has been completed on the rodlet deemed FCA-L3 from the simple proof-of-concept irradiation program. Initial results show possible breach of the rodlet under irradiation but further studies are needed to conclusively determine whether breach has occurred and the underlying reasons for such a possible failure. Further work includes characterizing additional rodlets following irradiation.« less

International academic program in technologies of light-water nuclear reactors. Phases of development and implementation

NASA Astrophysics Data System (ADS)

Geraskin, N. I.; Glebov, V. B.

2017-01-01

The results of implementation of European educational projects CORONA and CORONA II dedicated to preserving and further developing nuclear knowledge and competencies in the area of technologies of light-water nuclear reactors are analyzed. Present article addresses issues of design and implementation of the program for specialized training in the branch of technologies of light-water nuclear reactors. The systematic approach has been used to construct the program for students of nuclear specialties, which corresponding to IAEA standards and commonly accepted nuclear principles recognized in the European Union. Possibilities of further development of the international cooperation between countries and educational institutions are analyzed. Special attention is paid to e-learning/distance training, nuclear knowledge preservation and interaction with European Nuclear Education Network.

INL Experimental Program Roadmap for Thermal Hydraulic Code Validation

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

Glenn McCreery; Hugh McIlroy

2007-09-01

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

Returning HEU Fuel from the Czech Republic to Russia

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

Michael Tyacke; Dr. Igor Bolshinsky

In December 1999, representatives from the United States, Russian Federation, and International Atomic Energy Agency began working on a program to return Russian supplied, highly enriched, uranium fuel stored at foreign research reactors to Russia. Now, under the Global Threat Reduction Initiative’s Russian Research Reactor Fuel Return Program, this effort has repatriated over 800 kg of highly enriched uranium to Russia from over 10 countries. In May 2004, the “Agreement Between the Government of the United States of America and the Government of the Russian Federation Concerning Cooperation for the Transfer of Russian Produced Research Reactor Nuclear Fuel to themore » Russian Federation” was signed. This agreement provides legal authority for the Russian Research Reactor Fuel Return Program and establishes parameters whereby eligible countries may return highly enriched uranium spent and fresh fuel assemblies and other fissile materials to Russia. On December 8, 2007, one of the largest shipments of highly enriched uranium spent nuclear fuel was successfully made from a Russian-designed nuclear research reactor in the Czech Republic to the Russian Federation. This accomplishment is the culmination of years of planning, negotiations, and hard work. The United States, Russian Federation, and the International Atomic Energy Agency have been working together. In February 2003, Russian Research Reactor Fuel Return Program representatives met with the Nuclear Research Institute in Rež, Czech Republic, and discussed the return of their highly enriched uranium spent nuclear fuel to the Russian Federation for reprocessing. Nearly 5 years later, the shipment was made. This article discusses the planning, preparations, coordination, and cooperation required to make this important international shipment.« less

Status and progress of the RERTR program in the year 2003.

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

Travelli, A.; Nuclear Engineering Division

2003-01-01

One of the most important events affecting the RERTR program during the past year was the decision by the U.S. Department of Energy to request the U.S. Congress to significantly increase RERTR program funding. This decision was prompted, at least in part, by the terrible events of September 11, 2001, and by a high-level U.S./Russian Joint Expert Group recommendation to immediately accelerate RERTR program activities in both countries, with the goal of converting all the world's research reactors to low-enriched fuel at the earliest possible time, and including both Soviet-designed and United States-designed research reactors. The U.S. Congress is expectedmore » to approve this request very soon, and the RERTR program has prepared itself well for the intense activities that the 'Accelerated RERTR Program' will require. Promising results have been obtained in the development of a fabrication process for monolithic LEU U-Mo fuel. Most existing and future research reactors could be converted to LEU with this fuel, which has a uranium density between 15.4 and 16.4 g/cm{sup 3} and yielded promising irradiation results in 2002. The most promising method hinges on producing the monolithic meat by cold-rolling a thin ingot produced by casting. The aluminum clad and the meat are bonded by friction stir welding and the cladding surface is finished by a light cold roll. This method can be applied to the production of miniplates and appears to be extendable to the production of full-size plates, possibly with intermediate anneals. Other methods planned for investigation include high temperature bonding and hot isostatic pressing. The progress achieved within the Russian RERTR program, both for the traditional tube-type elements and for the new 'universal' LEU U-Mo pin-type elements, promises to enable soon the conversion of many Russian-designed research and test reactors. Irradiation testing of both fuel types with LEU U-Mo dispersion fuels has begun. Detailed studies are in progress to define the feasibility of converting each Russian-designed research and test reactor to either fuel type. The plan for the Accelerated RERTR Program is structured to achieve LEU conversion of all HEU research reactors supplied by the United States and Russia during the next nine years. This effort will address, in addition to the fuel development and qualification, the analyses and performance/economic/safety evaluations needed to implement the conversions. In combination with this over-arching goal, the RERTR program plans to achieve at the earliest possible date qualification of LEU U-Mo dispersion fuels with uranium densities of 6 g/cm{sup 3} and 7 g/cm{sup 3}. Reactors currently using or planning to use LEU silicide fuel will rely on this fuel after termination of the FRRSNFA program, because it is acceptable to COGEMA for reprocessing. Qualification of LEU U-Mo dispersion fuels has suffered some unavoidable delays but, to accelerate it as much as possible, the RERTR program, the French CEA, and the Australian ANSTO have agreed to jointly pursue a two-element qualification test of LEU U-Mo dispersion fuel with uranium density of 7.0 g/cm{sup 3} to be performed in the Osiris reactor during 2004. The RERTR program also intends to eliminate all obstacles to the utilization of LEU in targets for isotope production, so that this important function can be performed without the need for weapons-grade materials. All of us, working together as we have for many years, can ensure that all these goals will be achieved. By promoting the efficiency and safety of research reactors while eliminating the traffic in weapons-grade uranium, we can prevent the possibility that some of this material might fall in the wrong hands. Few causes can be more deserving of our joint efforts.« less

Exploratory development of a glass ceramic automobile thermal reactor. [anti-pollution devices

NASA Technical Reports Server (NTRS)

Gould, R. E.; Petticrew, R. W.

1973-01-01

This report summarizes the design, fabrication and test results obtained for glass-ceramic (CER-VIT) automotive thermal reactors. Several reactor designs were evaluated using both engine-dynamometer and vehicle road tests. A maximum reactor life of about 330 hours was achieved in engine-dynamometer tests with peak gas temperatures of about 1065 C (1950 F). Reactor failures were mechanically induced. No evidence of chemical degradation was observed. It was concluded that to be useful for longer times, the CER-VIT parts would require a mounting system that was an improvement over those tested in this program. A reactor employing such a system was designed and fabricated.

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

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

D. Kokkinos

2005-04-28

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

Researcher Poses with a Nuclear Rocket Model

NASA Image and Video Library

1961-11-21

A researcher at the NASA Lewis Research Center with slide ruler poses with models of the earth and a nuclear-propelled rocket. The Nuclear Engine for Rocket Vehicle Applications (NERVA) was a joint NASA and Atomic Energy Commission (AEC) endeavor to develop a nuclear-powered rocket for both long-range missions to Mars and as a possible upper-stage for the Apollo Program. The early portion of the program consisted of basic reactor and fuel system research. This was followed by a series of Kiwi reactors built to test nuclear rocket principles in a non-flying nuclear engine. The next phase, NERVA, would create an entire flyable engine. The AEC was responsible for designing the nuclear reactor and overall engine. NASA Lewis was responsible for developing the liquid-hydrogen fuel system. The nuclear rocket model in this photograph includes a reactor at the far right with a hydrogen propellant tank and large radiator below. The payload or crew would be at the far left, distanced from the reactor.

HEDL FACILITIES CATALOG 400 AREA

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

MAYANCSIK BA

1987-03-01

The purpose of this project is to provide a sodium-cooled fast flux test reactor designed specifically for irradiation testing of fuels and materials and for long-term testing and evaluation of plant components and systems for the Liquid Metal Reactor (LMR) Program. The FFTF includes the reactor, heat removal equipment and structures, containment, core component handling and examination, instrumentation and control, and utilities and other essential services. The complex array of buildings and equipment are arranged around the Reactor Containment Building.

Light-Water-Reactor safety research program. Quarterly progress report, January--March 1977

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

None

The report summarizes the Argonne National Laboratory work performed during January, February, and March 1977 on water-reactor-safety problems. The following research and development areas are covered: (1) loss-of-coolant accident research: heat transfer and fluid dynamics; (2) transient fuel response and fission-product release program; (3) mechanical properties of zircaloy containing oxygen; and (4) steam-explosion studies.

N.S. SAVANNAH, DRAFT OF FINAL SAFEGUARDS REPORT TEST, START-UP AND TRIALS, NEW YORK SHIPBUILDING CORPORATION

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

None

1960-04-01

The N. S. Savannah program for testing, start-up, and initial operation of all reactor and propulsion components and systems is discussed. Definitions of test phases are given and various stages of the test program are outlined. A list of tests for the various reactor, propulsion, and other system components is included. (C.J.G.)

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

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

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

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

Testing piezoelectric sensors in a nuclear reactor environment

NASA Astrophysics Data System (ADS)

Reinhardt, Brian T.; Suprock, Andy; Tittmann, Bernhard

2017-02-01

Several Department of Energy Office of Nuclear Energy (DOE-NE) programs, such as the Fuel Cycle Research and Development (FCRD), Advanced Reactor Concepts (ARC), Light Water Reactor Sustainability, and Next Generation Nuclear Power Plants (NGNP), are investigating new fuels, materials, and inspection paradigms for advanced and existing reactors. A key objective of such programs is to understand the performance of these fuels and materials during irradiation. In DOE-NE's FCRD program, ultrasonic based technology was identified as a key approach that should be pursued to obtain the high-fidelity, high-accuracy data required to characterize the behavior and performance of new candidate fuels and structural materials during irradiation testing. The radiation, high temperatures, and pressure can limit the available tools and characterization methods. In this work piezoelectric transducers capable of making these measurements are developed. Specifically, three piezoelectric sensors (Bismuth Titanate, Aluminum Nitride, and Zinc Oxide) are tested in the Massachusetts Institute of Technology Research reactor to a fast neutron fluence of 8.65×1020 nf/cm2. It is demonstrated that Bismuth Titanate is capable of transduction up to 5 × 1020 nf/cm2, Zinc Oxide is capable of transduction up to at least 6.27 × 1020 nf/cm2, and Aluminum Nitride is capable of transduction up to at least 8.65 × 1020 nf/cm2.

Strengthening the fission reactor nuclear science and engineering program at UCLA. Final technical report

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

Okrent, D.

1997-06-23

This is the final report on DOE Award No. DE-FG03-92ER75838 A000, a three year matching grant program with Pacific Gas and Electric Company (PG and E) to support strengthening of the fission reactor nuclear science and engineering program at UCLA. The program began on September 30, 1992. The program has enabled UCLA to use its strong existing background to train students in technological problems which simultaneously are of interest to the industry and of specific interest to PG and E. The program included undergraduate scholarships, graduate traineeships and distinguished lecturers. Four topics were selected for research the first year, withmore » the benefit of active collaboration with personnel from PG and E. These topics remained the same during the second year of this program. During the third year, two topics ended with the departure o the students involved (reflux cooling in a PWR during a shutdown and erosion/corrosion of carbon steel piping). Two new topics (long-term risk and fuel relocation within the reactor vessel) were added; hence, the topics during the third year award were the following: reflux condensation and the effect of non-condensable gases; erosion/corrosion of carbon steel piping; use of artificial intelligence in severe accident diagnosis for PWRs (diagnosis of plant status during a PWR station blackout scenario); the influence on risk of organization and management quality; considerations of long term risk from the disposal of hazardous wastes; and a probabilistic treatment of fuel motion and fuel relocation within the reactor vessel during a severe core damage accident.« less

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

NASA Technical Reports Server (NTRS)

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

1973-01-01

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

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

NASA Astrophysics Data System (ADS)

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

2014-09-01

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

Nuclear Engine System Simulation (NESS). Volume 1: Program user's guide

NASA Astrophysics Data System (ADS)

Pelaccio, Dennis G.; Scheil, Christine M.; Petrosky, Lyman J.

1993-03-01

A Nuclear Thermal Propulsion (NTP) engine system design analysis tool is required to support current and future Space Exploration Initiative (SEI) propulsion and vehicle design studies. Currently available NTP engine design models are those developed during the NERVA program in the 1960's and early 1970's and are highly unique to that design or are modifications of current liquid propulsion system design models. To date, NTP engine-based liquid design models lack integrated design of key NTP engine design features in the areas of reactor, shielding, multi-propellant capability, and multi-redundant pump feed fuel systems. Additionally, since the SEI effort is in the initial development stage, a robust, verified NTP analysis design tool could be of great use to the community. This effort developed an NTP engine system design analysis program (tool), known as the Nuclear Engine System Simulation (NESS) program, to support ongoing and future engine system and stage design study efforts. In this effort, Science Applications International Corporation's (SAIC) NTP version of the Expanded Liquid Engine Simulation (ELES) program was modified extensively to include Westinghouse Electric Corporation's near-term solid-core reactor design model. The ELES program has extensive capability to conduct preliminary system design analysis of liquid rocket systems and vehicles. The program is modular in nature and is versatile in terms of modeling state-of-the-art component and system options as discussed. The Westinghouse reactor design model, which was integrated in the NESS program, is based on the near-term solid-core ENABLER NTP reactor design concept. This program is now capable of accurately modeling (characterizing) a complete near-term solid-core NTP engine system in great detail, for a number of design options, in an efficient manner. The following discussion summarizes the overall analysis methodology, key assumptions, and capabilities associated with the NESS presents an example problem, and compares the results to related NTP engine system designs. Initial installation instructions and program disks are in Volume 2 of the NESS Program User's Guide.

Nuclear Engine System Simulation (NESS). Volume 1: Program user's guide

NASA Technical Reports Server (NTRS)

Pelaccio, Dennis G.; Scheil, Christine M.; Petrosky, Lyman J.

1993-01-01

A Nuclear Thermal Propulsion (NTP) engine system design analysis tool is required to support current and future Space Exploration Initiative (SEI) propulsion and vehicle design studies. Currently available NTP engine design models are those developed during the NERVA program in the 1960's and early 1970's and are highly unique to that design or are modifications of current liquid propulsion system design models. To date, NTP engine-based liquid design models lack integrated design of key NTP engine design features in the areas of reactor, shielding, multi-propellant capability, and multi-redundant pump feed fuel systems. Additionally, since the SEI effort is in the initial development stage, a robust, verified NTP analysis design tool could be of great use to the community. This effort developed an NTP engine system design analysis program (tool), known as the Nuclear Engine System Simulation (NESS) program, to support ongoing and future engine system and stage design study efforts. In this effort, Science Applications International Corporation's (SAIC) NTP version of the Expanded Liquid Engine Simulation (ELES) program was modified extensively to include Westinghouse Electric Corporation's near-term solid-core reactor design model. The ELES program has extensive capability to conduct preliminary system design analysis of liquid rocket systems and vehicles. The program is modular in nature and is versatile in terms of modeling state-of-the-art component and system options as discussed. The Westinghouse reactor design model, which was integrated in the NESS program, is based on the near-term solid-core ENABLER NTP reactor design concept. This program is now capable of accurately modeling (characterizing) a complete near-term solid-core NTP engine system in great detail, for a number of design options, in an efficient manner. The following discussion summarizes the overall analysis methodology, key assumptions, and capabilities associated with the NESS presents an example problem, and compares the results to related NTP engine system designs. Initial installation instructions and program disks are in Volume 2 of the NESS Program User's Guide.

Reactor Statics Module, RS-9: Multigroup Diffusion Program Using an Exponential Acceleration Technique.

ERIC Educational Resources Information Center

Macek, Victor C.

The nine Reactor Statics Modules are designed to introduce students to the use of numerical methods and digital computers for calculation of neutron flux distributions in space and energy which are needed to calculate criticality, power distribution, and fuel burnup for both slow neutron and fast neutron fission reactors. The last module, RS-9,…

Hanford Laboratories monthly activities report, March 1964

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

Not Available

1964-04-15

The monthly report for the Hanford Laboratories Operation, March 1964. Reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, biology operation, and physics and instrumentation research, and applied mathematics operation, and programming operations are discussed.

CERCA LEU fuel assemblies testing in Maria Reactor - safety analysis summary and testing program scope.

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

Pytel, K.; Mieleszczenko, W.; Lechniak, J.

2010-03-01

The presented paper contains neutronic and thermal-hydraulic (for steady and unsteady states) calculation results prepared to support annex to Safety Analysis Report for MARIA reactor in order to obtain approval for program of testing low-enriched uranium (LEU) lead test fuel assemblies (LTFA) manufactured by CERCA. This includes presentation of the limits and operational constraints to be in effect during the fuel testing investigations. Also, the scope of testing program (which began in August 2009), including additional measurements and monitoring procedures, is described.

Effects of Levels of Automation for Advanced Small Modular Reactors: Impacts on Performance, Workload, and Situation Awareness

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

Johanna Oxstrand; Katya Le Blanc

The Human-Automation Collaboration (HAC) research effort is a part of the Department of Energy (DOE) sponsored Advanced Small Modular Reactor (AdvSMR) program conducted at Idaho National Laboratory (INL). The DOE AdvSMR program focuses on plant design and management, reduction of capital costs as well as plant operations and maintenance costs (O&M), and factory production costs benefits.

Method for depleting BWRs using optimal control rod patterns

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

Taner, M.S.; Levine, S.H.; Hsiao, M.Y.

1991-01-01

Control rod (CR) programming is an essential core management activity for boiling water reactors (BWRs). After establishing a core reload design for a BWR, CR programming is performed to develop a sequence of exposure-dependent CR patterns that assure the safe and effective depletion of the core through a reactor cycle. A time-variant target power distribution approach has been assumed in this study. The authors have developed OCTOPUS to implement a new two-step method for designing semioptimal CR programs for BWRs. The optimization procedure of OCTOPUS is based on the method of approximation programming and uses the SIMULATE-E code for nucleonicsmore » calculations.« less

Hanford Laboratories monthly activities report, February 1964

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

Not Available

1964-03-16

This is the monthly report for the Hanford Laboratories Operation, February, 1964. Reactor fuels, chemistry, dosimetry, separation process, reactor technology financial activities, biology operation, physics and instrumentation research, employee relations, applied mathematics, programming, and radiation protection are discussed.

10 CFR 140.5 - Communications.

Code of Federal Regulations, 2012 CFR

2012-01-01

...: ATTN: Document Control Desk, Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of Federal and State Materials and Environmental Management Programs, or..., Rockville, Maryland; or, where practicable, by electronic submission, for example, via Electronic...

10 CFR 140.5 - Communications.

Code of Federal Regulations, 2011 CFR

2011-01-01

...: ATTN: Document Control Desk, Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of Federal and State Materials and Environmental Management Programs, or..., Rockville, Maryland; or, where practicable, by electronic submission, for example, via Electronic...

Biaxial Creep Specimen Fabrication

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

JL Bump; RF Luther

This report documents the results of the weld development and abbreviated weld qualification efforts performed by Pacific Northwest National Laboratory (PNNL) for refractory metal and superalloy biaxial creep specimens. Biaxial creep specimens were to be assembled, electron beam welded, laser-seal welded, and pressurized at PNNL for both in-pile (JOYO reactor, O-arai, Japan) and out-of-pile creep testing. The objective of this test campaign was to evaluate the creep behavior of primary cladding and structural alloys under consideration for the Prometheus space reactor. PNNL successfully developed electron beam weld parameters for six of these materials prior to the termination of the Navalmore » Reactors program effort to deliver a space reactor for Project Prometheus. These materials were FS-85, ASTAR-811C, T-111, Alloy 617, Haynes 230, and Nirnonic PE16. Early termination of the NR space program precluded the development of laser welding parameters for post-pressurization seal weldments.« less

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

Davis, R. S.

The following are specific topics of this paper: 1.There is much creativity in the manner in which Dimensional Generator can be applied to a specific programming task [2]. This paper tells how Dimensional Generator was applied to a reactor-physics task. 2. In this first practical use, Dimensional Generator itself proved not to need change, but a better user interface was found necessary, essentially because the relevance of Dimensional Generator to reactor physics was initially underestimated. It is briefly described. 3. The use of Dimensional Generator helps make reactor-physics source code somewhat simpler. That is explained here with brief examples frommore » BURFEL-PC and WIMSBURF. 4. Most importantly, with the help of Dimensional Generator, all erroneous physical expressions were automatically detected. The errors are detailed here (in spite of the author's embarrassment) because they show clearly, both in theory and in practice, how Dimensional Generator offers quality enhancement of reactor-physics programming. (authors)« less

Final summary report for 1989 inservice inspection (ISI) of SRS (Savannah River Site) 100-P Reactor tank

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

Morrison, J.M.; Loibl, M.W.

1989-12-15

The integrity of the SRS reactor tanks is a key factor affecting their suitability for continued service since, unlike the external piping system and components, the tanks are virtually irreplaceable. Cracking in various areas of the process water piping systems has occurred beginning in 1960 as a result of several degradation mechanisms, chiefly intergranular stress corrosion cracking (IGSCC) and chloride-induced transgranular cracking. IGSCC, currently the primary degradation mechanism, also occurred in the knuckle'' region (tank wall-to-bottom tube sheet transition piece) unique to C Reactor and was eventually responsible for that reactor being deactivated in 1985. A program of visual examinationsmore » of the SRS reactor tanks was initiated in 1968, which used a specially designed immersible periscope. Under that program the condition of the accessible tank welds and associated heat affected zones (HAZ) was evaluated on a five-year frequency. Prior to 1986, the scope of these inspections comprised approximately 20 percent of the accessible weld area. In late 1986 and early 1987 the scope of the inspections was expanded and a 100 percent visual inspection of accessible welds was performed of the P-, L-, and K-Reactor tanks. Supplemental dye penetrant examinations were performed in L Reactor on selected areas which showed visual indications. No evidence of cracking was detected in any of these inspections of the P-, L-, and K-Reactor tanks. 17 refs., 7 figs.« less

Analysis of JKT01 Neutron Flux Detector Measurements In RSG-GAS Reactor Using LabVIEW

NASA Astrophysics Data System (ADS)

Rokhmadi; Nur Rachman, Agus; Sujarwono; Taryo, Taswanda; Sunaryo, Geni Rina

2018-02-01

The RSG-GAS Reactor, one of the Indonesia research reactors and located in Serpong, is owned by the National Nuclear Energy Agency (BATAN). The RSG-GAS reactor has operated since 1987 and some instrumentation and control systems are considered to be degraded and ageing. It is therefore, necessary to evaluate the safety of all instrumentation and controls and one of the component systems to be evaluated is the performance of JKT01 neutron flux detector. Neutron Flux Detector JKT01 basically detects neutron fluxes in the reactor core and converts it into electrical signals. The electrical signal is then forwarded to the amplifier (Amplifier) to become the input of the reactor protection system. One output of it is transferred to the Main Control Room (RKU) showing on the analog meter as an indicator used by the reactor operator. To simulate all of this matter, a program to simulate the output of the JKT01 Neutron Flux Detector using LabVIEW was developed. The simulated data is estimated using a lot of equations also formulated in LabVIEW. The calculation results are also displayed on the interface using LabVIEW available in the PC. By using this simulation program, it is successful to perform anomaly detection experiments on the JKT01 detector of RSG-GAS Reactor. The simulation results showed that the anomaly JKT01 neutron flux using electrical-current-base are respectively, 1.5×,1.7× and 2.0×.

Methods and codes for neutronic calculations of the MARIA research reactor.

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

Andrzejewski, K.; Kulikowska, T.; Bretscher, M. M.

2002-02-18

The core of the MARIA high flux multipurpose research reactor is highly heterogeneous. It consists of beryllium blocks arranged in 6 x 8 matrix, tubular fuel assemblies, control rods and irradiation channels. The reflector is also heterogeneous and consists of graphite blocks clad with aluminum. Its structure is perturbed by the experimental beam tubes. This paper presents methods and codes used to calculate the MARIA reactor neutronics characteristics and experience gained thus far at IAE and ANL. At ANL the methods of MARIA calculations were developed in connection with the RERTR program. At IAE the package of programs was developedmore » to help its operator in optimization of fuel utilization.« less

Lunar in-core thermionic nuclear reactor power system conceptual design

NASA Technical Reports Server (NTRS)

Mason, Lee S.; Schmitz, Paul C.; Gallup, Donald R.

1991-01-01

This paper presents a conceptual design of a lunar in-core thermionic reactor power system. The concept consists of a thermionic reactor located in a lunar excavation with surface mounted waste heat radiators. The system was integrated with a proposed lunar base concept representative of recent NASA Space Exploration Initiative studies. The reference mission is a permanently-inhabited lunar base requiring a 550 kWe, 7 year life central power station. Performance parameters and assumptions were based on the Thermionic Fuel Element (TFE) Verification Program. Five design cases were analyzed ranging from conservative to advanced. The cases were selected to provide sensitivity effects on the achievement of TFE program goals.

THE SM-1 ENVIRONMENTAL RADIOLOGICAL MONITORING PROGRAM, NOVEMBER 1954- DECEMBER 1960

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

Pressman, M; Pruett, P B

1961-08-31

BS>An environmental radiological monitoring program was conducted. All data obtained during a period extending from l 1/2 years prior to SM-1 reactor start-up through more than 3 years of reactor operation are summarized. The period extended from November 1954 through December 1960. Samples assayed for radioactivity include river water and bottom silt, SM-1 condenser cooling water, subsurface ground water, rain and snow, atmospheric particles obtained by air filtration and fallout, and biota. The report concludes that after more than 3 years of SM-1 reactor operation, no significant increase has been noted in the radiological background level in the Fort Belvoirmore » area.« less

The current state of the Russian reduced enrichment research reactors program

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

Aden, V.G.; Kartashov, E.F.; Lukichev, V.A.

1997-08-01

During the last year after the 16-th International Conference on Reducing Fuel Enrichment in Research Reactors held in October, 1993 in Oarai, Japan, the conclusive stage of the Program on reducing fuel enrichment (to 20% in U-235) in research reactors was finally made up in Russia. The Program was started late in 70th and the first stage of the Program was completed by 1986 which allowed to reduce fuel enrichment from 80-90% to 36%. The completion of the Program current stage, which is counted for 5-6 years, will exclude the use of the fuel enriched by more than 20% frommore » RF to other countries such as: Poland, Czeck Republick, Hungary, Roumania, Bulgaria, Libya, Viet-Nam, North Korea, Egypt, Latvia, Ukraine, Uzbekistan and Kazakhstan. In 1994 the Program, approved by RF Minatom authorities, has received the status of an inter-branch program since it was admitted by the RF Ministry for Science and Technical Policy. The Head of RF Minatom central administrative division N.I.Ermakov was nominated as the Head of the Russian Program, V.G.Aden, RDIPE Deputy Director, was nominated as the scientific leader. The Program was submitted to the Commission for Scientific, Technical and Economical Cooperation between USA and Russia headed by Vice-President A. Gore and Prime Minister V. Chemomyrdin and was given support also.« less

Eddy Current Flow Measurements in the FFTF

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

Nielsen, Deborah L.; Polzin, David L.; Omberg, Ronald P.

2017-02-02

The Fast Flux Test Facility (FFTF) is the most recent liquid metal reactor (LMR) to be designed, constructed, and operated by the U.S. Department of Energy (DOE). The 400-MWt sodium-cooled, fast-neutron flux reactor plant was designed for irradiation testing of nuclear reactor fuels and materials for liquid metal fast breeder reactors. Following shut down of the Clinch River Breeder Reactor Plant (CRBRP) project in 1983, FFTF continued to play a key role in providing a test bed for demonstrating performance of advanced fuel designs and demonstrating operation, maintenance, and safety of advanced liquid metal reactors. The FFTF Program provides valuablemore » information for potential follow-on reactor projects in the areas of plant system and component design, component fabrication, fuel design and performance, prototype testing, site construction, and reactor control and operations. This report provides HEDL-TC-1344, “ECFM Flow Measurements in the FFTF Using Phase-Sensitive Detectors”, March 1979.« less

Gaseous fuel reactors for power systems

NASA Technical Reports Server (NTRS)

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

1977-01-01

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

PLASTIC-SASS--A COMPUTER PROGRAM FOR STRESSES AND DEFLECTIONS IN A REACTOR SUBASSEMBLY UNDER THERMAL, HYDRAULIC, AND FUEL EXPANSION LOADS

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

Friedrich, C.M.

1963-05-01

PLASTlC-SASS, an ALTAC-3 computer program that determines stresses and deflections in a flat-plate, rectangular reactor subassembly is described. Elastic, plastic, and creep properties are used to calculate the results of temperature, pressure, and fuel expansion. Plate deflections increase or decrease local channel thicknesses and thus produce a hydraulic load which is a function of fuel plate deflection. (auth)

Progress in the Production of JP-8 Based Hydrogen and Advanced Tactical Fuels for Military Applications

DTIC Science & Technology

2011-02-01

of a multi- year program to develop, optimize, and demonstrate the military viability of a technology for on-demand production of high...continuous reactor system used for kinetic rate data experiment 86 52 Schematic of a differential reactor. The catalyst bed is kept small , and...program to develop, optimize, and demonstrate the military viability of a technology for on-demand production of high-pressure hydrogen for fuel

Nonproliferation and Threat Reduction Assistance: U.S. Programs in the Former Soviet Union

DTIC Science & Technology

2011-04-26

large - scale former BW-related facilities so that they can perform peaceful research issues such as infectious diseases. The Global Threat Reduction...indicated that it may not pursue the MOX program to eliminate its plutonium, opting instead for the construction of fast breeder reactors that could...burn plutonium directly for energy production. The United States might not fund this effort, as many in the United States argue that breeder reactors

Progress of the RERTR program in 2001.

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

Travelli, A.

2002-03-07

This paper describes the 2001 progress achieved by the Reduced Enrichment for Research and Test Reactors (RERTR) Program in collaboration with its many international partners. Postirradiation examinations of microplates have continued to reveal excellent irradiation behavior of U-Mo dispersion fuels in a variety of compositions and irradiating conditions. Irradiation of two new batches of miniplates of greater sizes was completed in the ATR to investigate the swelling behavior of these fuels under prototypic conditions. These materials hold the promise of achieving the program goal of developing LEU research reactor fuels with uranium densities in the 8-9 g/cm{sup 3} range. Qualificationmore » of the U-Mo dispersion fuels has been delayed by a patent issue involving KAERI. Test fuel elements with uranium density of 6 g/cm{sup 3} are being fabricated by BWXT and are expected to begin undergoing irradiation in the HFR-Petten reactor around March 2003, with a goal of qualifying this fuel by mid-2005. U-Mo fuel with uranium density of 8-9 g/cm{sup 3} is expected to be qualified by mid-2007. Final irradiation tests of LEU {sup 99}Mo targets in the RAS-GAS reactor at BATAN, in Indonesia, had to be postponed because of the 9/11 attacks, but the results collected to date indicate that these targets will soon be ready for commercial production. Excellent cooperation is also in progress with the CNEA in Argentina, MDSN/AECL in Canada, and ANSTO in Australia. Irradiation testing of five WWR-M2 tube-type fuel assemblies fabricated by the NZChK and containing LEU UO{sub 2} dispersion fuel was successfully completed within the Russian RERTR program. A new LEU U-Mo pin-type fuel that could be used to convert most Russian-designed research reactors has been developed by VNIINM and is ready for testing. Four additional shipments containing 822 spent fuel assemblies from foreign research reactors were accepted by the U.S. by September 30, 2001. Altogether, 4,562 spent fuel assemblies from foreign research reactors had been received by that date by the U.S. under the FRR SNF acceptance policy. The RERTR program is aggressively pursuing qualification of high-density LEU U-Mo dispersion fuels, with the dual goal of enabling further conversions and of developing a substitute for LEU silicide fuels that can be more easily disposed of after expiration of the U.S. FRR SNF Acceptance Program. As in the past, the success of the RERTR program will depend on the international friendship and cooperation that has always been its trademark.« less

Review of the TREAT Conversion Conceptual Design and Fuel Qualification Plan

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

Diamond, David

The U.S. Department of Energy (DOE) is preparing to re establish the capability to conduct transient testing of nuclear fuels at the Idaho National Laboratory (INL) Transient Reactor Test (TREAT) facility. The original TREAT core went critical in February 1959 and operated for more than 6,000 reactor startups before plant operations were suspended in 1994. DOE is now planning to restart the reactor using the plant's original high-enriched uranium (HEU) fuel. At the same time, the National Nuclear Security Administration (NNSA) Office of Material Management and Minimization Reactor Conversion Program is supporting analyses and fuel fabrication studies that will allowmore » for reactor conversion to low-enriched uranium (LEU) fuel (i.e., fuel with less than 20% by weight 235U content) after plant restart. The TREAT Conversion Program's objectives are to perform the design work necessary to generate an LEU replacement core, to restore the capability to fabricate TREAT fuel element assemblies, and to implement the physical and operational changes required to convert the TREAT facility to use LEU fuel.« less

Current Abstracts Nuclear Reactors and Technology

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

Bales, J.D.; Hicks, S.C.

1993-01-01

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 acquisition programs or interagency agreements and international information obtained through the International Energy Agency`smore » 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.« less

The CABRI fast neutron Hodoscope: Renovation, qualification program and first results following the experimental reactor restart

NASA Astrophysics Data System (ADS)

Chevalier, V.; Mirotta, S.; Guillot, J.; Biard, B.

2018-01-01

The CABRI experimental pulse reactor, located at the Cadarache nuclear research center, southern France, is devoted to the study of Reactivity Initiated Accidents (RIA). For the purpose of the CABRI International Program (CIP), managed and funded by IRSN, in the framework of an OECD/NEA agreement, a huge renovation of the facility has been conducted since 2003. The Cabri Water Loop was then installed to ensure prototypical Pressurized Water Reactor (PWR) conditions for testing irradiated fuel rods. The hodoscope installed in the CABRI reactor is a unique online fuel motion monitoring system, operated by IRSN and dedicated to the measurement of the fast neutrons emitted by the tested rod during the power pulse. It is one of the distinctive features of the CABRI reactor facility, which is operated by CEA. The system is able to determine the fuel motion, if any, with a time resolution of 1 ms and a spatial resolution of 3 mm. The hodoscope equipment has been upgraded as well during the CABRI facility renovation. This paper presents the main outcomes achieved with the hodoscope since October 2015, date of the first criticality of the CABRI reactor in its new Cabri Water Loop configuration. Results obtained during reactor commissioning phase functioning, either in steady-state mode (at low and high power, up to 23 MW) or in transient mode (start-up, possibly beyond 20 GW), are discussed.

Evaluation of the magnitude and effects of bundle duct interaction in fuel assemblies at developmental plant conditions

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

Serell, D.C.; Kaplan, S.

1980-09-01

Purpose of this evaluation is to estimate the magnitude and effects of irradiation and creep induced fuel bundle deformations in the developmental plant. This report focuses on the trends of the results and the ability of present models to evaluate the assembly temperatures in the presence of bundle deformation. Although this analysis focuses on the developmental plant, the conclusions are applicable to LMFBR fuel assemblies in general if they have wire spacers.

Hanford Laboratories Operation monthly activities report, September 1961

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

Not Available

1961-10-16

This is the monthly report for the Hanford Laboratories Operation September 1961. Reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, biology operation, physics and instrumentation research, operations research and synthesis, programming, and radiation protection operation are discussed.

77 FR 68162 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Regulatory...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-11-15

... Water Reactors,'' Revision 2, and Regulatory Guide 1.79.1, ``Initial Test Program of Emergency Core... White Flint North building, 11555 Rockville Pike, Rockville, MD. After registering with security, please...

Overview of DOE-NE Proliferation and Terrorism Risk Assessment

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

Sadasivan, Pratap

2012-08-24

Research objectives are: (1) Develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of current reactors; (2) Develop improvements in the affordability of new reactors to enable nuclear energy; (3) Develop Sustainable Nuclear Fuel Cycles; and (4) Understand and minimize the risks of nuclear proliferation and terrorism. The goal is to enable the use of risk information to inform NE R&D program planning. The PTRA program supports DOE-NE's goal of using risk information to inform R&D program planning. The FY12 PTRA program is focused on terrorism risk. The program includes a mixmore » of innovative methods that support the general practice of risk assessments, and selected applications.« less

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

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

Rice, R. E.

1960-02-01

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

Naval Reactors Prime Contractor Team (NRPCT) Experiences and Considerations With Irradiation Test Performance in an International Environment

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

MH Lane

2006-02-15

This letter forwards a compilation of knowledge gained regarding international interactions and issues associated with Project Prometheus. The following topics are discussed herein: (1) Assessment of international fast reactor capability and availability; (2) Japanese fast reactor (JOYO) contracting strategy; (3) NRPCT/Program Office international contract follow; (4) Completion of the Japan Atomic Energy Agency (JAEA)/Pacific Northwest National Laboratory (PNNL) contract for manufacture of reactor test components; (5) US/Japanese Departmental interactions and required Treaties and Agreements; and (6) Non-technical details--interactions and considerations.

The Rockwell SR-100G reactor turboelectric space power system

NASA Technical Reports Server (NTRS)

Anderson, R. V.

1985-01-01

During FY 1982 and 1983, Rockwell International performed system and subsystem studies for space reactor power systems. These studies drew on the expertise gained from the design and flight of the SNAP-10A space nuclear reactor system. These studies, performed for the SP-100 Program, culminated in the selection of a reactor-turboelectric (gas Brayton) system for the SP-100 application; this system is called the SR-100G. This paper describes the features of the system and provides references where more detailed information can be obtained.

Cladding and duct materials for advanced nuclear recycle reactors

NASA Astrophysics Data System (ADS)

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

2008-01-01

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

Next generation fuel irradiation capability in the High Flux Reactor Petten

NASA Astrophysics Data System (ADS)

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

2009-07-01

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

The IRIS Spool-Type Reactor Coolant Pump

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

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

2002-07-01

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

Digital computer program for nuclear reactor design water properties (LWBR Development Program)

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

Lynn, L.L.

1967-07-01

An edit program MO899 for the tabulation of thermodynamic and transport properties of liquid and vapor water, frequently used in design calculations for pressurized water nuclear reactors, is described. The data tabulated are obtained from a FORTRAN IV subroutine named HOH. Values of enthalpy, specific volume, viscosity, and thermal conductivity are given for the following ranges: pressure from one bar (14.5 psia) to 175 bars (2538 psia) and temperature from as much as 320 deg C (608 deg F) below saturation up to 500 deg C (932 deg F) above saturation. (NSA 21: 38472)

Space reactor power 1986 - A year of choices and transition

NASA Technical Reports Server (NTRS)

Wiley, R. L.; Verga, R. L.; Schnyer, A. D.; Sholtis, J. A., Jr.; Wahlquist, E. J.

1986-01-01

Both the SP-100 and Multimegawatt programs have made significant progress over the last year and that progress is the focus of this paper. In the SP-100 program the thermoelectric energy conversion concept powered by a compact, high-temperature, lithium-cooled, uranium-nitride-fueled fast spectrum reactor was selected for engineering development and ground demonstration testing at an electrical power level of 300 kilowatts. In the Multimegawatt program, activities moved from the planning phase into one of technology development and assessment with attendant preliminary definition and evaluation of power concepts against requirements of the Strategic Defense Initiative.

Review of the Tri-Agency Space Nuclear Reactor Power System Technology Program

NASA Technical Reports Server (NTRS)

Ambrus, J. H.; Wright, W. E.; Bunch, D. F.

1984-01-01

The Space Nuclear Reactor Power System Technology Program designated SP-100 was created in 1983 by NASA, the U.S. Department of Defense, and the Defense Advanced Research Projects Agency. Attention is presently given to the development history of SP-100 over the course of its first year, in which it has been engaged in program objectives' definition, the analysis of civil and military missions, nuclear power system functional requirements' definition, concept definition studies, the selection of primary concepts for technology feasibility validation, and the acquisition of initial experimental and analytical results.

Nuclear safety for the space exploration initiative

NASA Technical Reports Server (NTRS)

Dix, Terry E.

1991-01-01

The results of a study to identify potential hazards arising from nuclear reactor power systems for use on the lunar and Martian surfaces, related safety issues, and resolutions of such issues by system design changes, operating procedures, and other means are presented. All safety aspects of nuclear reactor power systems from prelaunch ground handling to eventual disposal were examined consistent with the level of detail for SP-100 reactor design at the 1988 System Design Review and for launch vehicle and space transport vehicle designs and mission descriptions as defined in the 90-day Space Exploration Initiative (SEI) study. Information from previous aerospace nuclear safety studies was used where appropriate. Safety requirements for the SP-100 space nuclear reactor system were compiled. Mission profiles were defined with emphasis on activities after low earth orbit insertion. Accident scenarios were then qualitatively defined for each mission phase. Safety issues were identified for all mission phases with the aid of simplified event trees. Safety issue resolution approaches of the SP-100 program were compiled. Resolution approaches for those safety issues not covered by the SP-100 program were identified. Additionally, the resolution approaches of the SP-100 program were examined in light of the moon and Mars missions.

77 FR 64148 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Regulatory...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-18

... Regulatory Guides (RG) RG 1.79, ````Preoperational Testing of Emergency Core Cooling Systems for Pressurized Water Reactors,'' Revision 2 and RG 1.79.1, ``Initial Test Program of Emergency Core Cooling Systems for...

77 FR 34367 - Proposed Subsequent Arrangement

Federal Register 2010, 2011, 2012, 2013, 2014

2012-06-11

... reactors, and a research reactor, at the Post Irradiation Examination Facility (PIEF), the Irradiated.../2011, ``Post-Irradiation Examination and R&D Programs Using Irradiated Fuels at KAERI,'' dated June... fuel elements for post-irradiation examination and for research, development and manufacture of DUPIC...

Thermal Hydraulic Analysis of a Packed Bed Reactor Fuel Element

DTIC Science & Technology

1989-05-25

Engineer and Master of Science in Nuclear Engineering. ABSTRACT A model of the behavior of a packed bed nuclear reactor fuel element is developed . It...RECOMMENDATIONS FOR FURTHER INVESTIGATION .................... 150 APPENDIX A FUEL ELEMENT MODEL PROGRAM DESIGN AND OPERA- T IO N...follow describe the details of the packed bed reactor and then discuss the development of the mathematical representations of the fuel element. These are

COST FUNCTION STUDIES FOR POWER REACTORS

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

Heestand, J.; Wos, L.T.

1961-11-01

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

Status of Fuel Development and Manufacturing for Space Nuclear Reactors at BWX Technologies

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

Carmack, W.J.; Husser, D.L.; Mohr, T.C.

2004-02-04

New advanced nuclear space propulsion systems will soon seek a high temperature, stable fuel form. BWX Technologies Inc (BWXT) has a long history of fuel manufacturing. UO2, UCO, and UCx have been fabricated at BWXT for various US and international programs. Recent efforts at BWXT have focused on establishing the manufacturing techniques and analysis capabilities needed to provide a high quality, high power, compact nuclear reactor for use in space nuclear powered missions. To support the production of a space nuclear reactor, uranium nitride has recently been manufactured by BWXT. In addition, analytical chemistry and analysis techniques have been developedmore » to provide verification and qualification of the uranium nitride production process. The fabrication of a space nuclear reactor will require the ability to place an unclad fuel form into a clad structure for assembly into a reactor core configuration. To this end, BWX Technologies has reestablished its capability for machining, GTA welding, and EB welding of refractory metals. Specifically, BWX Technologies has demonstrated GTA welding of niobium flat plate and EB welding of niobium and Nb-1Zr tubing. In performing these demonstration activities, BWX Technologies has established the necessary infrastructure to manufacture UO2, UCx, or UNx fuel, components, and complete reactor assemblies in support of space nuclear programs.« less

DIFFUSE: a FORTRAN program for design computation of tritium transport through thermonuclear reactor components by combined ordinary and thermal diffusion when the principal resistance to diffusion is the bulk metal

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

Pendergrass, J.H.

1977-10-01

Based on the theory developed in an earlier report, a FORTRAN computer program, DIFFUSE, was written. It computes, for design purposes, rates of transport of hydrogen isotopes by temperature-dependent quasi-unidirectional, and quasi-static combined ordinary and thermal diffusion through thin, hot thermonuclear reactor components that can be represented by composites of plane, cylindrical-shell, and spherical-shell elements when the dominant resistance to transfer is that of the bulk metal. The program is described, directions for its use are given, and a listing of the program, together with sample problem results, is presented.

SP-100 design, safety, and testing

NASA Technical Reports Server (NTRS)

Cox, Carl. M.; Mahaffey, Michael M.; Smith, Gary L.

1991-01-01

The SP-100 Program is developing a nuclear reactor power system that can enhance and/or enable future civilian and military space missions. The program is directed to develop space reactor technology to provide electrical power in the range of tens to hundreds of kilowatts. The major nuclear assembly test is to be conducted at the Hanford Site near Richland, Washington, and is designed to validate the performance of the 2.4-MWt nuclear and heat transport assembly.

Nonproliferation and Threat Reduction Assistance: U.S, Programs in the Former Soviet Union

DTIC Science & Technology

2008-03-26

reconfigure its large - scale former BW-related facilities so that they can perform peaceful research issues such as infectious diseases. For FY2004, the Bush...program to eliminate its plutonium, opting instead for the construction of fast breeder reactors that could burn plutonium directly for energy production...The United States might not fund this effort, as many in the United States argue that breeder reactors , which produce more plutonium than they

A two-step method for developing a control rod program for boiling water reactors

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

Taner, M.S.; Levine, S.H.; Hsiao, M.Y.

1992-01-01

This paper reports on a two-step method that is established for the generation of a long-term control rod program for boiling water reactors (BWRs). The new method assumes a time-variant target power distribution in core depletion. In the new method, the BWR control rod programming is divided into two steps. In step 1, a sequence of optimal, exposure-dependent Haling power distribution profiles is generated, utilizing the spectral shift concept. In step 2, a set of exposure-dependent control rod patterns is developed by using the Haling profiles generated at step 1 as a target. The new method is implemented in amore » computer program named OCTOPUS. The optimization procedure of OCTOPUS is based on the method of approximation programming, in which the SIMULATE-E code is used to determine the nucleonics characteristics of the reactor core state. In a test in cycle length over a time-invariant, target Haling power distribution case because of a moderate application of spectral shift. No thermal limits of the core were violated. The gain in cycle length could be increased further by broadening the extent of the spetral shift.« less

Modernization of existing VVER-1000 surveillance programs

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

Kochkin, V.; Erak, D.; Makhotin, D.

2011-07-01

According to generally accepted world practice, evaluation of the reactor pressure vessel (RPV) material behavior during operation is carried out using tests of surveillance specimens. The main objective of the surveillance program consists in insurance of safe RPV operation during the design lifetime and lifetime-extension period. At present, the approaches of pressure vessels residual life validation based on the test results of their surveillance specimens have been developed and introduced in Russia and are under consideration in other countries where vodo-vodyanoi energetichesky reactors- (VVER-) 1000 are in operation. In this case, it is necessary to ensure leading irradiation of surveillancemore » specimens (as compared to the pressure vessel wall) and to provide uniformly irradiated specimen groups for mechanical testing. Standard surveillance program of VVER-1000 has several significant shortcomings and does not meet these requirements. Taking into account program of lifetime extension of VVER-1000 operating in Russia, it is necessary to carry out upgrading of the VVER-1000 surveillance program. This paper studies the conditions of a surveillance specimen's irradiation and upgrading of existing sets to provide monitoring and prognosis of RPV material properties for extension of the reactor's lifetime up to 60 years or more. (authors)« less

Conversion Preliminary Safety Analysis Report for the NIST Research Reactor

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

Diamond, D. J.; Baek, J. S.; Hanson, A. L.

The NIST Center for Neutron Research (NCNR) is a reactor-laboratory complex providing the National Institute of Standards and Technology (NIST) and the nation with a world-class facility for the performance of neutron-based research. The heart of this facility is the NIST research reactor (aka NBSR); a heavy water moderated and cooled reactor operating at 20 MW. It is fueled with high-enriched uranium (HEU) fuel elements. A Global Threat Reduction Initiative (GTRI) program is underway to convert the reactor to low-enriched uranium (LEU) fuel. This program includes the qualification of the proposed fuel, uranium and molybdenum alloy foil clad in anmore » aluminum alloy, and the development of the fabrication techniques. This report is a preliminary version of the Safety Analysis Report (SAR) that would be submitted to the U.S. Nuclear Regulatory Commission (NRC) for approval prior to conversion. The report follows the recommended format and content from the NRC codified in NUREG-1537, “Guidelines for Preparing and Reviewing Applications for the Licensing of Non-power Reactors,” Chapter 18, “Highly Enriched to Low-Enriched Uranium Conversions.” The emphasis in any conversion SAR is to explain the differences between the LEU and HEU cores and to show the acceptability of the new design; there is no need to repeat information regarding the current reactor that will not change upon conversion. Hence, as seen in the report, the bulk of the SAR is devoted to Chapter 4, Reactor Description, and Chapter 13, Safety Analysis.« less

Status and progress of the RERTR program in the year 2000.

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

Travelli, A.

2000-09-28

This paper describes the progress achieved by the Reduced Enrichment for Research and Test Reactors (RERTR) Program in collaboration with its many international partners during the year 2000 and discusses the main activities planned for the year 2001. The past year was characterized by important accomplishments and events for the RERTR program. Four additional shipments containing 503 spent fuel assemblies from foreign research reactors were accepted by the U.S. Altogether, 3,740 spent fuel assemblies from foreign research reactors have been received by the U.S. under the acceptance policy. Postirradiation examinations of three batches of microplates have continued to reveal excellentmore » irradiation behavior of U-MO dispersion fuels in a variety of compositions and irradiating conditions. h-radiation of two new batches of miniplates of greater sizes is in progress in the ATR to investigate me swelling behavior of these fuels under prototypic conditions. These materials hold the promise of achieving the program goal of developing LEU research reactor fuels with uranium densities in the 8-9 g /cm{sup 3} range. Qualification of the U-MO dispersion fuels is proceeding on schedule. Test fuel elements with 6 gU/cm{sup 3} are being fabricated by BWXT and are scheduled to begin undergoing irradiation in the HFR-Petten in the spring of 2001, with a goal of qualifying this fuel by the end of 2003. U-Mo with 8-9 gU/cm{sup 3} is planned to be qualified by the end of 2005. Joint LEU conversion feasibility studies were completed for HFR-Petten and for SAFARI-1. Significant improvements were made in the design of LEU metal-foil annular targets that would allow efficient production of fission {sup 99}Mo. Irradiations in the RAS-GAS reactor showed that these targets can formed from aluminum tubes, and that the yield and purity of their product from the acidic process were at least as good as those from the HEU Cintichem targets. Progress was made on irradiation testing of LEU UO{sub 2} dispersion fuel and on LEU conversion feasibility studies in the Russian RERTR program. Conversion of the BER-11reactor in Berlin, Germany, was completed and conversion of the La Reins reactor in Santiago, Chile, began. These are exciting times for the program. In the fuel development area, the RERTR program is aggressively pursuing qualification of high-density LEU U-Mo dispersion fuels, with the dual goal of enabling fi.uther conversions and of developing a substitute for LEU silicide fuels that can be more easily disposed of after expiration of the FRR SNF Acceptance Program. The {sup 99}Mo effort has reached the point where it appears feasible for all the {sup 99}Mo producers of the world to agree jointly to a common course of action leading to the elimination of HEU use in their processes. As in the past, the success of the RERTR program will depend on the international friendship and cooperation that has always been its trademark.« less

KWU's high conversion reactor concept - An economical evolution of modern pressurized water reactor technology toward improved uranium ore utilization

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

Markl, H.; Goetzmann, C.A.; Moldaschl, H.

The Kraftwerk Union AG high conversion reactor represents a quasi-standard PWR with fuel assemblies of more or less uniformly enriched fuel rods, arranged in a tight hexagonal array with a pitch-to-diameter ratio p/d approx. = 1.12. High fuel enrichment as well as a high conversion ratio of --0.9 will provide the potential for high burnup values up to 70 000 MWd/tonne and a low fissile material consumption. The overall objective of the actual RandD program is to have the technical feasibility, including that for licensibility, established by the early 1990s as a prerequisite for deciding whether to enter a demonstrationmore » plant program.« less

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

NASA Astrophysics Data System (ADS)

Carlson, William F.; Bitten, Ernest J.

1993-01-01

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

Space station prototype Sabatier reactor design verification testing

NASA Technical Reports Server (NTRS)

Cusick, R. J.

1974-01-01

A six-man, flight prototype carbon dioxide reduction subsystem for the SSP ETC/LSS (Space Station Prototype Environmental/Thermal Control and Life Support System) was developed and fabricated for the NASA-Johnson Space Center between February 1971 and October 1973. Component design verification testing was conducted on the Sabatier reactor covering design and off-design conditions as part of this development program. The reactor was designed to convert a minimum of 98 per cent hydrogen to water and methane for both six-man and two-man reactant flow conditions. Important design features of the reactor and test conditions are described. Reactor test results are presented that show design goals were achieved and off-design performance was stable.

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

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

D.M. McEligot; K. G. Condie; G. E. McCreery

2005-10-01

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

Scaling Studies for Advanced High Temperature Reactor Concepts, Final Technical Report: October 2014—December 2017

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

Woods, Brian; Gutowska, Izabela; Chiger, Howard

Computer simulations of nuclear reactor thermal-hydraulic phenomena are often used in the design and licensing of nuclear reactor systems. In order to assess the accuracy of these computer simulations, computer codes and methods are often validated against experimental data. This experimental data must be of sufficiently high quality in order to conduct a robust validation exercise. In addition, this experimental data is generally collected at experimental facilities that are of a smaller scale than the reactor systems that are being simulated due to cost considerations. Therefore, smaller scale test facilities must be designed and constructed in such a fashion tomore » ensure that the prototypical behavior of a particular nuclear reactor system is preserved. The work completed through this project has resulted in scaling analyses and conceptual design development for a test facility capable of collecting code validation data for the following high temperature gas reactor systems and events— 1. Passive natural circulation core cooling system, 2. pebble bed gas reactor concept, 3. General Atomics Energy Multiplier Module reactor, and 4. prismatic block design steam-water ingress event. In the event that code validation data for these systems or events is needed in the future, significant progress in the design of an appropriate integral-type test facility has already been completed as a result of this project. Where applicable, the next step would be to begin the detailed design development and material procurement. As part of this project applicable scaling analyses were completed and test facility design requirements developed. Conceptual designs were developed for the implementation of these design requirements at the Oregon State University (OSU) High Temperature Test Facility (HTTF). The original HTTF is based on a ¼-scale model of a high temperature gas reactor concept with the capability for both forced and natural circulation flow through a prismatic core with an electrical heat source. The peak core region temperature capability is 1400°C. As part of this project, an inventory of test facilities that could be used for these experimental programs was completed. Several of these facilities showed some promise, however, upon further investigation it became clear that only the OSU HTTF had the power and/or peak temperature limits that would allow for the experimental programs envisioned herein. Thus the conceptual design and feasibility study development focused on examining the feasibility of configuring the current HTTF to collect validation data for these experimental programs. In addition to the scaling analyses and conceptual design development, a test plan was developed for the envisioned modified test facility. This test plan included a discussion on an appropriate shakedown test program as well as the specific matrix tests. Finally, a feasibility study was completed to determine the cost and schedule considerations that would be important to any test program developed to investigate these designs and events.« less

10 CFR 140.6 - Reports.

Code of Federal Regulations, 2012 CFR

2012-01-01

... Nuclear Reactor Regulation, Director, Office of New Reactors, Director, Office of Federal and State Materials and Environmental Management Programs, or Director, Office of Nuclear Material Safety and... 10 Energy 2 2012-01-01 2012-01-01 false Reports. 140.6 Section 140.6 Energy NUCLEAR REGULATORY...

Assessment of Current Inservice Inspection and Leak Monitoring Practices for Detecting Materials Degradation in Light Water Reactors

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

Anderson, Michael T.; Simonen, Fredric A.; Muscara, Joseph

2016-09-01

An assessment was performed to determine the effectiveness of existing inservice inspection (ISI) and leak monitoring techniques, and recommend improvements, as necessary, to the programs as currently performed for light water reactor (LWR) components. Information from nuclear power plant (NPP) aging studies and from the U. S. Nuclear Regulatory Commission’s Generic Aging Lessons Learned (GALL) report (NUREG-1801) was used to identify components that have already experienced, or are expected to experience, degradation. This report provides a discussion of the key aspects and parameters that constitute an effective ISI program and a discussion of the basis and background against which themore » effectiveness of the ISI and leak monitoring programs for timely detection of degradation was evaluated. Tables based on the GALL components were used to systematically guide the process, and table columns were included that contained the ISI requirements and effectiveness assessment. The information in the tables was analyzed using histograms to reduce the data and help identify any trends. The analysis shows that the overall effectiveness of the ISI programs is very similar for both boiling water reactors (BWRs) and pressurized water reactors (PWRs). The evaluations conducted as part of this research showed that many ISI programs are not effective at detecting degradation before its extent reached 75% of the component wall thickness. This work should be considered as an assessment of NDE practices at this time; however, industry and regulatory activities are currently underway that will impact future effectiveness assessments. A number of actions have been identified to improve the current ISI programs so that degradation can be more reliably detected.« less

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

NASA Technical Reports Server (NTRS)

Srivastava, R.; Gould, R. K.

1979-01-01

The program aims at developing mathematical models and computer codes based on these models, which allow prediction of the product distribution in chemical reactors for converting gaseous silicon compounds to condensed-phase silicon. The major interest is in collecting silicon as a liquid on the reactor walls and other collection surfaces. Two reactor systems are of major interest, a SiCl4/Na reactor in which Si(l) is collected on the flow tube reactor walls and a reactor in which Si(l) droplets formed by the SiCl4/Na reaction are collected by a jet impingement method. During this quarter the following tasks were accomplished: (1) particle deposition routines were added to the boundary layer code; and (2) Si droplet sizes in SiCl4/Na reactors at temperatures below the dew point of Si are being calculated.

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

NASA Astrophysics Data System (ADS)

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

2006-01-01

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

Sister Lab Program Prospective Partner Nuclear Profile: Indonesia

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

Bissani, M; Tyson, S

2006-12-14

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

THE ARMOUR DUST FUELED REACTOR (ADFR)

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

Krucoff, D.

1958-01-01

The A-DFR is based on the use of a fissionable dust carried in a gas. This fuel ferm offers promise of a major economic advance through the use of 2,000 to 3,000 F operating temperatures and a low cost fuel cycle. The development program is described that was initiated to investigate experimentally the proposed fuel and study analytically other reactor characteristics. A brief review of the reactor concept is presented. (W.D.M.)

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

NASA Astrophysics Data System (ADS)

Permana, Sidik

2017-07-01

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

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

NASA Astrophysics Data System (ADS)

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

2015-11-01

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

FFTF Passive Safety Test Data for Benchmarks for New LMR Designs

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

Wootan, David W.; Casella, Andrew M.

Liquid Metal Reactors (LMRs) continue to be considered as an attractive concept for advanced reactor design. Software packages such as SASSYS are being used to im-prove new LMR designs and operating characteristics. Significant cost and safety im-provements can be realized in advanced liquid metal reactor designs by emphasizing inherent or passive safety through crediting the beneficial reactivity feedbacks associ-ated with core and structural movement. This passive safety approach was adopted for the Fast Flux Test Facility (FFTF), and an experimental program was conducted to characterize the structural reactivity feedback. The FFTF passive safety testing pro-gram was developed to examine howmore » specific design elements influenced dynamic re-activity feedback in response to a reactivity input and to demonstrate the scalability of reactivity feedback results to reactors of current interest. The U.S. Department of En-ergy, Office of Nuclear Energy Advanced Reactor Technology program is in the pro-cess of preserving, protecting, securing, and placing in electronic format information and data from the FFTF, including the core configurations and data collected during the passive safety tests. Benchmarks based on empirical data gathered during operation of the Fast Flux Test Facility (FFTF) as well as design documents and post-irradiation examination will aid in the validation of these software packages and the models and calculations they produce. Evaluation of these actual test data could provide insight to improve analytical methods which may be used to support future licensing applications for LMRs« less

Research Reactor Preparations for the Air Shipment of Highly Enriched Uranium from Romania

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

K. J. Allen; I. Bolshinsky; L. L. Biro

2010-03-01

In June 2009 two air shipments transported both unirradiated (fresh) and irradiated (spent) Russian-origin highly enriched uranium (HEU) nuclear fuel from two research reactors in Romania to the Russian Federation for conversion to low enriched uranium. The Institute for Nuclear Research at Pitesti (SCN Pitesti) shipped 30.1 kg of HEU fresh fuel pellets to Dimitrovgrad, Russia and the Horia Hulubei National Institute of Physics and Nuclear Engineering (IFIN-HH) shipped 23.7 kilograms of HEU spent fuel assemblies from the VVR S research reactor at Magurele, Romania, to Chelyabinsk, Russia. Both HEU shipments were coordinated by the Russian Research Reactor Fuel Returnmore » Program (RRRFR) as part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), were managed in Romania by the National Commission for Nuclear Activities Control (CNCAN), and were conducted in cooperation with the Russian Federation State Corporation Rosatom and the International Atomic Energy Agency. Both shipments were transported by truck to and from respective commercial airports in Romania and the Russian Federation and stored at secure nuclear facilities in Russia until the material is converted into low enriched uranium. These shipments resulted in Romania becoming the 3rd country under the RRRFR program and the 14th country under the GTRI program to remove all HEU. This paper describes the research reactor preparations and license approvals that were necessary to safely and securely complete these air shipments of nuclear fuel.« less

Light Water Reactor Sustainability Program Advanced Instrumentation, Information, and Control Systems Technologies Technical Program Plan for FY 2016

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

Hallbert, Bruce Perry; Thomas, Kenneth David

2015-10-01

Reliable instrumentation, information, and control (II&C) systems technologies are essential to ensuring safe and efficient operation of the U.S. light water reactor (LWR) fleet. These technologies affect every aspect of nuclear power plant (NPP) and balance-of-plant operations. In 1997, the National Research Council conducted a study concerning the challenges involved in modernization of digital instrumentation and control systems in NPPs. Their findings identified the need for new II&C technology integration.

Computer modeling of a hot filament diamond deposition reactor

NASA Technical Reports Server (NTRS)

Kuczmarski, Maria A.; Washlock, Paul A.; Angus, John C.

1991-01-01

A commercial fluid mechanics program, FLUENT, has been applied to the modeling of a hot-filament diamond deposition reactor. Streamlines and contours of constant temperature and species concentrations are obtained for practical reactor geometries and conditions. The modeling is presently restricted to two-dimensional simulations and to a chemical mechanism of ten independent homogeneous and surface reactions. Comparisons are made between predicted power consumption, substrate temperature, and concentrations of atomic hydrogen and methyl-radical with values taken from the literature. The results to date indicate that the modeling can aid in the rational design and analysis of practical reactor configurations.

Schedule and status of irradiation experiments

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

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

1998-09-01

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

Clean catalytic combustor program

NASA Technical Reports Server (NTRS)

Ekstedt, E. E.; Lyon, T. F.; Sabla, P. E.; Dodds, W. J.

1983-01-01

A combustor program was conducted to evolve and to identify the technology needed for, and to establish the credibility of, using combustors with catalytic reactors in modern high-pressure-ratio aircraft turbine engines. Two selected catalytic combustor concepts were designed, fabricated, and evaluated. The combustors were sized for use in the NASA/General Electric Energy Efficient Engine (E3). One of the combustor designs was a basic parallel-staged double-annular combustor. The second design was also a parallel-staged combustor but employed reverse flow cannular catalytic reactors. Subcomponent tests of fuel injection systems and of catalytic reactors for use in the combustion system were also conducted. Very low-level pollutant emissions and excellent combustor performance were achieved. However, it was obvious from these tests that extensive development of fuel/air preparation systems and considerable advancement in the steady-state operating temperature capability of catalytic reactor materials will be required prior to the consideration of catalytic combustion systems for use in high-pressure-ratio aircraft turbine engines.

A probabilistic safety analysis of incidents in nuclear research reactors.

PubMed

Lopes, Valdir Maciel; Agostinho Angelo Sordi, Gian Maria; Moralles, Mauricio; Filho, Tufic Madi

2012-06-01

This work aims to evaluate the potential risks of incidents in nuclear research reactors. For its development, two databases of the International Atomic Energy Agency (IAEA) were used: the Research Reactor Data Base (RRDB) and the Incident Report System for Research Reactor (IRSRR). For this study, the probabilistic safety analysis (PSA) was used. To obtain the result of the probability calculations for PSA, the theory and equations in the paper IAEA TECDOC-636 were used. A specific program to analyse the probabilities was developed within the main program, Scilab 5.1.1. for two distributions, Fischer and chi-square, both with the confidence level of 90 %. Using Sordi equations, the maximum admissible doses to compare with the risk limits established by the International Commission on Radiological Protection (ICRP) were obtained. All results achieved with this probability analysis led to the conclusion that the incidents which occurred had radiation doses within the stochastic effects reference interval established by the ICRP-64.

TECHNICAL SCOPE OF GAS-COOLED REACTOR FUEL ELEMENT IRRADIATION PROGRAM

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

None

A set of 55 experiments hss been outiined to provide a minimum irradiation program for selection of UO/sub 2/, pellet geometry and fabricntion techniques, and canning technology. These experiments fall into three catagories: prototype: untts in which radial dimension and heat fluxes sre close to proposed design values, but irradiation times are long; reduced-size prototype for accelerated tests in which most variables will be studied; and miniaurized pellet irradiation to obtain high burnup for fission gas release studies. Reactor space has been found generally available and several installations are now examining their capabilities to participate in the program. A tentativemore » schedule has been drawn to illustrate the feasibility of the program. (auth)« less

Aerosol reactor production of uniform submicron powders

NASA Technical Reports Server (NTRS)

Flagan, Richard C. (Inventor); Wu, Jin J. (Inventor)

1991-01-01

A method of producing submicron nonagglomerated particles in a single stage reactor includes introducing a reactant or mixture of reactants at one end while varying the temperature along the reactor to initiate reactions at a low rate. As homogeneously small numbers of seed particles generated in the initial section of the reactor progress through the reactor, the reaction is gradually accelerated through programmed increases in temperature along the length of the reactor to promote particle growth by chemical vapor deposition while minimizing agglomerate formation by maintaining a sufficiently low number concentration of particles in the reactor such that coagulation is inhibited within the residence time of particles in the reactor. The maximum temperature and minimum residence time is defined by a combination of temperature and residence time that is necessary to bring the reaction to completion. In one embodiment, electronic grade silane and high purity nitrogen are introduced into the reactor and temperatures of approximately 770.degree. K. to 1550.degree. K. are employed. In another embodiment silane and ammonia are employed at temperatures from 750.degree. K. to 1800.degree. K.

Control console replacement at the WPI Reactor

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

Not Available

1992-01-01

With partial funding from the Department of Energy (DOE) University Reactor Instrumentation Upgrade Program (DOE Grant No. DE-FG02-90ER12982), the original control console at the Worcester Polytechnic Institute (WPI) Reactor has been replaced with a modern system. The new console maintains the original design bases and functionality while utilizing current technology. An advanced remote monitoring system has been added to augment the educational capabilities of the reactor. Designed and built by General Electric in 1959, the open pool nuclear training reactor at WPI was one of the first such facilities in the nation located on a university campus. Devoted to undergraduatemore » use, the reactor and its related facilities have been since used to train two generations of nuclear engineers and scientists for the nuclear industry. The reactor power level was upgraded from 1 to 10 kill in 1969, and its operating license was renewed for 20 years in 1983. In 1988, the reactor was converted to low enriched uranium. The low power output of the reactor and ergonomic facility design make it an ideal tool for undergraduate nuclear engineering education and other training.« less

Aerosol reactor production of uniform submicron powders

DOEpatents

Flagan, Richard C.; Wu, Jin J.

1991-02-19

A method of producing submicron nonagglomerated particles in a single stage reactor includes introducing a reactant or mixture of reactants at one end while varying the temperature along the reactor to initiate reactions at a low rate. As homogeneously small numbers of seed particles generated in the initial section of the reactor progress through the reactor, the reaction is gradually accelerated through programmed increases in temperature along the length of the reactor to promote particle growth by chemical vapor deposition while minimizing agglomerate formation by maintaining a sufficiently low number concentration of particles in the reactor such that coagulation is inhibited within the residence time of particles in the reactor. The maximum temperature and minimum residence time is defined by a combination of temperature and residence time that is necessary to bring the reaction to completion. In one embodiment, electronic grade silane and high purity nitrogen are introduced into the reactor and temperatures of approximately 770.degree. K. to 1550.degree. K. are employed. In another embodiment silane and ammonia are employed at temperatures from 750.degree. K. to 1800.degree. K.

Oxidative coupling of methane using inorganic membrane reactor

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

Ma, Y.H.; Moser, W.R.; Dixon, A.G.

1995-12-31

The goal of this research is to improve the oxidative coupling of methane in a catalytic inorganic membrane reactor. A specific target is to achieve conversion of methane to C{sub 2} hydrocarbons at very high selectivity and relatively higher yields than in fixed bed reactors by controlling the oxygen supply through the membrane. A membrane reactor has the advantage of precisely controlling the rate of delivery of oxygen to the catalyst. This facility permits balancing the rate of oxidation and reduction of the catalyst. In addition, membrane reactors minimize the concentration of gas phase oxygen thus reducing non selective gasmore » phase reactions, which are believed to be a main route for formation of CO{sub x} products. Such gas phase reactions are a cause for decreased selectivity in oxidative coupling of methane in conventional flow reactors. Membrane reactors could also produce higher product yields by providing better distribution of the reactant gases over the catalyst than the conventional plug flow reactors. Modeling work which aimed at predicting the observed experimental trends in porous membrane reactors was also undertaken in this research program.« less

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

NASA Astrophysics Data System (ADS)

Darmawan, R.

2018-01-01

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

Separations in the STATS report

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

Choppin, G.R.

1996-12-31

The Separations Technology and Transmutation Systems (STATS) Committee formed a Subcommittee on Separations. This subcommittee was charged with evaluating the separations proposed for the several reactor and accelerator transmutation systems. It was also asked to review the processing options for the safe management of high-level waste generated by the defense programs, in particular, the special problems involved in dealing with the waste at the U.S. Department of Energy (DOE) facility in Hanford, Washington. Based on the evaluations from the Subcommittee on Separations, the STATS Committee concluded that for the reactor transmutation programs, aqueous separations involving a combination of PUREX andmore » TRUEX solvent extraction processes could be used. However, additional research and development (R&D) would be required before full plant-scale use of the TRUEX technology could be employed. Alternate separations technology for the reactor transmutation program involves pyroprocessing. This process would require a significant amount of R&D before its full-scale application can be evaluated.« less

TRAC-P1: an advanced best estimate computer program for PWR LOCA analysis. I. Methods, models, user information, and programming details

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

Not Available

1978-05-01

The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos Scientific Laboratory (LASL) to provide an advanced ''best estimate'' predictive capability for the analysis of postulated accidents in light water reactors (LWRs). TRAC-Pl provides this analysis capability for pressurized water reactors (PWRs) and for a wide variety of thermal-hydraulic experimental facilities. It features a three-dimensional treatment of the pressure vessel and associated internals; two-phase nonequilibrium hydrodynamics models; flow-regime-dependent constitutive equation treatment; reflood tracking capability for both bottom flood and falling film quench fronts; and consistent treatment of entire accident sequences including the generation of consistent initial conditions.more » The TRAC-Pl User's Manual is composed of two separate volumes. Volume I gives a description of the thermal-hydraulic models and numerical solution methods used in the code. Detailed programming and user information is also provided. Volume II presents the results of the developmental verification calculations.« less

Conceptual Design and Neutronics Analyses of a Fusion Reactor Blanket Simulation Facility

DTIC Science & Technology

1986-01-01

Laboratory (LLL) ORNL Oak Ridge National Laboratory PPPL Princeton Plasma Physics Laboratory RSIC Reactor Shielding Information Center (at ORNL) SS...Module (LBM) to be placed in the TFTR at PPPL . Jassby et al. describe the program, including design, manufacturing techniques. neutronics analyses, and

Isomer Energy Source for Space Propulsion Systems

DTIC Science & Technology

2004-03-01

1,590 Engine F/W (no shield) 3.4 5.0 20.0 A similar core design replacing the fission fuel with the isomer 178Hfm2 is the starting point for this...particles interact and collide with other atoms in the fuel material, reactor core , or coolant, their energy can be transferred to thermal energy...thrust (44). The program produced several reactors that made it all the way through the testing stages of development . The reactors used uranium-235

Environmental Information Document: L-reactor reactivation

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

Mackey, H.E. Jr.

1982-04-01

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

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

NASA Astrophysics Data System (ADS)

Volman, M.

2017-01-01

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

Nuclear electric propulsion reactor control systems status

NASA Technical Reports Server (NTRS)

Ferg, D. A.

1973-01-01

The thermionic reactor control system design studies conducted over the past several years for a nuclear electric propulsion system are described and summarized. The relevant reactor control system studies are discussed in qualitative terms, pointing out the significant advantages and disadvantages including the impact that the various control systems would have on the nuclear electric propulsion system design. A recommendation for the reference control system is made, and a program for future work leading to an engineering model is described.

Environmental Effect on Evolutionary Cyclic Plasticity Material Parameters of 316 Stainless Steel: An Experimental & Material Modeling Approach

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

Mohanty, Subhasish; Soppet, William K.; Majumdar, Saurin

2014-09-20

This report provides an update on an earlier assessment of environmentally assisted fatigue for light water reactor (LWR) materials under extended service conditions. This report is a deliverable under the work package for environmentally assisted fatigue in the Light Water Reactor Sustainability (LWRS) program. The overall objective of this LWRS project is to assess the degradation by environmentally assisted cracking/fatigue of LWR materials such as various alloy base metals and their welds used in reactor coolant system piping. This effort is to support the Department of Energy LWRS program for developing tools to understand the aging/failure mechanism and to predictmore » the remaining life of LWR components for anticipated 60-80 year operation.« less

NGNP Data Management and Analysis System Modeling Capabilities

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

Cynthia D. Gentillon

2009-09-01

Projects for the very-high-temperature reactor (VHTR) program provide data in support of Nuclear Regulatory Commission licensing of the VHTR. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high temperature and high fluence environments. In addition, thermal-hydraulic experiments are conducted to validate codes used to assess reactor safety. The VHTR Program has established the NGNP Data Management and Analysis System (NDMAS) to ensure that VHTR data are (1) qualified for use, (2) stored in a readily accessible electronic form, and (3) analyzed to extract useful results. This document focuses on the thirdmore » NDMAS objective. It describes capabilities for displaying the data in meaningful ways and identifying relationships among the measured quantities that contribute to their understanding.« less

Tritium program at Chalk River Laboratories

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

Brown, R.M.; Workman, W.J.; Kotzer, T.G.

1993-01-01

Control of tritium dispersal within and around the research and power stations of the Canadian nuclear program has always been recognized as particularly important because of the high production of tritium in heavy-water-moderated reactors. At the Chalk River Labs, (CRL), two major research reactors have operated for more than 30 yr. Over the years, emissions have been from 300 to 700 TBq/yr (8 to 19 kCi/yr) to the atmosphere and from 100 to 200 TBq/yr (3 to 5 kCi/yr) to local water systems. This results in concentrations in atmospheric moisture of [approximately]600 Bq/[ell] water in the immediate reactor area, 80more » Bq/[ell] at the exclusion area boundary (7 km distant), and 50 Bq/[ell] at the nearest downwind community (12 km).« less

Autonomous Control Capabilities for Space Reactor Power Systems

NASA Astrophysics Data System (ADS)

Wood, Richard T.; Neal, John S.; Brittain, C. Ray; Mullens, James A.

2004-02-01

The National Aeronautics and Space Administration's (NASA's) Project Prometheus, the Nuclear Systems Program, is investigating a possible Jupiter Icy Moons Orbiter (JIMO) mission, which would conduct in-depth studies of three of the moons of Jupiter by using a space reactor power system (SRPS) to provide energy for propulsion and spacecraft power for more than a decade. Terrestrial nuclear power plants rely upon varying degrees of direct human control and interaction for operations and maintenance over a forty to sixty year lifetime. In contrast, an SRPS is intended to provide continuous, remote, unattended operation for up to fifteen years with no maintenance. Uncertainties, rare events, degradation, and communications delays with Earth are challenges that SRPS control must accommodate. Autonomous control is needed to address these challenges and optimize the reactor control design. In this paper, we describe an autonomous control concept for generic SRPS designs. The formulation of an autonomous control concept, which includes identification of high-level functional requirements and generation of a research and development plan for enabling technologies, is among the technical activities that are being conducted under the U.S. Department of Energy's Space Reactor Technology Program in support of the NASA's Project Prometheus. The findings from this program are intended to contribute to the successful realization of the JIMO mission.

A Discrepancy-Based Methodology for Nuclear Training Program Evaluation.

ERIC Educational Resources Information Center

Cantor, Jeffrey A.

1991-01-01

A three-phase comprehensive process for commercial nuclear power training program evaluation is presented. The discrepancy-based methodology was developed after the Three Mile Island nuclear reactor accident. It facilitates analysis of program components to identify discrepancies among program specifications, actual outcomes, and industry…

Evaluation of rotating-cylinder and piston-cylinder reactors for ground-based emulsion polymerization

NASA Technical Reports Server (NTRS)

Vanderhoff, J. W.; El-Aasser, M. S.

1987-01-01

The objectives of this program are to apply ground-based emulsion polymerization reactor technology to improve the production of: monodisperse latex particles for calibration standards, chromatographic separation column packing, and medical research; and commercial latexes such as those used for coatings, foams, and adhesives.

A PROGRAM OF RESEARCH ON MECHANICAL METALLURGY AS RELATED TO FUEL-ELEMENT FABRICATION. Summary Report, January 1-September 30, 1961

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

Trozera, T.A.; White, J.L.; Chambers, R.H.

Research progress on mechanical metallurgy of reactor materials is reported in three sections: deformation characteristics of reactor materials, stored energy of cold work, and microplastic propenties and mechanical relaxation spectra of very pure refractory bcc metals. (M.C.G.)

Hydrogen and water reactor safety: proceedings

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

Not Available

1982-01-01

Separate abstracts were prepared for papers presented in the following areas of interest: 1) hydrogen research programs; 2) hydrogen behavior during light water reactor accidents; 3) combustible gas generation; 4) hydrogen transport and mixing; 5) combustion modeling and experiments; 6) accelerated flames and detonations; 7) combustion mitigation and control; and 8) equipment survivability.

TECHNOLOGY EVALUATION REPORT: SITE PROGRAM DEMON- STRATION TEST - HORSEHEAD RESOURCE DEVELOPMENT COMPANY, INC. - FLAME REACTOR TECHNOLOGY - MONACA, PENNSYLVANIA

EPA Science Inventory

A SITE demonstration of the Horsehead Resource Development (HRD) Company, Inc. Flame Reactor Technology was conducted in March 1991 at the HRD facility in Monaca, Pennsylvania. or this demonstration, secondary lead smelter soda slag was treated to produce a potentially recyclable...

10 CFR Appendix H to Part 50 - Reactor Vessel Material Surveillance Program Requirements

Code of Federal Regulations, 2010 CFR

2010-01-01

... light water nuclear power reactors which result from exposure of these materials to neutron irradiation... the beltline region so that the specimen irradiation history duplicates, to the extent practicable... insertion of replacement capsules. Accelerated irradiation capsules may be used in addition to the required...

10 CFR Appendix H to Part 50 - Reactor Vessel Material Surveillance Program Requirements

Code of Federal Regulations, 2014 CFR

2014-01-01

... light water nuclear power reactors which result from exposure of these materials to neutron irradiation... the beltline region so that the specimen irradiation history duplicates, to the extent practicable... insertion of replacement capsules. Accelerated irradiation capsules may be used in addition to the required...

10 CFR Appendix H to Part 50 - Reactor Vessel Material Surveillance Program Requirements

Code of Federal Regulations, 2012 CFR

2012-01-01

... light water nuclear power reactors which result from exposure of these materials to neutron irradiation... the beltline region so that the specimen irradiation history duplicates, to the extent practicable... insertion of replacement capsules. Accelerated irradiation capsules may be used in addition to the required...

10 CFR Appendix H to Part 50 - Reactor Vessel Material Surveillance Program Requirements

Code of Federal Regulations, 2011 CFR

2011-01-01

... light water nuclear power reactors which result from exposure of these materials to neutron irradiation... the beltline region so that the specimen irradiation history duplicates, to the extent practicable... insertion of replacement capsules. Accelerated irradiation capsules may be used in addition to the required...

10 CFR Appendix H to Part 50 - Reactor Vessel Material Surveillance Program Requirements

Code of Federal Regulations, 2013 CFR

2013-01-01

... light water nuclear power reactors which result from exposure of these materials to neutron irradiation... the beltline region so that the specimen irradiation history duplicates, to the extent practicable... insertion of replacement capsules. Accelerated irradiation capsules may be used in addition to the required...

10 CFR 2.318 - Commencement and termination of jurisdiction of presiding officer.

Code of Federal Regulations, 2014 CFR

2014-01-01

..., whichever is earliest. (b) The Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, the Director, Office of Federal and State Materials and Environmental Management Programs, or the... officer. 2.318 Section 2.318 Energy NUCLEAR REGULATORY COMMISSION AGENCY RULES OF PRACTICE AND PROCEDURE...

10 CFR 2.318 - Commencement and termination of jurisdiction of presiding officer.

Code of Federal Regulations, 2013 CFR

2013-01-01

..., whichever is earliest. (b) The Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, the Director, Office of Federal and State Materials and Environmental Management Programs, or the... officer. 2.318 Section 2.318 Energy NUCLEAR REGULATORY COMMISSION AGENCY RULES OF PRACTICE AND PROCEDURE...

Current status of nuclear engineering education

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

Palladino, N.J.

1975-09-01

The 65 colleges and universities offering undergraduate degrees in nuclear engineering and the 15 schools offering strong nuclear engineering options are, in general, doing a good job to meet the current spectrum of job opportunities. But, nuclear engineering programs are not producing enough graduates to meet growing demands. They currently receive little aid and support from their customers --industry and government--in the form of scholarships, grants, faculty research support, student thesis and project support, or student summer jobs. There is not enough interaction between industry and universities. Most nuclear engineering programs are geared too closely to the technology of themore » present family of reactors and too little to the future breeder reactors and controlled thermonuclear reactors. In addition, nuclear engineering programs attract too few women and members of minority ethnic groups. Further study of the reasons for this fact is needed so that effective corrective action can be taken. Faculty in nuclear engineering programs should assume greater initiative to provide attractive and objective nuclear energy electives for technical and nontechnical students in other disciplines to improve their technical understanding of the safety and environmental issues involved. More aggressive and persistent efforts must be made by nuclear engineering schools to obtain industry support and involvement in their programs. (auth)« less

Renewing Liquid Fueled Molten Salt Reactor Research and Development

NASA Astrophysics Data System (ADS)

Towell, Rusty; NEXT Lab Team

2016-09-01

Globally there is a desperate need for affordable, safe, and clean energy on demand. More than anything else, this would raise the living conditions of those in poverty around the world. An advanced reactor that utilizes liquid fuel and molten salts is capable of meeting these needs. Although, this technology was demonstrated in the Molten Salt Reactor Experiment (MSRE) at ORNL in the 60's, little progress has been made since the program was cancelled over 40 years ago. A new research effort has been initiated to advance the technical readiness level of key reactor components. This presentation will explain the motivation and initial steps for this new research initiative.

Evaluation of nuclear-facility decommissioning projects. Summary report: Ames Laboratory Research Reactor

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

Link, B.W.; Miller, R.L.

1983-07-01

This document summarizes the available information concerning the decommissioning of the Ames Laboratory Research Reactor (ALRR), a five-megawatt heavy water moderated and cooled research reactor. The data were placed in a computerized information retrieval/manipulation system which permits its future utilization for purposes of comparative analysis. This information is presented both in detail in its computer output form and also as a manually assembled summarization which highlights the more important aspects of the decommissioning program. Some comparative information with reference to generic decommissioning data extracted from NUREG/CR 1756, Technology, Safety and Costs of Decommissioning Nuclear Research and Test Reactors, is included.

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

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

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

2014-10-01

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

The WSTIAC Quarterly. Volume 9, Number 3

DTIC Science & Technology

2010-01-25

program .[8] THE THORIUM FUEL CYCLE AND LFTR POWER PLANT The thorium fuel cycle is based on a series of neutron absorp- tion and beta decay processes...the fig- ure is a graphite matrix moderated MSR reactor with fuel salt mixture (ThF4-U233F4) being circulated by a pump through the core and to a...the core as purified salt. As one of the unique safety features, a melt-plug at the reactor bottom would permit the reactor fluid fuel to be drained

Design Study of a Modular Gas-Cooled, Closed-Brayton Cycle Reactor for Marine Use

DTIC Science & Technology

1989-06-01

materials in the core and surroundings. To investigate this design point in the marine variant I developed the program HEAT.BAS to perform a one-dimensional...helium as the working fluid. The core is a graphite moderated, epithermal spectrum reactor, using TRISO fuel particles in extruded graphite fuel elements...The fuel is highly enriched U2315 . The containment is shaped in an inverted ’T’ with two sections. The upper section contains the reactor core

Low Energy Neutrino Physics at the Kuo-Sheng Reactor Laboratory in Taiwan

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

Lin, S.-T.

2006-11-17

A laboratory has been constructed by the TEXONO Collaboration at the Kuo-Sheng Reactor Power Plant in Taiwan to study low energy neutrino physics. A limit on the neutrino magnetic moment of {mu}{nu}({nu}-bare) < 7.2 x 10-11 {mu}B at 90% confidence level has been achieved from measurements with a high-purity germanium detector, as well as the electron neutrinos ({nu}{sub e}) produced from nuclear power reactors has been studied. Other research program at Kuo-Sheng are surveyed.

Current and anticipated uses of thermal hydraulic codes at the Japan Atomic Energy Research Institute

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

Akimoto, Hajime; Kukita; Ohnuki, Akira

1997-07-01

The Japan Atomic Energy Research Institute (JAERI) is conducting several research programs related to thermal-hydraulic and neutronic behavior of light water reactors (LWRs). These include LWR safety research projects, which are conducted in accordance with the Nuclear Safety Commission`s research plan, and reactor engineering projects for the development of innovative reactor designs or core/fuel designs. Thermal-hydraulic and neutronic codes are used for various purposes including experimental analysis, nuclear power plant (NPP) safety analysis, and design assessment.

Next Generation Nuclear Plant Methods Research and Development Technical Program Plan -- PLN-2498

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

Richard R. Schultz; Abderrafi M. Ougouag; David W. Nigg

2008-09-01

One of the great challenges of designing and licensing the Very High Temperature Reactor (VHTR) is to confirm that the intended VHTR analysis tools can be used confidently to make decisions and to assure all that the reactor systems are safe and meet the performance objectives of the Generation IV Program. The research and development (R&D) projects defined in the Next Generation Nuclear Plant (NGNP) Design Methods Development and Validation Program will ensure that the tools used to perform the required calculations and analyses can be trusted. The Methods R&D tasks are designed to ensure that the calculational envelope ofmore » the tools used to analyze the VHTR reactor systems encompasses, or is larger than, the operational and transient envelope of the VHTR itself. The Methods R&D focuses on the development of tools to assess the neutronic and thermal fluid behavior of the plant. The fuel behavior and fission product transport models are discussed in the Advanced Gas Reactor (AGR) program plan. Various stress analysis and mechanical design tools will also need to be developed and validated and will ultimately also be included in the Methods R&D Program Plan. The calculational envelope of the neutronics and thermal-fluids software tools intended to be used on the NGNP is defined by the scenarios and phenomena that these tools can calculate with confidence. The software tools can only be used confidently when the results they produce have been shown to be in reasonable agreement with first-principle results, thought-problems, and data that describe the “highly ranked” phenomena inherent in all operational conditions and important accident scenarios for the VHTR.« less

Next Generation Nuclear Plant Methods Technical Program Plan

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

Richard R. Schultz; Abderrafi M. Ougouag; David W. Nigg

2010-12-01

One of the great challenges of designing and licensing the Very High Temperature Reactor (VHTR) is to confirm that the intended VHTR analysis tools can be used confidently to make decisions and to assure all that the reactor systems are safe and meet the performance objectives of the Generation IV Program. The research and development (R&D) projects defined in the Next Generation Nuclear Plant (NGNP) Design Methods Development and Validation Program will ensure that the tools used to perform the required calculations and analyses can be trusted. The Methods R&D tasks are designed to ensure that the calculational envelope ofmore » the tools used to analyze the VHTR reactor systems encompasses, or is larger than, the operational and transient envelope of the VHTR itself. The Methods R&D focuses on the development of tools to assess the neutronic and thermal fluid behavior of the plant. The fuel behavior and fission product transport models are discussed in the Advanced Gas Reactor (AGR) program plan. Various stress analysis and mechanical design tools will also need to be developed and validated and will ultimately also be included in the Methods R&D Program Plan. The calculational envelope of the neutronics and thermal-fluids software tools intended to be used on the NGNP is defined by the scenarios and phenomena that these tools can calculate with confidence. The software tools can only be used confidently when the results they produce have been shown to be in reasonable agreement with first-principle results, thought-problems, and data that describe the “highly ranked” phenomena inherent in all operational conditions and important accident scenarios for the VHTR.« less

Next Generation Nuclear Plant Methods Technical Program Plan -- PLN-2498

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

Richard R. Schultz; Abderrafi M. Ougouag; David W. Nigg

2010-09-01

One of the great challenges of designing and licensing the Very High Temperature Reactor (VHTR) is to confirm that the intended VHTR analysis tools can be used confidently to make decisions and to assure all that the reactor systems are safe and meet the performance objectives of the Generation IV Program. The research and development (R&D) projects defined in the Next Generation Nuclear Plant (NGNP) Design Methods Development and Validation Program will ensure that the tools used to perform the required calculations and analyses can be trusted. The Methods R&D tasks are designed to ensure that the calculational envelope ofmore » the tools used to analyze the VHTR reactor systems encompasses, or is larger than, the operational and transient envelope of the VHTR itself. The Methods R&D focuses on the development of tools to assess the neutronic and thermal fluid behavior of the plant. The fuel behavior and fission product transport models are discussed in the Advanced Gas Reactor (AGR) program plan. Various stress analysis and mechanical design tools will also need to be developed and validated and will ultimately also be included in the Methods R&D Program Plan. The calculational envelope of the neutronics and thermal-fluids software tools intended to be used on the NGNP is defined by the scenarios and phenomena that these tools can calculate with confidence. The software tools can only be used confidently when the results they produce have been shown to be in reasonable agreement with first-principle results, thought-problems, and data that describe the “highly ranked” phenomena inherent in all operational conditions and important accident scenarios for the VHTR.« less

Gas-cooled reactor programs. High-temperature gas-cooled reactor technology development program. Annual progress report, December 31, 1983

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

Kasten, P.R.; Rittenhouse, P.L.; Bartine, D.E.

1984-06-01

ORNL continues to make significant contributions to the national program. In the HTR fuels area, we are providing detailed statistical information on the fission product retention performance of irradiated fuel. Our studies are also providing basic data on the mechanical, physical, and chemical behavior of HTR materials, including metals, ceramics, graphite, and concrete. The ORNL has an important role in the development of improved HTR graphites and in the specification of criteria that need to be met by commercial products. We are also developing improved reactor physics design methods. Our work in component development and testing centers in the Componentmore » Flow Test Loop (CFTL), which is being used to evaluate the performance of the HTR core support structure. Other work includes experimental evaluation of the shielding effectiveness of the lower portions of an HTR core. This evaluation is being performed at the ORNL Tower Shielding Facility. Researchers at ORNL are developing welding techniques for attaching steam generator tubing to the tubesheets and are testing ceramic pads on which the core posts rest. They are also performing extensive testing of aggregate materials obtained from potential HTR site areas for possible use in prestressed concrete reactor vessels. During the past year we continued to serve as a peer reviewer of small modular reactor designs being developed by GA and GE with balance-of-plant layouts being developed by Bechtel Group, Inc. We have also evaluated the national need for developing HTRs with emphasis on the longer term applications of the HTRs to fossil conversion processes.« less

Technicians Manufacture a Nozzle for the Kiwi B-1-B Engine

NASA Image and Video Library

1964-05-21

Technicians manufacture a nozzle for the Kiwi B-1-B nuclear rocket engine in the Fabrication Shop’s vacuum oven at the National Aeronautics and Space Administration (NASA) Lewis Research Center. The Nuclear Engine for Rocket Vehicle Applications (NERVA) was a joint NASA and Atomic Energy Commission (AEC) endeavor to develop a nuclear-powered rocket for both long-range missions to Mars and as a possible upper-stage for the Apollo Program. The early portion of the program consisted of basic reactor and fuel system research. This was followed by a series of Kiwi reactors built to test basic nuclear rocket principles in a non-flying nuclear engine. The next phase, NERVA, would create an entire flyable engine. The final phase of the program, called Reactor-In-Flight-Test, would be an actual launch test. The AEC was responsible for designing the nuclear reactor and overall engine. NASA Lewis was responsible for developing the liquid-hydrogen fuel system. The turbopump, which pumped the fuels from the storage tanks to the engine, was the primary tool for restarting the engine. The NERVA had to be able to restart in space on its own using a safe preprogrammed startup system. Lewis researchers endeavored to design and test this system. This non-nuclear Kiwi engine, seen here, was being prepared for tests at Lewis’ High Energy Rocket Engine Research Facility (B-1) located at Plum Brook Station. The tests were designed to start an unfueled Kiwi B-1-B reactor and its Aerojet Mark IX turbopump without any external power.

Goals of thermionic program for space power

NASA Technical Reports Server (NTRS)

English, R. E.

1981-01-01

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

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

NASA Astrophysics Data System (ADS)

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

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

Space station program phase B definition: Nuclear reactor-powered space station cost and schedules

NASA Technical Reports Server (NTRS)

1971-01-01

Tabulated data are presented on the costs, schedules, and technical characteristics for the space station phases C and D program. The work breakdown structure, schedule data, program ground rules, program costs, cost-estimating rationale, funding schedules, and supporting data are included.

Nuclear Education and Training Programs of Potential Interest to Utilities.

ERIC Educational Resources Information Center

Atomic Energy Commission, Washington, DC.

This compilation of education and training programs related to nuclear applications in electric power generation covers programs conducted by nuclear reactor vendors, public utilities, universities, technical institutes, and community colleges, which were available in December 1968. Several training-program consultant services are also included.…

Needs of Accurate Prompt and Delayed γ-spectrum and Multiplicity for Nuclear Reactor Designs

NASA Astrophysics Data System (ADS)

Rimpault, G.; Bernard, D.; Blanchet, D.; Vaglio-Gaudard, C.; Ravaux, S.; Santamarina, A.

The local energy photon deposit must be accounted accurately for Gen-IV fast reactors, advanced light-water nuclear reactors (Gen-III+) and the new experimental Jules Horowitz Reactor (JHR). The γ energy accounts for about 10% of the total energy released in the core of a thermal or fast reactor. The γ-energy release is much greater in the core of the reactor than in its structural sub-assemblies (such as reflector, control rod followers, dummy sub-assemblies). However, because of the propagation of γ from the core regions to the neighboring fuel-free assemblies, the contribution of γ energy to the total heating can be dominant. For reasons related to their performance, power reactors require a 7.5% (1σ) uncertainty for the energy deposition in non-fuelled zones. For the JHR material-testing reactor, a 5% (1 s) uncertainty is required in experimental positions. In order to verify the adequacy of the calculation of γ-heating, TLD and γ-fission chambers were used to derive the experimental heating values. Experimental programs were and are still conducted in different Cadarache facilities such as MASURCA (for SFR), MINERVE and EOLE (for JHR and Gen-III+ reactors). The comparison of calculated and measured γ-heating values shows an underestimation in all experimental programs indicating that for the most γ-production data from 239Pu in current nuclear-data libraries is highly suspicious.The first evaluation priority is for prompt γ-multiplicity for U and Pu fission but similar values for otheractinides such as Pu and U are also required. The nuclear data library JEFF3.1.1 contains most of the photon production data. However, there are some nuclei for which there are missing or erroneous data which need to be completed or modified. A review of the data available shows a lack of measurements for conducting serious evaluation efforts. New measurements are needed to guide new evaluation efforts which benefit from consolidated modeling techniques.

Consolidated fuel reprocessing program

NASA Astrophysics Data System (ADS)

1985-04-01

A survey of electrochemical methods applications in fuel reprocessing was completed. A dummy fuel assembly shroud was cut using the remotely operated laser disassembly equipment. Operations and engineering efforts have continued to correct equipment operating, software, and procedural problems experienced during the previous uranium compaigns. Fuel cycle options were examined for the liquid metal reactor fuel cycle. In high temperature gas cooled reactor spent fuel studies, preconceptual designs were completed for the concrete storage cask and open field drywell storage concept. These and other tasks operating under the consolidated fuel reprocessing program are examined.

Reactor physics teaching and research in the Swiss nuclear engineering master

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

Chawla, R.; Paul Scherrer Inst., CH-5232 Villigen PSI

Since 2008, a Master of Science program in Nuclear Engineering (NE) has been running in Switzerland, thanks to the combined efforts of the country's key players in nuclear teaching and research, viz. the Swiss Federal Inst.s of Technology at Lausanne (EPFL) and at Zurich (ETHZ), the Paul Scherrer Inst. (PSI) at Villigen and the Swiss Nuclear Utilities (Swissnuclear). The present paper, while outlining the academic program as a whole, lays emphasis on the reactor physics teaching and research training accorded to the students in the framework of the developed curriculum. (authors)

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

Mulder, R.U.; Benneche, P.E.; Hosticka, B.

The University of Virginia Reactor Facility is an integral part of the Department of Nuclear Engineering and Engineering Physics (to become the Department of Mechanical, Aerospace and Nuclear Engineering on July 1, 1992). As such, it is effectively used to support educational programs in engineering and science at the University of Virginia as well as those at other area colleges and universities. The expansion of support to educational programs in the mid-east region is a major objective. To assist in meeting this objective, the University of Virginia has been supported under the US Department of Energy (DOE) Reactor Sharing Programmore » since 1978. Due to the success of the program, this proposal requests continued DOE support through August 1993.« less

Laboratory instrumentation modernization at the WPI Nuclear Reactor Facility

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

Not Available

1995-01-01

With partial funding from the Department of Energy (DOE) University Reactor Instrumentation Program several laboratory instruments utilized by students and researchers at the WPI Nuclear Reactor Facility have been upgraded or replaced. Designed and built by General Electric in 1959, the open pool nuclear training reactor at WPI was one of the first such facilities in the nation located on a university campus. Devoted to undergraduate use, the reactor and its related facilities have been since used to train two generations of nuclear engineers and scientists for the nuclear industry. The low power output of the reactor and an ergonomicmore » facility design make it an ideal tool for undergraduate nuclear engineering education and other training. The reactor, its control system, and the associate laboratory equipment are all located in the same room. Over the years, several important milestones have taken place at the WPI reactor. In 1969, the reactor power level was upgraded from 1 kW to 10 kW. The reactor`s Nuclear Regulatory Commission operating license was renewed for 20 years in 1983. In 1988, under DOE Grant No. DE-FG07-86ER75271, the reactor was converted to low-enriched uranium fuel. In 1992, again with partial funding from DOE (Grant No. DE-FG02-90ER12982), the original control console was replaced.« less

CHAP-2 heat-transfer analysis of the Fort St. Vrain reactor core

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

Kotas, J.F.; Stroh, K.R.

1983-01-01

The Los Alamos National Laboratory is developing the Composite High-Temperature Gas-Cooled Reactor Analysis Program (CHAP) to provide advanced best-estimate predictions of postulated accidents in gas-cooled reactor plants. The CHAP-2 reactor-core model uses the finite-element method to initialize a two-dimensional temperature map of the Fort St. Vrain (FSV) core and its top and bottom reflectors. The code generates a finite-element mesh, initializes noding and boundary conditions, and solves the nonlinear Laplace heat equation using temperature-dependent thermal conductivities, variable coolant-channel-convection heat-transfer coefficients, and specified internal fuel and moderator heat-generation rates. This paper discusses this method and analyzes an FSV reactor-core accident thatmore » simulates a control-rod withdrawal at full power.« less

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

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

Fitriyani, Dian; Su'ud, Zaki

2010-06-22

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

Development of Cross Section Library and Application Programming Interface (API)

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

Lee, C. H.; Marin-Lafleche, A.; Smith, M. A.

2014-04-09

The goal of NEAMS neutronics is to develop a high-fidelity deterministic neutron transport code termed PROTEUS for use on all reactor types of interest, but focused primarily on sodium-cooled fast reactors. While PROTEUS-SN has demonstrated good accuracy for homogeneous fast reactor problems and partially heterogeneous fast reactor problems, the simulation results were not satisfactory when applied on fully heterogeneous thermal problems like the Advanced Test Reactor (ATR). This is mainly attributed to the quality of cross section data for heterogeneous geometries since the conventional cross section generation approach does not work accurately for such irregular and complex geometries. Therefore, onemore » of the NEAMS neutronics tasks since FY12 has been the development of a procedure to generate appropriate cross sections for a heterogeneous geometry core.« less

Loss-of-Flow and Loss-of-Pressure Simulations of the BR2 Research Reactor with HEU and LEU Fuel

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

Licht, J.; Bergeron, A.; Dionne, B.

2016-01-01

Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The reactor core of BR2 is located inside a pressure vessel that contains 79 channels in a hyperboloid configuration. The core configuration is highly variable as each channel can contain a fuel assembly, a control or regulating rod, an experimentalmore » device, or a beryllium or aluminum plug. Because of this variability, a representative core configuration, based on current reactor use, has been defined for the fuel conversion analyses. The code RELAP5/Mod 3.3 was used to perform the transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. The input model has been modernized relative to that historically used at BR2 taking into account the best modeling practices developed by Argonne National Laboratory (ANL) and BR2 engineers.« less

Control console replacement at the WPI Reactor. [Final report

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

Not Available

1992-12-31

With partial funding from the Department of Energy (DOE) University Reactor Instrumentation Upgrade Program (DOE Grant No. DE-FG02-90ER12982), the original control console at the Worcester Polytechnic Institute (WPI) Reactor has been replaced with a modern system. The new console maintains the original design bases and functionality while utilizing current technology. An advanced remote monitoring system has been added to augment the educational capabilities of the reactor. Designed and built by General Electric in 1959, the open pool nuclear training reactor at WPI was one of the first such facilities in the nation located on a university campus. Devoted to undergraduatemore » use, the reactor and its related facilities have been since used to train two generations of nuclear engineers and scientists for the nuclear industry. The reactor power level was upgraded from 1 to 10 kill in 1969, and its operating license was renewed for 20 years in 1983. In 1988, the reactor was converted to low enriched uranium. The low power output of the reactor and ergonomic facility design make it an ideal tool for undergraduate nuclear engineering education and other training.« less

Co-Production of Electricity and Hydrogen Using a Novel Iron-based Catalyst

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

Hilaly, Ahmad; Georgas, Adam; Leboreiro, Jose

2011-09-30

The primary objective of this project was to develop a hydrogen production technology for gasification applications based on a circulating fluid-bed reactor and an attrition resistant iron catalyst. The work towards achieving this objective consisted of three key activities: Development of an iron-based catalyst suitable for a circulating fluid-bed reactor; Design, construction, and operation of a bench-scale circulating fluid-bed reactor system for hydrogen production; Techno-economic analysis of the steam-iron and the pressure swing adsorption hydrogen production processes. This report describes the work completed in each of these activities during this project. The catalyst development and testing program prepared and iron-basedmore » catalysts using different support and promoters to identify catalysts that had sufficient activity for cyclic reduction with syngas and steam oxidation and attrition resistance to enable use in a circulating fluid-bed reactor system. The best performing catalyst from this catalyst development program was produced by a commercial catalyst toll manufacturer to support the bench-scale testing activities. The reactor testing systems used during material development evaluated catalysts in a single fluid-bed reactor by cycling between reduction with syngas and oxidation with steam. The prototype SIP reactor system (PSRS) consisted of two circulating fluid-bed reactors with the iron catalyst being transferred between the two reactors. This design enabled demonstration of the technical feasibility of the combination of the circulating fluid-bed reactor system and the iron-based catalyst for commercial hydrogen production. The specific activities associated with this bench-scale circulating fluid-bed reactor systems that were completed in this project included design, construction, commissioning, and operation. The experimental portion of this project focused on technical demonstration of the performance of an iron-based catalyst and a circulating fluid-bed reactor system for hydrogen production. Although a technology can be technically feasible, successful commercial deployment also requires that a technology offer an economic advantage over existing commercial technologies. To effective estimate the economics of this steam-iron process, a techno-economic analysis of this steam iron process and a commercial pressure swing adsorption process were completed. The results from this analysis described in this report show the economic potential of the steam iron process for integration with a gasification plant for coproduction of hydrogen and electricity.« less

ATOMIC PHYSICS, AN AUTOINSTRUCTIONAL PROGRAM, VOLUME 4, SUPPLEMENT.

ERIC Educational Resources Information Center

DETERLINE, WILLIAM A.; KLAUS, DAVID J.

THE AUTOINSTRUCTIONAL MATERIALS IN THIS TEXT WERE PREPARED FOR USE IN AN EXPERIMENTAL STUDY, OFFERING SELF-TUTORING MATERIAL FOR LEARNING ATOMIC PHYSICS. THE TOPICS COVERED ARE (1) RADIATION USES AND NUCLEAR FISSION, (2) NUCLEAR REACTORS, (3) ENERGY FROM NUCLEAR REACTORS, (4) NUCLEAR EXPLOSIONS AND FUSION, (5) A COMPREHENSIVE REVIEW, AND (6) A…

Nuclear Propulsion for Space, Understanding the Atom Series.

ERIC Educational Resources Information Center

Corliss, William R.; Schwenk, Francis C.

The operation of nuclear rockets with respect both to rocket theory and to various fuels is described. The development of nuclear reactors for use in nuclear rocket systems is provided, with the Kiwi and NERVA programs highlighted. The theory of fuel element and reactor construction and operation is explained with particular reference to rocket…

Reactors Save Energy, Costs for Hydrogen Production

NASA Technical Reports Server (NTRS)

2014-01-01

While examining fuel-reforming technology for fuel cells onboard aircraft, Glenn Research Center partnered with Garrettsville, Ohio-based Catacel Corporation through the Glenn Alliance Technology Exchange program and a Space Act Agreement. Catacel developed a stackable structural reactor that is now employed for commercial hydrogen production and results in energy savings of about 20 percent.

78 FR 37591 - Entergy Nuclear Operations, Inc., Entergy Nuclear Indian Point Unit 2, LLC, Issuance of Director...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-21

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0284; Docket No. 50-247; License No. DPR-26] Entergy Nuclear Operations, Inc., Entergy Nuclear Indian Point Unit 2, LLC, Issuance of Director's Decision Notice is hereby given that the Deputy Director, Reactor Safety Programs, Office of Nuclear Reactor...

First-wall structural analysis of the self-cooled water blanket concept

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

O'Brien, D.A.; Steiner, D.; Embrechts, M.J.

1986-01-01

A novel blanket concept recently proposed utilizes water with small amounts of dissolved lithium compound as both coolant and breeder. The inherent simplicity of this idea should result in an attractive breeding blanket for fusion reactors. In addition, the available base of relevant information accumulated through water-cooled fission reactor programs should greatly facilitate the R and D effort required to validate this concept. First-wall and blanket designs have been developed first for the tandem mirror reactor (TMR) due to the obvious advantages of this geometry. First-wall and blanket designs will also be developed for toroidal reactors. A simple plate designmore » with coolant tubes welded on the back (side away from plasma) was chosen as the first wall for the TMR application. Dimensions and materials were chosen to minimize temperature differences and thermal stresses. A finite element code (STRAW), originally developed for the analysis of core components subjected to high-pressure transients in the fast breeder program, was utilized to evaluate stresses in the first wall.« less

LWRS ATR Irradiation Testing Readiness Status

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

Kristine Barrett

2012-09-01

The Light Water Reactor Sustainability (LWRS) Program was established by the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) to develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors. The LWRS Program is divided into four R&D Pathways: (1) Materials Aging and Degradation; (2) Advanced Light Water Reactor Nuclear Fuels; (3) Advanced Instrumentation, Information and Control Systems; and (4) Risk-Informed Safety Margin Characterization. This report describes an irradiation testing readiness analysis in preparation of LWRS experiments for irradiation testing at the Idaho National Laboratory (INL) Advanced Testmore » Reactor (ATR) under Pathway (2). The focus of the Advanced LWR Nuclear Fuels Pathway is to improve the scientific knowledge basis for understanding and predicting fundamental performance of advanced nuclear fuel and cladding in nuclear power plants during both nominal and off-nominal conditions. This information will be applied in the design and development of high-performance, high burn-up fuels with improved safety, cladding integrity, and improved nuclear fuel cycle economics« less

Space Nuclear Thermal Propulsion (SNTP) Air Force facility

NASA Technical Reports Server (NTRS)

Beck, David F.

1993-01-01

The Space Nuclear Thermal Propulsion (SNTP) Program is an initiative within the US Air Force to acquire and validate advanced technologies that could be used to sustain superior capabilities in the area or space nuclear propulsion. The SNTP Program has a specific objective of demonstrating the feasibility of the particle bed reactor (PBR) concept. The term PIPET refers to a project within the SNTP Program responsible for the design, development, construction, and operation of a test reactor facility, including all support systems, that is intended to resolve program technology issues and test goals. A nuclear test facility has been designed that meets SNTP Facility requirements. The design approach taken to meet SNTP requirements has resulted in a nuclear test facility that should encompass a wide range of nuclear thermal propulsion (NTP) test requirements that may be generated within other programs. The SNTP PIPET project is actively working with DOE and NASA to assess this possibility.

ZPR-6 assembly 7 high {sup 240} PU core : a cylindrical assemby with mixed (PU, U)-oxide fuel and a central high {sup 240} PU zone.

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

Lell, R. M.; Schaefer, R. W.; McKnight, R. D.

Over a period of 30 years more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited to form the basis for criticality safety benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactormore » physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. The term 'benchmark' in a ZPR program connotes a particularly simple loading aimed at gaining basic reactor physics insight, as opposed to studying a reactor design. In fact, the ZPR-6/7 Benchmark Assembly (Reference 1) had a very simple core unit cell assembled from plates of depleted uranium, sodium, iron oxide, U3O8, and plutonium. The ZPR-6/7 core cell-average composition is typical of the interior region of liquid-metal fast breeder reactors (LMFBRs) of the era. It was one part of the Demonstration Reactor Benchmark Program,a which provided integral experiments characterizing the important features of demonstration-size LMFBRs. As a benchmark, ZPR-6/7 was devoid of many 'real' reactor features, such as simulated control rods and multiple enrichment zones, in its reference form. Those kinds of features were investigated experimentally in variants of the reference ZPR-6/7 or in other critical assemblies in the Demonstration Reactor Benchmark Program.« less

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

Melber, B.D.; Saari, L.M.; White, A.S.

This report assesses the job-relatedness of specialized educational programs for licensed nuclear reactor operators. The approach used involved systematically comparing the curriculum of specialized educational programs for college credit, to academic knowledge identified as necessary for carrying out the jobs of licenses reactor operators. A sample of eight programs, including A.S. degree, B.S. degree, and coursework programs were studied. Subject matter experts in the field of nuclear operations curriculum and training determined the extent to which individual program curricula covered the identified job-related academic knowledge. The major conclusions of the report are: There is a great deal of variation amongmore » individual programs, ranging from coverage of 15% to 65% of the job-related academic knowledge. Four schools cover at least half, and four schools cover less than one-third of this knowledge content; There is no systematic difference in the job-relatedness of the different types of specialized educational programs, A.S. degree, B.S. degree, and coursework; and Traditional B.S. degree programs in nuclear engineering cover as much job-related knowledge (about one-half of this knowledge content) as most of the specialized educational programs.« less

AGR-1 Compact 1-3-1 Post-Irradiation Examination Results

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

Demkowicz, Paul Andrew

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform the requisite research and development on tristructural isotropic (TRISO) coated particle fuel to support deployment of a high-temperature gas-cooled reactor (HTGR). The work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel program. The overarching program goal is to provide a baseline fuel qualification data set to support licensing and operation of an HTGR. To achieve these goals, the program includes the elements of fuel fabrication, irradiation, post-irradiation examination (PIE) and safety testing, fuel performance modeling, and fission product transport (INL 2015). A seriesmore » of fuel irradiation experiments is being planned and conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). These experiments will provide data on fuel performance under irradiation, support fuel process development, qualify the fuel for normal operating conditions, provide irradiated fuel for safety testing, and support the development of fuel performance and fission product transport models. The first of these irradiation tests, designated AGR-1, began in the ATR in December 2006 and ended in November 2009. This experiment was conducted primarily to act as a shakedown test of the multicapsule test train design and provide early data on fuel performance for use in fuel fabrication process development. It also provided samples for post-irradiation safety testing, where fission product retention of the fuel at high temperatures will be experimentally measured. The capsule design and details of the AGR-1 experiment have been presented previously (Grover, Petti, and Maki 2010, Maki 2009).« less

AGR-1 Compact 5-3-1 Post-Irradiation Examination Results

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

Demkowicz, Paul; Harp, Jason; Winston, Phil

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform the requisite research and development on tristructural isotropic (TRISO) coated particle fuel to support deployment of a high-temperature gas-cooled reactor (HTGR). The work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel program. The overarching program goal is to provide a baseline fuel qualification data set to support licensing and operation of an HTGR. To achieve these goals, the program includes the elements of fuel fabrication, irradiation, post-irradiation examination (PIE) and safety testing, fuel performance, and fission product transport (INL 2015). A series ofmore » fuel irradiation experiments is being planned and conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). These experiments will provide data on fuel performance under irradiation, support fuel process development, qualify the fuel for normal operating conditions, provide irradiated fuel for safety testing, and support the development of fuel performance and fission product transport models. The first of these irradiation tests, designated AGR-1, began in the ATR in December 2006 and ended in November 2009. This experiment was conducted primarily to act as a shakedown test of the multicapsule test train design and provide early data on fuel performance for use in fuel fabrication process development. It also provided samples for post-irradiation safety testing, where fission product retention of the fuel at high temperatures will be experimentally measured. The capsule design and details of the AGR-1 experiment have been presented previously.« less

Interface design of VSOP'94 computer code for safety analysis

NASA Astrophysics Data System (ADS)

Natsir, Khairina; Yazid, Putranto Ilham; Andiwijayakusuma, D.; Wahanani, Nursinta Adi

2014-09-01

Today, most software applications, also in the nuclear field, come with a graphical user interface. VSOP'94 (Very Superior Old Program), was designed to simplify the process of performing reactor simulation. VSOP is a integrated code system to simulate the life history of a nuclear reactor that is devoted in education and research. One advantage of VSOP program is its ability to calculate the neutron spectrum estimation, fuel cycle, 2-D diffusion, resonance integral, estimation of reactors fuel costs, and integrated thermal hydraulics. VSOP also can be used to comparative studies and simulation of reactor safety. However, existing VSOP is a conventional program, which was developed using Fortran 65 and have several problems in using it, for example, it is only operated on Dec Alpha mainframe platforms and provide text-based output, difficult to use, especially in data preparation and interpretation of results. We develop a GUI-VSOP, which is an interface program to facilitate the preparation of data, run the VSOP code and read the results in a more user friendly way and useable on the Personal 'Computer (PC). Modifications include the development of interfaces on preprocessing, processing and postprocessing. GUI-based interface for preprocessing aims to provide a convenience way in preparing data. Processing interface is intended to provide convenience in configuring input files and libraries and do compiling VSOP code. Postprocessing interface designed to visualized the VSOP output in table and graphic forms. GUI-VSOP expected to be useful to simplify and speed up the process and analysis of safety aspects.

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

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

Weigl, M.

2008-07-01

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

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

Boyack, B.E.

The PIUS reactor utilizes simplified, inherent, passive, or other innovative means to accomplish safety functions. Accordingly, the PIUS reactor is subject to the requirements of 10CFR52.47(b)(2)(i)(A). This regulation requires that the applicant adequately demonstrate the performance of each safety feature, interdependent effects among the safety features, and a sufficient data base on the safety features of the design to assess the analytical tools used for safety analysis. Los Alamos has assessed the quality and completeness of the existing and planned data bases used by Asea Brown Boveri to validate its safety analysis codes and other relevant data bases. Only amore » limited data base of separate effect and integral tests exist at present. This data base is not adequate to fulfill the requirements of 10CFR52.47(b)(2)(i)(A). Asea Brown Boveri has stated that it plans to conduct more separate effect and integral test programs. If appropriately designed and conducted, these test programs have the potential to satisfy most of the data base requirements of 10CFR52.47(b)(2)(i)(A) and remedy most of the deficiencies of the currently existing combined data base. However, the most important physical processes in PIUS are related to reactor shutdown because the PIUS reactor does not contain rodded shutdown and control systems. For safety-related reactor shutdown, PIUS relies on negative reactivity insertions from the moderator temperature coefficient and from boron entering the core from the reactor pool. Asea Brown Boveri has neither developed a direct experimental data base for these important processes nor provided a rationale for indirect testing of these key PIUS processes. This is assessed as a significant shortcoming. In preparing the conclusions of this report, test documentation and results have been reviewed for only one integral test program, the small-scale integral tests conducted in the ATLE facility.« less

The benefits of an advanced fast reactor fuel cycle for plutonium management

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

Hannum, W.H.; McFarlane, H.F.; Wade, D.C.

1996-12-31

The United States has no program to investigate advanced nuclear fuel cycles for the large-scale consumption of plutonium from military and civilian sources. The official U.S. position has been to focus on means to bury spent nuclear fuel from civilian reactors and to achieve the spent fuel standard for excess separated plutonium, which is considered by policy makers to be an urgent international priority. Recently, the National Research Council published a long awaited report on its study of potential separation and transmutation technologies (STATS), which concluded that in the nuclear energy phase-out scenario that they evaluated, transmutation of plutonium andmore » long-lived radioisotopes would not be worth the cost. However, at the American Nuclear Society Annual Meeting in June, 1996, the STATS panelists endorsed further study of partitioning to achieve superior waste forms for burial, and suggested that any further consideration of transmutation should be in the context of energy production, not of waste management. 2048 The U.S. Department of Energy (DOE) has an active program for the short-term disposition of excess fissile material and a `focus area` for safe, secure stabilization, storage and disposition of plutonium, but has no current programs for fast reactor development. Nevertheless, sufficient data exist to identify the potential advantages of an advanced fast reactor metallic fuel cycle for the long-term management of plutonium. Advantages are discussed.« less

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

Stillman, J. A.; Feldman, E. E.; Jaluvka, D.

This report contains the results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members in the Research and Test Reactor Department at the Argonne National Laboratory (ANL) and the MURR Facility. MURR LEU conversion is part of an overall effort to develop and qualify high-density fuel within the U.S. High Performance Research Reactor Conversion (USHPRR) program conducted by the U.S. Department of Energy National Nuclearmore » Security Administration’s Office of Material Management and Minimization (M 3).« less

Advanced reactors and associated fuel cycle facilities: safety and environmental impacts.

PubMed

Hill, R N; Nutt, W M; Laidler, J J

2011-01-01

The safety and environmental impacts of new technology and fuel cycle approaches being considered in current U.S. nuclear research programs are contrasted to conventional technology options in this paper. Two advanced reactor technologies, the sodium-cooled fast reactor (SFR) and the very high temperature gas-cooled reactor (VHTR), are being developed. In general, the new reactor technologies exploit inherent features for enhanced safety performance. A key distinction of advanced fuel cycles is spent fuel recycle facilities and new waste forms. In this paper, the performance of existing fuel cycle facilities and applicable regulatory limits are reviewed. Technology options to improve recycle efficiency, restrict emissions, and/or improve safety are identified. For a closed fuel cycle, potential benefits in waste management are significant, and key waste form technology alternatives are described. Copyright © 2010 Health Physics Society

Space power reactor in-core thermionic multicell evolutionary (S-prime) design

NASA Astrophysics Data System (ADS)

Determan, William R.; Van Hagan, Tom H.

1993-01-01

A 5- to 40-kWe moderated in-core thermionic space nuclear power system (TI-SNPS) concept was developed to address the TI-SNPS program requirements. The 40-kWe baseline design uses multicell Thermionic Fuel Elements (TFEs) in a zirconium hydride moderated reactor to achieve a specific mass of 18.2 We/kg and a net end-of-mission (EOM) efficiency of 8.2%. The reactor is cooled with a single NaK-78 pumped loop, which rejects the heat through a 24 m2 heat pipe space radiator.

BOILING WATER REACTOR TECHNOLOGY STATUS OF THE ART REPORT. VOLUME II. WATER CHEMISTRY AND CORROSION

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

Breden, C.R.

1963-02-01

Information concerning the corrosive effects of water in power reactor moderator-coolant systems is presented. The information is based on investigations reported in the unclassified literature believed to be fairly complete to 1959, but less complete since then. The material is presented in sections on water decomposition, water chemistry, materials corrosion, corrosion product deposits, and radioactivity. It is noted that the report is presented as a part of a continuing program in development of less expensive materials for use in reactors. (J.R.D.)

Validation of the WIMSD4M cross-section generation code with benchmark results

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

Deen, J.R.; Woodruff, W.L.; Leal, L.E.

1995-01-01

The WIMSD4 code has been adopted for cross-section generation in support of the Reduced Enrichment Research and Test Reactor (RERTR) program at Argonne National Laboratory (ANL). Subsequently, the code has undergone several updates, and significant improvements have been achieved. The capability of generating group-collapsed micro- or macroscopic cross sections from the ENDF/B-V library and the more recent evaluation, ENDF/B-VI, in the ISOTXS format makes the modified version of the WIMSD4 code, WIMSD4M, very attractive, not only for the RERTR program, but also for the reactor physics community. The intent of the present paper is to validate the WIMSD4M cross-section librariesmore » for reactor modeling of fresh water moderated cores. The results of calculations performed with multigroup cross-section data generated with the WIMSD4M code will be compared against experimental results. These results correspond to calculations carried out with thermal reactor benchmarks of the Oak Ridge National Laboratory (ORNL) unreflected HEU critical spheres, the TRX LEU critical experiments, and calculations of a modified Los Alamos HEU D{sub 2}O moderated benchmark critical system. The benchmark calculations were performed with the discrete-ordinates transport code, TWODANT, using WIMSD4M cross-section data. Transport calculations using the XSDRNPM module of the SCALE code system are also included. In addition to transport calculations, diffusion calculations with the DIF3D code were also carried out, since the DIF3D code is used in the RERTR program for reactor analysis and design. For completeness, Monte Carlo results of calculations performed with the VIM and MCNP codes are also presented.« less

Fusion breeder

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

Moir, R.W.

1982-02-22

The fusion breeder is a fusion reactor designed with special blankets to maximize the transmutation by 14 MeV neutrons of uranium-238 to plutonium or thorium to uranium-233 for use as a fuel for fission reactors. Breeding fissile fuels has not been a goal of the US fusion energy program. This paper suggests it is time for a policy change to make the fusion breeder a goal of the US fusion program and the US nuclear energy program. The purpose of this paper is to suggest this policy change be made and tell why it should be made, and to outlinemore » specific research and development goals so that the fusion breeder will be developed in time to meet fissile fuel needs.« less

Fusion breeder

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

Moir, R.W.

1982-04-20

The fusion breeder is a fusion reactor designed with special blankets to maximize the transmutation by 14 MeV neutrons of uranium-238 to plutonium or thorium to uranium-233 for use as a fuel for fission reactors. Breeding fissile fuels has not been a goal of the US fusion energy program. This paper suggests it is time for a policy change to make the fusion breeder a goal of the US fusion program and the US nuclear energy program. The purpose of this paper is to suggest this policy change be made and tell why it should be made, and to outlinemore » specific research and development goals so that the fusion breeder will be developed in time to meet fissile fuel needs.« less

Standard interface files and procedures for reactor physics codes, version III

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

Carmichael, B.M.

Standards and procedures for promoting the exchange of reactor physics codes are updated to Version-III status. Standards covering program structure, interface files, file handling subroutines, and card input format are included. The implementation status of the standards in codes and the extension of the standards to new code areas are summarized. (15 references) (auth)

75 FR 8411 - Office of New Reactors: Interim Staff Guidance on Assessing the Consequences of an Accidental...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-02-24

... the public will be better served by being able to review and comment on both documents at this time... Construction Inspection and Operational Programs, Office of New Reactors, U.S. Nuclear Regulatory Commission... conditions for such releases and define acceptable assumptions to describe exposure scenarios and pathways to...

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

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-03

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

The United Arab Emirates Nuclear Program and Proposed U.S. Nuclear Cooperation

DTIC Science & Technology

2009-05-14

fuel for future civilian light water reactors deployed” in the UAE. The agreement also states that future cooperation may encompass training...planned nuclear reactor . (...continued) May 4, 2008; and, Chris Stanton and Ivan...already taken place. In August 2008, Virginia’s Thorium Power Ltd. signed two consulting and advisory services contracts related to the establishment

Fabrication Technological Development of the Oxide Dispersion Strengthened Alloy MA957 for Fast Reactor Applications

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

Hamilton, Margaret L.; Gelles, David S.; Lobsinger, Ralph J.

A significant amount of effort has been devoted to determining the properties and understanding the behavior of the alloy MA957 to define its potential usefulness as a cladding material in the fast breeder reactor program. The numerous characterization and fabrication studies that were conducted are documented in this report.

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

Monteleone, S.

This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25--27, 1993. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updatedmore » and may differ from those that appeared in the final program of the meeting. Individual papers have been cataloged separately. This document, Volume 1 covers the following topics: Advanced Reactor Research; Advanced Instrumentation and Control Hardware; Advanced Control System Technology; Human Factors Research; Probabilistic Risk Assessment Topics; Thermal Hydraulics; and Thermal Hydraulic Research for Advanced Passive Light Water Reactors.« less

Successful Completion of the Largest Shipment of Russian Research Reactor High-Enriched Uranium Spent Nuclear Fuel from Czech Republic to Russian Federation

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

Michael Tyacke; Dr. Igor Bolshinsky; Jeff Chamberlin

On December 8, 2007, the largest shipment of high-enriched uranium spent nuclear fuel was successfully made from a Russian-designed nuclear research reactor in the Czech Republic to the Russian Federation. This accomplishment is the culmination of years of planning, negotiations, and hard work. The United States, Russian Federation, and the International Atomic Energy Agency have been working together on the Russian Research Reactor Fuel Return (RRRFR) Program in support of the Global Threat Reduction Initiative. In February 2003, RRRFR Program representatives met with the Nuclear Research Institute in Rež, Czech Republic, and discussed the return of their high-enriched uranium spentmore » nuclear fuel to the Russian Federation for reprocessing. Nearly 5 years later, the shipment was made. This paper discusses the planning, preparations, coordination, and cooperation required to make this important international shipment.« less

Analysis of irradiated U-7wt%Mo dispersion fuel microstructures using automated image processing

DOE PAGES

Collette, R.; King, J.; Buesch, C.; ...

2016-04-01

The High Performance Research Reactor Fuel Development (HPPRFD) program is responsible for developing low enriched uranium (LEU) fuel substitutes for high performance reactors fueled with highly enriched uranium (HEU) that have not yet been converted to LEU. The uranium-molybdenum (U-Mo) fuel system was selected for this effort. In this study, fission gas pore segmentation was performed on U-7wt%Mo dispersion fuel samples at three separate fission densities using an automated image processing interface developed in MATLAB. Pore size distributions were attained that showed both expected and unexpected fission gas behavior. In general, it proved challenging to identify any dominant trends whenmore » comparing fission bubble data across samples from different fuel plates due to varying compositions and fabrication techniques. Here, the results exhibited fair agreement with the fission density vs. porosity correlation developed by the Russian reactor conversion program.« less

Analysis of irradiated U-7wt%Mo dispersion fuel microstructures using automated image processing

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

Collette, R.; King, J.; Buesch, C.

The High Performance Research Reactor Fuel Development (HPPRFD) program is responsible for developing low enriched uranium (LEU) fuel substitutes for high performance reactors fueled with highly enriched uranium (HEU) that have not yet been converted to LEU. The uranium-molybdenum (U-Mo) fuel system was selected for this effort. In this study, fission gas pore segmentation was performed on U-7wt%Mo dispersion fuel samples at three separate fission densities using an automated image processing interface developed in MATLAB. Pore size distributions were attained that showed both expected and unexpected fission gas behavior. In general, it proved challenging to identify any dominant trends whenmore » comparing fission bubble data across samples from different fuel plates due to varying compositions and fabrication techniques. Here, the results exhibited fair agreement with the fission density vs. porosity correlation developed by the Russian reactor conversion program.« less

Design of a Low Power, Fast-Spectrum, Liquid-Metal Cooled Surface Reactor System

NASA Astrophysics Data System (ADS)

Marcille, T. F.; Dixon, D. D.; Fischer, G. A.; Doherty, S. P.; Poston, D. I.; Kapernick, R. J.

2006-01-01

In the current 2005 US budget environment, competition for fiscal resources make funding for comprehensive space reactor development programs difficult to justify and accommodate. Simultaneously, the need to develop these systems to provide planetary and deep space-enabling power systems is increasing. Given that environment, designs intended to satisfy reasonable near-term surface missions, using affordable technology-ready materials and processes warrant serious consideration. An initial lunar application design incorporating a stainless structure, 880 K pumped NaK coolant system and a stainless/UO2 fuel system can be designed, fabricated and tested for a fraction of the cost of recent high-profile reactor programs (JIMO, SP-100). Along with the cost reductions associated with the use of qualified materials and processes, this design offers a low-risk, high-reliability implementation associated with mission specific low temperature, low burnup, five year operating lifetime requirements.

The past as prologue - A look at historical flight qualifications for space nuclear systems

NASA Technical Reports Server (NTRS)

Bennett, Gary L.

1992-01-01

Currently the U.S. is sponsoring production of radioisotope thermoelectric generators (RTGs) for the Cassini mission to Saturn; the SP-100 space nuclear reactor power system for NASA applications; a thermionic space reactor program for DoD applications as well as early work on nuclear propulsion. In an era of heightened public concern about having successful space ventures it is important that a full understanding be developed of what it means to 'flight qualify' a space nuclear system. As a contribution to the ongoing work this paper reviews several qualification programs, including the general-purpose heat source radioisotope thermoelectric generators (GPHS-RTGs) as developed for the Galileo and Ulysses missions, the SNAP-10A space reactor, the Nuclear Engine for Rocket Vehicle Applications (NERVA), the F-1 chemical engine used on the Saturn-V, and the Space Shuttle Main Engines (SSMEs). Similarities and contrasts are noted.

The past as prologue - A look at historical flight qualifications for space nuclear systems

NASA Astrophysics Data System (ADS)

Bennett, Gary L.

Currently the U.S. is sponsoring production of radioisotope thermoelectric generators (RTGs) for the Cassini mission to Saturn; the SP-100 space nuclear reactor power system for NASA applications; a thermionic space reactor program for DoD applications as well as early work on nuclear propulsion. In an era of heightened public concern about having successful space ventures it is important that a full understanding be developed of what it means to 'flight qualify' a space nuclear system. As a contribution to the ongoing work this paper reviews several qualification programs, including the general-purpose heat source radioisotope thermoelectric generators (GPHS-RTGs) as developed for the Galileo and Ulysses missions, the SNAP-10A space reactor, the Nuclear Engine for Rocket Vehicle Applications (NERVA), the F-1 chemical engine used on the Saturn-V, and the Space Shuttle Main Engines (SSMEs). Similarities and contrasts are noted.

ReactorHealth Physics operations at the NIST center for neutron research.

PubMed

Johnston, Thomas P

2015-02-01

Performing health physics and radiation safety functions under a special nuclear material license and a research and test reactor license at a major government research and development laboratory encompasses many elements not encountered by industrial, general, or broad scope licenses. This article reviews elements of the health physics and radiation safety program at the NIST Center for Neutron Research, including the early history and discovery of the neutron, applications of neutron research, reactor overview, safety and security of radiation sources and radioactive material, and general health physics procedures. These comprise precautions and control of tritium, training program, neutron beam sample processing, laboratory audits, inventory and leak tests, meter calibration, repair and evaluation, radioactive waste management, and emergency response. In addition, the radiation monitoring systems will be reviewed including confinement building monitoring, ventilation filter radiation monitors, secondary coolant monitors, gaseous fission product monitors, gas monitors, ventilation tritium monitor, and the plant effluent monitor systems.

Development costs for a nuclear electric propulsion stage.

NASA Technical Reports Server (NTRS)

Mondt, J. F.; Prickett, W. Z.

1973-01-01

Development costs are presented for an unmanned nuclear electric propulsion (NEP) stage based upon a liquid metal cooled, in-core thermionic reactor. A total of 120 kWe are delivered to the thrust subsystem which employs mercury ion engines for electric propulsion. This study represents the most recent cost evaluation of the development of a reactor power system for a wide range of nuclear space power applications. These include geocentric, and outer planet and other deep space missions. The development program is described for the total NEP stage, based upon specific development programs for key NEP stage components and subsystems.

Rover nuclear rocket engine program: Overview of rover engine tests

NASA Technical Reports Server (NTRS)

Finseth, J. L.

1991-01-01

The results of nuclear rocket development activities from the inception of the ROVER program in 1955 through the termination of activities on January 5, 1973 are summarized. This report discusses the nuclear reactor test configurations (non cold flow) along with the nuclear furnace demonstrated during this time frame. Included in the report are brief descriptions of the propulsion systems, test objectives, accomplishments, technical issues, and relevant test results for the various reactor tests. Additionally, this document is specifically aimed at reporting performance data and their relationship to fuel element development with little or no emphasis on other (important) items.

Demonstration of catalytic combustion with residual fuel

NASA Technical Reports Server (NTRS)

Dodds, W. J.; Ekstedt, E. E.

1981-01-01

An experimental program was conducted to demonstrate catalytic combustion of a residual fuel oil. Three catalytic reactors, including a baseline configuration and two backup configurations based on baseline test results, were operated on No. 6 fuel oil. All reactors were multielement configurations consisting of ceramic honeycomb catalyzed with palladium on stabilized alumina. Stable operation on residual oil was demonstrated with the baseline configuration at a reactor inlet temperature of about 825 K (1025 F). At low inlet temperature, operation was precluded by apparent plugging of the catalytic reactor with residual oil. Reduced plugging tendency was demonstrated in the backup reactors by increasing the size of the catalyst channels at the reactor inlet, but plugging still occurred at inlet temperature below 725 K (845 F). Operation at the original design inlet temperature of 589 K (600 F) could not be demonstrated. Combustion efficiency above 99.5% was obtained with less than 5% reactor pressure drop. Thermally formed NO sub x levels were very low (less than 0.5 g NO2/kg fuel) but nearly 100% conversion of fuel-bound nitrogen to NO sub x was observed.

Reactor technology assessment and selection utilizing systems engineering approach

NASA Astrophysics Data System (ADS)

Zolkaffly, Muhammed Zulfakar; Han, Ki-In

2014-02-01

The first Nuclear power plant (NPP) deployment in a country is a complex process that needs to consider technical, economic and financial aspects along with other aspects like public acceptance. Increased interest in the deployment of new NPPs, both among newcomer countries and those with expanding programs, necessitates the selection of reactor technology among commercially available technologies. This paper reviews the Systems Decision Process (SDP) of Systems Engineering and applies it in selecting the most appropriate reactor technology for the deployment in Malaysia. The integrated qualitative and quantitative analyses employed in the SDP are explored to perform reactor technology assessment and to select the most feasible technology whose design has also to comply with the IAEA standard requirements and other relevant requirements that have been established in this study. A quick Malaysian case study result suggests that the country reside with PWR (pressurized water reactor) technologies with more detailed study to be performed in the future for the selection of the most appropriate reactor technology for Malaysia. The demonstrated technology assessment also proposes an alternative method to systematically and quantitatively select the most appropriate reactor technology.

Design options for a bunsen reactor.

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

Moore, Robert Charles

2013-10-01

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

Top shield temperatures, C and K Reactors

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

Agar, J.D.

1964-12-28

A modification program is now in progress at the C and K Reactors consisting of an extensive renovation of the graphite channels in the vertical safety rod ststems. The present VSR channels are being enlarged by a graphite coring operation and channel sleeves will be installed in the larger channels. One problem associated with the coring operation is the danger of damaging top thermal shield cooling tubes located close to the VSR channels to such an extent that these tubes will have to be removed from service. If such a condition should exist at one or a number of locationsmore » in the top shield of the reactors after reactor startup, the question remains -- what would the resulting temperatures be of the various components of the top shields? This study was initiated to determine temperature distributions in the top shield complex at the C and K Reactors for various top thermal shield coolant system conditions. Since the top thermal shield cooling system at C Reactor is different than those at the K Reactors, the study was conducted separately for the two different systems.« less

The U.S. Geological Survey's TRIGA® reactor

USGS Publications Warehouse

DeBey, Timothy M.; Roy, Brycen R.; Brady, Sally R.

2012-01-01

The U.S. Geological Survey (USGS) operates a low-enriched uranium-fueled, pool-type reactor located at the Federal Center in Denver, Colorado. The mission of the Geological Survey TRIGA® Reactor (GSTR) is to support USGS science by providing information on geologic, plant, and animal specimens to advance methods and techniques unique to nuclear reactors. The reactor facility is supported by programs across the USGS and is organizationally under the Associate Director for Energy and Minerals, and Environmental Health. The GSTR is the only facility in the United States capable of performing automated delayed neutron analyses for detecting fissile and fissionable isotopes. Samples from around the world are submitted to the USGS for analysis using the reactor facility. Qualitative and quantitative elemental analyses, spatial elemental analyses, and geochronology are performed. Few research reactor facilities in the United States are equipped to handle the large number of samples processed at the GSTR. Historically, more than 450,000 sample irradiations have been performed at the USGS facility. Providing impartial scientific information to resource managers, planners, and other interested parties throughout the world is an integral part of the research effort of the USGS.

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

NASA Technical Reports Server (NTRS)

Wetch, J. R.

1988-01-01

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

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

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

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

2012-11-01

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

Evolution of systems concepts for a 100 kWe class Space Nuclear Power System

NASA Technical Reports Server (NTRS)

Katucki, R.; Josloff, A.; Kirpich, A.; Florio, F.

1985-01-01

Conceptual designs for the SP-100 Space Nuclear Power System have been prepared that meet baseline, backup and growth program scenarios. Near-term advancement in technology was considered in the design of the Baseline Concept. An improved silicon-germanium thermoelectric technique is used to convert the heat from a fast-spectrum, liquid lithium cooled reactor. This system produces a net power of 100 kWe with a 10-year end of life, under the specific constraints of area and volume. Output of the Backup Concept is estimated to be 60 kWe for a 10-year end of life. This system differs from the Baseline Concept because currently available thermoelectric conversion is used from energy supplied by a liquid sodium cooled reactor. The Growth Concept uses Stirling engine conversion to produce 100 kWe within the constraints of mass and volume. The Growth Concept can be scaled up to produce a 1 MWe output that uses the same type reactor developed for the Baseline Concept. Assessments made for each of the program scenarios indicate the key development efforts needed to initiate detailed design and hardware program phases. Development plans were prepared for each scenario that detail the work elements and show the program activities leading to a state of flight readiness.

Baseline Fracture Toughness and CGR testing of alloys X-750 and XM-19 (EPRI Phase I)

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

J. H. Jackson; S. P. Teysseyre

2012-10-01

The Advanced Test Reactor National Scientific User Facility (ATR NSUF) and Electric Power Research Institute (EPRI) formed an agreement to test representative alloys used as reactor structural materials as a pilot program toward establishing guidelines for future ATR NSUF research programs. This report contains results from the portion of this program established as Phase I (of three phases) that entails baseline fracture toughness, stress corrosion cracking (SCC), and tensile testing of selected materials for comparison to similar tests conducted at GE Global Research. The intent of this Phase I research program is to determine baseline properties for the materials ofmore » interest prior to irradiation, and to ensure comparability between laboratories using similar testing techniques, prior to applying these techniques to the same materials after having been irradiated at the Advanced Test Reactor (ATR). The materials chosen for this research are the nickel based super alloy X-750, and nitrogen strengthened austenitic stainless steel XM-19. A spare core shroud upper support bracket of alloy X-750 was purchased by EPRI from Southern Co. and a section of XM-19 plate was purchased by EPRI from GE-Hitachi. These materials were sectioned at GE Global Research and provided to INL.« less

Baseline Fracture Toughness and CGR testing of alloys X-750 and XM-19 (EPRI Phase I)

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

J. H. Jackson; S. P. Teysseyre

2012-02-01

The Advanced Test Reactor National Scientific User Facility (ATR NSUF) and Electric Power Research Institute (EPRI) formed an agreement to test representative alloys used as reactor structural materials as a pilot program toward establishing guidelines for future ATR NSUF research programs. This report contains results from the portion of this program established as Phase I (of three phases) that entails baseline fracture toughness, stress corrosion cracking (SCC), and tensile testing of selected materials for comparison to similar tests conducted at GE Global Research. The intent of this Phase I research program is to determine baseline properties for the materials ofmore » interest prior to irradiation, and to ensure comparability between laboratories using similar testing techniques, prior to applying these techniques to the same materials after having been irradiated at the Advanced Test Reactor (ATR). The materials chosen for this research are the nickel based super alloy X-750, and nitrogen strengthened austenitic stainless steel XM-19. A spare core shroud upper support bracket of alloy X-750 was purchased by EPRI from Southern Co. and a section of XM-19 plate was purchased by EPRI from GE-Hitachi. These materials were sectioned at GE Global Research and provided to INL.« less

Neutron Fluence And DPA Rate Analysis In Pebble-Bed HTR Reactor Vessel Using MCNP

NASA Astrophysics Data System (ADS)

Hamzah, Amir; Suwoto; Rohanda, Anis; Adrial, Hery; Bakhri, Syaiful; Sunaryo, Geni Rina

2018-02-01

In the Pebble-bed HTR reactor, the distance between the core and the reactor vessel is very close and the media inside are carbon and He gas. Neutron moderation capability of graphite material is theoretically lower than that of water-moderated reactors. Thus, it is estimated much more the fast neutrons will reach the reactor vessel. The fast neutron collisions with the atoms in the reactor vessel will result in radiation damage and could be reducing the vessel life. The purpose of this study was to obtain the magnitude of neutron fluence in the Pebble-bed HTR reactor vessel. Neutron fluence calculations in the pebble-bed HTR reactor vessel were performed using the MCNP computer program. By determining the tally position, it can be calculated flux, spectrum and neutron fluence in the position of Pebble-bed HTR reactor vessel. The calculations results of total neutron flux and fast neutron flux in the reactor vessel of 1.82x108 n/cm2/s and 1.79x108 n/cm2/s respectively. The fast neutron fluence in the reactor vessel is 3.4x1017 n/cm2 for 60 years reactor operation. Radiation damage in stainless steel material caused by high-energy neutrons (> 1.0 MeV) will occur when it has reached the neutron flux level of 1.0x1024 n/cm2. The neutron fluence results show that there is no radiation damage in the Pebble-bed HTR reactor vessel, so it is predicted that it will be safe to operate at least for 60 years.

Trench fast reactor design using the microcomputer

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

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

1987-01-01

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

Analysis of fluid fuel flow to the neutron kinetics on molten salt reactor FUJI-12

NASA Astrophysics Data System (ADS)

Aji, Indarta Kuncoro; Waris, Abdul; Permana, Sidik

2015-09-01

Molten Salt Reactor is a reactor are operating with molten salt fuel flowing. This condition interpret that the neutron kinetics of this reactor is affected by the flow rate of the fuel. This research analyze effect by the alteration velocity of the fuel by MSR type Fuji-12, with fuel composition LiF-BeF2-ThF4-233UF4 respectively 71.78%-16%-11.86%-0.36%. Calculation process in this study is performed numerically by SOR and finite difference method use C programming language. Data of reactivity, neutron flux, and the macroscopic fission cross section for calculation process obtain from SRAC-CITATION (Standard thermal Reactor Analysis Code) and JENDL-4.0 data library. SRAC system designed and developed by JAEA (Japan Atomic Energy Agency). This study aims to observe the effect of the velocity of fuel salt to the power generated from neutron precursors at fourth year of reactor operate (last critical condition) with number of multiplication effective; 1.0155.

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

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

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

1994-07-01

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

Development of a Polysilicon Process Based on Chemical Vapor Deposition of Dichlorosilane in an Advanced Siemen's Reactor

NASA Technical Reports Server (NTRS)

Arevidson, A. N.; Sawyer, D. H.; Muller, D. M.

1983-01-01

Dichlorosilane (DCS) was used as the feedstock for an advanced decomposition reactor for silicon production. The advanced reactor had a cool bell jar wall temperature, 300 C, when compared to Siemen's reactors previously used for DCS decomposition. Previous reactors had bell jar wall temperatures of approximately 750 C. The cooler wall temperature allows higher DCS flow rates and concentrations. A silicon deposition rate of 2.28 gm/hr-cm was achieved with power consumption of 59 kWh/kg. Interpretation of data suggests that a 2.8 gm/hr-cm deposition rate is possible. Screening of lower cost materials of construction was done as a separate program segment. Stainless Steel (304 and 316), Hastalloy B, Monel 400 and 1010-Carbon Steel were placed individually in an experimental scale reactor. Silicon was deposited from trichlorosilane feedstock. The resultant silicon was analyzed for electrically active and metallic impurities as well as carbon. No material contributed significant amounts of electrically active or metallic impurities, but all contributed carbon.

Investigation of materials for fusion power reactors

NASA Astrophysics Data System (ADS)

Bouhaddane, A.; Slugeň, V.; Sojak, S.; Veterníková, J.; Petriska, M.; Bartošová, I.

2014-06-01

The possibility of application of nuclear-physical methods to observe radiation damage to structural materials of nuclear facilities is nowadays a very actual topic. The radiation damage to materials of advanced nuclear facilities, caused by extreme radiation stress, is a process, which significantly limits their operational life as well as their safety. In the centre of our interest is the study of the radiation degradation and activation of the metals and alloys for the new nuclear facilities (Generation IV fission reactors, fusion reactors ITER and DEMO). The observation of the microstructure changes in the reactor steels is based on experimental investigation using the method of positron annihilation spectroscopy (PAS). The experimental part of the work contains measurements focused on model reactor alloys and ODS steels. There were 12 model reactor steels and 3 ODS steels. We were investigating the influence of chemical composition on the production of defects in crystal lattice. With application of the LT 9 program, the spectra of specimen have been evaluated and the most convenient samples have been determined.

PARTIAL ECONOMIC STUDY OF STEAM COOLED HEAVY WATER MODERATED REACTORS

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

None

1960-04-01

Steam-cooled reactors are compared with CAHDU for costs of Calandria tubes, pressure tubes. heavy water moderator, heavy water reflector, fuel supply, heat exchanger, and turbine generator. A direct-cycle lightsteam-cooled heavy- water-moderated pressure-tube reactor formed the basic reactor design for the study. Two methods of steam circulation through the reactor were examined. In both cases the steam was generated outside the reactor and superheated in the reactor core. One method consisted of a series of reactor and steam generator passes. The second method consisted of the Loeffler cycle and its modifications. The fuel was assumed to be natural cylindrical UO/sub 2/more » pellets sheathed in a hypothetical material with the nuclear properties of Zircaloy, but able to function at temperatures to 900 deg F. For the conditions assumed, the longer the rod, the higher the outlet temperature and therefore the higher the efficiency. The turbine cycle efficiency was calculated on the assumption that suitable steam generators are available. As the neutron losses to the pressure tubes were significant, an economic analysis of insulated pressure tubes is included. A description of the physics program for steam-cooled reactors is included. Results indicated that power from the steam-cooled reactor would cost 1.4 mills/ kwh compared with 1.25 mills/kwh for CANDU. (M.C.G.)« less

Review of Nuclear Thermal Propulsion Ground Test Options

NASA Technical Reports Server (NTRS)

Coote, David J.; Power, Kevin P.; Gerrish, Harold P.; Doughty, Glen

2015-01-01

High efficiency rocket propulsion systems are essential for humanity to venture beyond the moon. Nuclear Thermal Propulsion (NTP) is a promising alternative to conventional chemical rockets with relatively high thrust and twice the efficiency of highest performing chemical propellant engines. NTP utilizes the coolant of a nuclear reactor to produce propulsive thrust. An NTP engine produces thrust by flowing hydrogen through a nuclear reactor to cool the reactor, heating the hydrogen and expelling it through a rocket nozzle. The hot gaseous hydrogen is nominally expected to be free of radioactive byproducts from the nuclear reactor; however, it has the potential to be contaminated due to off-nominal engine reactor performance. NTP ground testing is more difficult than chemical engine testing since current environmental regulations do not allow/permit open air testing of NTP as was done in the 1960's and 1970's for the Rover/NERVA program. A new and innovative approach to rocket engine ground test is required to mitigate the unique health and safety risks associated with the potential entrainment of radioactive waste from the NTP engine reactor core into the engine exhaust. Several studies have been conducted since the ROVER/NERVA program in the 1970's investigating NTP engine ground test options to understand the technical feasibility, identify technical challenges and associated risks and provide rough order of magnitude cost estimates for facility development and test operations. The options can be divided into two distinct schemes; (1) real-time filtering of the engine exhaust and its release to the environment or (2) capture and storage of engine exhaust for subsequent processing.

Secure Retrieval of FFTF Testing, Design, and Operating Information

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

Butner, R. Scott; Wootan, David W.; Omberg, Ronald P.

One of the goals of the Advanced Fuel Cycle Initiative (AFCI) is to preserve the knowledge that has been gained in the United States on Liquid Metal Reactors (LMR). In addition, preserving LMR information and knowledge is part of a larger international collaborative activity conducted under the auspices of the International Atomic Energy Agency (IAEA). A similar program is being conducted for EBR-II at the Idaho Nuclear Laboratory (INL) and international programs are also in progress. Knowledge preservation at the FFTF is focused on the areas of design, construction, startup, and operation of the reactor. As the primary function ofmore » the FFTF was testing, the focus is also on preserving information obtained from irradiation testing of fuels and materials. This information will be invaluable when, at a later date, international decisions are made to pursue new LMRs. In the interim, this information may be of potential use for international exchanges with other LMR programs around the world. At least as important in the United States, which is emphasizing large-scale computer simulation and modeling, this information provides the basis for creating benchmarks for validating and testing these large scale computer programs. Although the preservation activity with respect to FFTF information as discussed below is still underway, the team of authors above is currently retrieving and providing experimental and design information to the LMR modeling and simulation efforts for use in validating their computer models. On the Hanford Site, the FFTF reactor plant is one of the facilities intended for decontamination and decommissioning consistent with the cleanup mission on this site. The reactor facility has been deactivated and is being maintained in a cold and dark minimal surveillance and maintenance mode until final decommissioning is pursued. In order to ensure protection of information at risk, the program to date has focused on sequestering and secure retrieval. Accomplishments include secure retrieval of: more than 400 boxes of FFTF information, several hundred microfilm reels including Clinch River Breeder Reactor (CRBR) information, and 40 boxes of information on the Fuels and Materials Examination Facility (FMEF). All information preserved to date is now being stored and categorized consistent with the IAEA international standardized taxonomy. Earlier information largely related to irradiation testing is likewise being categorized. The fuel test results information exists in several different formats depending upon the final stage of the test evaluation. In some cases there is information from both non-destructive and destructive examination while in other cases only non-destructive results are available. Non-destructive information would include disassembly records, dimensional profilometry, gamma spectrometry, and neutron radiography. Information from destructive examinations would include fission gas analysis, metallography, and photomicrographs. Archiving of FFTF data, including both the reactor plant and the fuel test information, is being performed in coordination with other data archiving efforts underway under the aegis of the AFCI program. In addition to the FFTF efforts, archiving of data from the EBR-II reactor is being carried out by INL. All material at risk associated with FFTF documentation has been secured in a timely manner consistent with the stated plan. This documentation is now being categorized consistent with internationally agreed upon IAEA standards. Documents are being converted to electronic format for transfer to a large searchable electronic database being developed by INL. In addition, selected FFTF information is being used to generate test cases for large-scale simulation modeling efforts and for providing Design Data Need (DDN) packages as requested by the AFCI program.« less

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

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

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

2006-10-01

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

Application of the Enabler to nuclear electric propulsion

NASA Astrophysics Data System (ADS)

Pierce, Bill L.

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

BRENDA: a dynamic simulator for a sodium-cooled fast reactor power plant

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

Hetrick, D.L.; Sowers, G.W.

1978-06-01

This report is a users' manual for one version of BRENDA (Breeder Reactor Nuclear Dynamic Analysis), which is a digital program for simulating the dynamic behavior of a sodium-cooled fast reactor power plant. This version, which contains 57 differential equations, represents a simplified model of the Clinch River Breeder Reactor Project (CRBRP). BRENDA is an input deck for DARE P (Differential Analyzer Replacement, Portable), which is a continuous-system simulation language developed at the University of Arizona. This report contains brief descriptions of DARE P and BRENDA, instructions for using BRENDA in conjunction with DARE P, and some sample output. Amore » list of variable names and a listing for BRENDA are included as appendices.« less

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

Yung, Matthew M.; Stanton, Alexander R.; Iisa, Kristiina

Metal-impregnated (Ni or Ga) ZSM-5 catalysts were studied for biomass pyrolysis vapor upgrading to produce hydrocarbons using three reactors constituting a 100 000x change in the amount of catalyst used in experiments. Catalysts were screened for pyrolysis vapor phase upgrading activity in two small-scale reactors: (i) a Pyroprobe with a 10 mg catalyst in a fixed bed and (ii) a fixed-bed reactor with 500 mg of catalyst. The best performing catalysts were then validated with a larger scale fluidized-bed reactor (using ~1 kg of catalyst) that produced measurable quantities of bio-oil for analysis and evaluation of mass balances. Despite somemore » inherent differences across the reactor systems (such as residence time, reactor type, analytical techniques, mode of catalyst and biomass feed) there was good agreement of reaction results for production of aromatic hydrocarbons, light gases, and coke deposition. Relative to ZSM-5, Ni or Ga addition to ZSM-5 increased production of fully deoxygenated aromatic hydrocarbons and light gases. In the fluidized bed reactor, Ga/ZSM-5 slightly enhanced carbon efficiency to condensed oil, which includes oxygenates in addition to aromatic hydrocarbons, and reduced oil oxygen content compared to ZSM-5. Ni/ZSM-5, while giving the highest yield of fully deoxygenated aromatic hydrocarbons, gave lower overall carbon efficiency to oil but with the lowest oxygen content. Reaction product analysis coupled with fresh and spent catalyst characterization indicated that the improved performance of Ni/ZSM-5 is related to decreasing deactivation by coking, which keeps the active acid sites accessible for the deoxygenation and aromatization reactions that produce fully deoxygenated aromatic hydrocarbons. The addition of Ga enhances the dehydrogenation activity of the catalyst, which leads to enhanced olefin formation and higher fully deoxygenated aromatic hydrocarbon yields compared to unmodified ZSM-5. Catalyst characterization by ammonia temperature programmed desorption, surface area measurements, and postreaction temperature-programmed oxidation (TPO) also showed that the metal-modified zeolites retained a greater percentage of their initial acidity and surface area, which was consistent between the reactor scales. These results demonstrate that the trends observed with smaller (milligram to gram) catalyst reactors are applicable to larger, more industrially relevant (kg) scales to help guide catalyst research toward application.« less

Reactor Operations Monitoring System

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

Hart, M.M.

1989-01-01

The Reactor Operations Monitoring System (ROMS) is a VME based, parallel processor data acquisition and safety action system designed by the Equipment Engineering Section and Reactor Engineering Department of the Savannah River Site. The ROMS will be analyzing over 8 million signal samples per minute. Sixty-eight microprocessors are used in the ROMS in order to achieve a real-time data analysis. The ROMS is composed of multiple computer subsystems. Four redundant computer subsystems monitor 600 temperatures with 2400 thermocouples. Two computer subsystems share the monitoring of 600 reactor coolant flows. Additional computer subsystems are dedicated to monitoring 400 signals from assortedmore » process sensors. Data from these computer subsystems are transferred to two redundant process display computer subsystems which present process information to reactor operators and to reactor control computers. The ROMS is also designed to carry out safety functions based on its analysis of process data. The safety functions include initiating a reactor scram (shutdown), the injection of neutron poison, and the loadshed of selected equipment. A complete development Reactor Operations Monitoring System has been built. It is located in the Program Development Center at the Savannah River Site and is currently being used by the Reactor Engineering Department in software development. The Equipment Engineering Section is designing and fabricating the process interface hardware. Upon proof of hardware and design concept, orders will be placed for the final five systems located in the three reactor areas, the reactor training simulator, and the hardware maintenance center.« less

Evaluation of ilmenite serpentine concrete and ordinary concrete as nuclear reactor shielding

NASA Astrophysics Data System (ADS)

Abulfaraj, Waleed H.; Kamal, Salah M.

1994-07-01

The present study involves adapting a formal decision methodology to the selection of alternative nuclear reactor concretes shielding. Multiattribute utility theory is selected to accommodate decision makers' preferences. Multiattribute utility theory (MAU) is here employed to evaluate two appropriate nuclear reactor shielding concretes in terms of effectiveness to determine the optimal choice in order to meet the radiation protection regulations. These concretes are Ordinary concrete (O.C.) and Ilmenite Serpentile concrete (I.S.C.). These are normal weight concrete and heavy heat resistive concrete, respectively. The effectiveness objective of the nuclear reactor shielding is defined and structured into definite attributes and subattributes to evaluate the best alternative. Factors affecting the decision are dose received by reactor's workers, the material properties as well as cost of concrete shield. A computer program is employed to assist in performing utility analysis. Based upon data, the result shows the superiority of Ordinary concrete over Ilmenite Serpentine concrete.

Generic Stellarator-like Magnetic Fusion Reactor

NASA Astrophysics Data System (ADS)

Sheffield, John; Spong, Donald

2015-11-01

The Generic Magnetic Fusion Reactor paper, published in 1985, has been updated, reflecting the improved science and technology base in the magnetic fusion program. Key changes beyond inflation are driven by important benchmark numbers for technologies and costs from ITER construction, and the use of a more conservative neutron wall flux and fluence in modern fusion reactor designs. In this paper the generic approach is applied to a catalyzed D-D stellarator-like reactor. It is shown that an interesting power plant might be possible if the following parameters could be achieved for a reference reactor: R/ < a > ~ 4 , confinement factor, fren = 0.9-1.15, < β > ~ 8 . 0 -11.5 %, Zeff ~ 1.45 plus a relativistic temperature correction, fraction of fast ions lost ~ 0.07, Bm ~ 14-16 T, and R ~ 18-24 m. J. Sheffield was supported under ORNL subcontract 4000088999 with the University of Tennessee.

Sodium Based Heat Pipe Modules for Space Reactor Concepts: Stainless Steel SAFE-100 Core

NASA Technical Reports Server (NTRS)

Martin, James J.; Reid, Robert S.

2004-01-01

A heat pipe cooled reactor is one of several candidate reactor cores being considered for advanced space power and propulsion systems to support future space exploration applications. Long life heat pipe modules, with designs verified through a combination of theoretical analysis and experimental lifetime evaluations, would be necessary to establish the viability of any of these candidates, including the heat pipe reactor option. A hardware-based program was initiated to establish the infrastructure necessary to build heat pipe modules. This effort, initiated by Los Alamos National Laboratory and referred to as the Safe Affordable Fission Engine (SAFE) project, set out to fabricate and perform non-nuclear testing on a modular heat pipe reactor prototype that can provide 100 kilowatt from the core to an energy conversion system at 700 C. Prototypic heat pipe hardware was designed, fabricated, filled, closed-out and acceptance tested.

Small and medium power reactors 1987

NASA Astrophysics Data System (ADS)

1987-12-01

This TECDOC follows the publication of TECDOC-347: Small and Medium Power Reactors (SMPR) Project Initiation Study, Phase 1, published in 1985 and TECDOC-376: Small and Medium Power Reactors 1985 published in 1986. It is mainly intended for decision makers in Developing Member States interested in embarking on a nuclear power program. It consists of two parts: (1) guidelines for the introduction of small and medium power reactors in developing countries. These Guidelines were established during the Advisory Group Meeting held in Vienna from 11 to 15 May 1987. Their purpose is to review key aspects relating to the introduction of small and medium power reactors in developing countries; (2) up-dated information on SMPR Concepts Contributed by Supplier Industries. According to the recommendations of the Second Technical Committee Meeting on SMPRs held in Vienna in March 1985, this part contains the up-dated information formerly published in Annex 1 of the above mentioned TECDOC-347.

Fast Breeder Reactors in Sweden: Vision and Reality.

PubMed

Fjaestad, Maja

2015-01-01

The fast breeder is a type of nuclear reactor that aroused much attention in the 1950s and '60s. Its ability to produce more nuclear fuel than it consumes offered promises of cheap and reliable energy. Sweden had advanced plans for a nuclear breeder program, but canceled them in the middle of the 1970s with the rise of nuclear skepticism. The article investigates the nuclear breeder as a technological vision. The nuclear breeder reactor is an example of a technological future that did not meet its industrial expectations. But that does not change the fact that the breeder was an influential technology. Decisions about the contemporary reactors were taken with the idea that in a foreseeable future they would be replaced with the efficient breeder. The article argues that general themes in the history of the breeder reactor can deepen our understanding of the mechanisms behind technological change.

75 FR 56597 - University of Wisconsin; University of Wisconsin Nuclear Reactor Environmental Assessment and...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-09-16

...,000 gallons being typical. The licensee maintains a pool leak surveillance program. The pool water leak surveillance program continues to monitor the pool water evaporation rate, the pool water make-up volume, and pool water radioactivity. The pool leak surveillance program indicated that approximately 2...

10 CFR 72.144 - Quality assurance program.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Quality assurance program. 72.144 Section 72.144 Energy... NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Quality Assurance § 72.144 Quality assurance program. (a) The licensee, applicant for a license, certificate holder...

77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2012-06-15

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0134] Initial Test Program of Emergency Core Cooling... for public comment draft regulatory guide (DG), DG-1277, ``Initial Test Program of Emergency Core... acceptable to implement with regard to initial testing features of emergency core cooling systems (ECCSs) for...

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

Bucholz, J.A.

The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory is in the midst of a massive upgrade program to enhance experimental facilities. The reactor presently has four horizontal experimental beam tubes, all of which will be replaced or redesigned. The HB-2 beam tube will be enlarged to support more guide tubes, while the HB-4 beam tube will soon include a cold neutron source.

Reduced enrichment for research and test reactors: Proceedings

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

Not Available

1993-07-01

The 15th annual Reduced Enrichment for Research and Test Reactors (RERTR) international meeting was organized by Ris{o} National Laboratory in cooperation with the International Atomic Energy Agency and Argonne National Laboratory. The topics of the meeting were the following: National Programs, Fuel Fabrication, Licensing Aspects, States of Conversion, Fuel Testing, and Fuel Cycle. Individual papers have been cataloged separately.

Cermet-fueled reactors for advanced space applications

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

Cowan, C.L.; Palmer, R.S.; Taylor, I.N.

Cermet-fueled nuclear reactors are attractive candidates for high-performance advanced space power systems. The cermet consists of a hexagonal matrix of a refractory metal and a ceramic fuel, with multiple tubular flow channels. The high performance characteristics of the fuel matrix come from its high strength at elevated temperatures and its high thermal conductivity. The cermet fuel concept evolved in the 1960s with the objective of developing a reactor design that could be used for a wide range of mobile power generating sytems, including both Brayton and Rankine power conversion cycles. High temperature thermal cycling tests for the cermet fuel weremore » carried out by General Electric as part of the 710 Project (General Electric 1966), and by Argonne National Laboratory in the Direct Nuclear Rocket Program (1965). Development programs for cermet fuel are currently under way at Argonne National Laboratory and Pacific Northwest Laboratory. The high temperature qualification tests from the 1960s have provided a base for the incorporation of cermet fuel in advanced space applications. The status of the cermet fuel development activities and descriptions of the key features of the cermet-fueled reactor design are summarized in this paper.« less

A cost/benefit analysis of commercial fusion-fission hybrid reactor development

NASA Astrophysics Data System (ADS)

Kostoff, Ronald N.

1983-04-01

A simple algorithm was developed that allows rapid computation of the ratio, R, of present worth of benefits to present worth of hybrid R&D program costs as a function of potential hybrid unit electricity cost savings, discount rate, electricity demand growth rate, total hybrid R&D program cost, and time to complete a demonstration reactor. In the sensitivity study, these variables were assigned nominal values (unit electricity cost savings of 4 mills/kW-hr, discount rate of 4%/year, growth rate of 2.25%/year, total R&D program cost of 20 billion, and time to complete a demonstration reactor of 30 years), and the variable of interest was varied about its nominal value. Results show that R increases with decreasing discount rate and increasing unit electricity savings and ranges from 4 to 94 as discount rate ranges from 5 to 3%/year and unit electricity savings range from 2 to 6 mills/kW-hr. R increases with increasing growth rate and ranges from 3 to 187 as growth rate ranges from 1 to 3.5%/year and unit electricity cost savings range from 2 to 6 mills/kW-hr. R attains a maximum value when plotted against time to complete a demonstration reactor. The location of this maximum value occurs at shorter completion times as discount rate increases, and this optimal completion time ranges from 20 years for a discount rate of 4%/year to 45 years for a discount rate of 3%/year.

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

Moren, Richard J.; Grindstaff, Keith D.

Hanford's Long-Term Stewardship (LTS) Program has evolved from a small, informal process, with minimal support, to a robust program that provides comprehensive transitions from cleanup contractors to long-term stewardship for post-cleanup requirements specified in the associated cleanup decision documents. The LTS Program has the responsibility for almost 100,000 acres of land, along with over 200 waste sites and will soon have six cocooned reactors. Close to 2,600 documents have been identified and tagged for storage in the LTS document library. The program has successfully completed six consecutive transitions over the last two years in support of the U.S. DOE Richlandmore » Operations Office's (DOE-RL) near-term cleanup objectives of significantly reducing the footprint of active cleanup operations for the River Corridor. The program has evolved from one that was initially responsible for defining and measuring Institutional Controls for the Hanford Site, to a comprehensive, post remediation surveillance and maintenance program that begins early in the transition process. In 2013, the first reactor area -- the cocooned 105-F Reactor and its surrounding 1,100 acres, called the F Area was transitioned. In another first, the program is expected to transition the five remaining cocooned reactors into the program through using a Transition and Turnover Package (TTP). As Hanford's LTS Program moves into the next few years, it will continue to build on a collaborative approach. The program has built strong relationships between contractors, regulators, tribes and stakeholders and with the U.S. Department of Energy's Office of Legacy Management (LM). The LTS Program has been working with LM since its inception. The transition process utilized LM's Site Transition Framework as one of the initial requirement documents and the Hanford Program continues to collaborate with LM today. One example of this collaboration is the development of the LTS Program's records management system in which, LM has been instrumental. The development of a rigorous data collection and records management systems has been influenced and built off of LMs success, which also ensures compatibility between what Hanford's LTS Program develops and LM. In another example, we are exploring a pilot project to ship records from the Hanford Site directly to LM for long-term storage. This pilot would gain program efficiencies so that records would be handled only once. Rather than storage on-site, then shipment to an interim Federal Records Center in Seattle, records would be shipped directly to LM. The Hanford LTS Program is working to best align programmatic processes, find efficiencies, and to benchmark site transition requirements. Involving the Hanford LTS Program early in the transition process with an integrated contractor and DOE team is helping to ensure that there is time to work through details on the completed remediation of transitioning areas. It also will allow for record documentation and storage for the future, and is an opportunity for the program to mature through the experiences that will be gained by implementing LTS Program activities over time.« less

ATF Neutron Irradiation Program Technical Plan

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

Geringer, J. W.; Katoh, Yutai

The Japan Atomic Energy Agency (JAEA) under the Civil Nuclear Energy Working Group (CNWG) is engaged in a cooperative research effort with the U.S. Department of Energy (DOE) to explore issues related to nuclear energy, including research on accident-tolerant fuels and materials for use in light water reactors. This work develops a draft technical plan for a neutron irradiation program on the candidate accident-tolerant fuel cladding materials and elements using the High Flux Isotope Reactor (HFIR). The research program requires the design of a detailed experiment, development of test vehicles, irradiation of test specimens, possible post-irradiation examination and characterization ofmore » irradiated materials and the shipment of irradiated materials to JAEA in Japan. This report discusses the technical plan of the experimental study.« less

ATF Neutron Irradiation Program Irradiation Vehicle Design Concepts

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

Geringer, J. W.; Katoh, Yutai; Howard, Richard H.

The Japan Atomic Energy Agency (JAEA) under the Civil Nuclear Energy Working Group (CNWG) is engaged in a cooperative research effort with the U.S. Department of Energy (DOE) to explore issues related to nuclear energy, including research on accident-tolerant fuels and materials for use in light water reactors. This work develops a draft technical plan for a neutron irradiation program on the candidate accident-tolerant fuel cladding materials and elements using the High Flux Isotope Reactor (HFIR). The research program requires the design of a detailed experiment, development of test vehicles, irradiation of test specimens, possible post irradiation examination and characterizationmore » of irradiated materials and the shipment of irradiated materials to Japan. This report discusses the conceptual design, the development and irradiation of the test vehicles.« less

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

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

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

2017-06-01

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

Assessment of Sensor Technologies for Advanced Reactors

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

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

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

Aluminum Carbothermic Technology

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

Bruno, Marshall J.

2005-03-31

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

DOE-NE Proliferation and Terrorism Risk Assessment: FY12 Plans Update

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

Sadasivan, Pratap

2012-06-21

This presentation provides background information on FY12 plans for the DOE Office of Nuclear Energy Proliferation and Terrorism Risk Assessment program. Program plans, organization, and individual project elements are described. Research objectives are: (1) Develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of current reactors; (2) Develop improvements in the affordability of new reactors to enable nuclear energy; (3) Develop Sustainable Nuclear Fuel Cycles; and (4) Understand and minimize the risks of nuclear proliferation and terrorism - Goal is to enable the use of risk information to inform NE R&D programmore » planning.« less

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

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

Stoner, K.J.

1999-11-05

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

Solid Polymer Electrolyte Fuel Cell Technology Program

NASA Technical Reports Server (NTRS)

1980-01-01

Work is reported on phase 5 of the Solid Polymer Electrolyte (SPE) Fuel Cell Technology Development program. The SPE fuel cell life and performance was established at temperatures, pressures, and current densities significantly higher than those previously demonstrated in sub-scale hardware. Operation of single-cell Buildup No. 1 to establish life capabilities of the full-scale hardware was continued. A multi-cell full-scale unit (Buildup No. 2) was designed, fabricated, and test evaluated laying the groundwork for the construction of a reactor stack. A reactor stack was then designed, fabricated, and successfully test-evaluated to demonstrate the readiness of SPE fuel cell technology for future space applications.

Application of a Systems Engineering Approach to Support Space Reactor Development

NASA Astrophysics Data System (ADS)

Wold, Scott

2005-02-01

In 1992, approximately 25 Russian and 12 U.S. engineers and technicians were involved in the transport, assembly, inspection, and testing of over 90 tons of Russian equipment associated with the Thermionic System Evaluation Test (TSET) Facility. The entire Russian Baikal Test Stand, consisting of a 5.79 m tall vacuum chamber and related support equipment, was reassembled and tested at the TSET facility in less than four months. In November 1992, the first non-nuclear operational test of a complete thermionic power reactor system in the U.S. was accomplished three months ahead of schedule and under budget. A major factor in this accomplishment was the application of a disciplined top-down systems engineering approach and application of a spiral development model to achieve the desired objectives of the TOPAZ International Program (TIP). Systems Engineering is a structured discipline that helps programs and projects conceive, develop, integrate, test and deliver products and services that meet customer requirements within cost and schedule. This paper discusses the impact of Systems Engineering and a spiral development model on the success of the TOPAZ International Program and how the application of a similar approach could help ensure the success of future space reactor development projects.

A Practical Approach to Starting Fission Surface Power Development

NASA Technical Reports Server (NTRS)

Mason, Lee S.

2006-01-01

The Prometheus Power and Propulsion Program has been reformulated to address NASA needs relative to lunar and Mars exploration. Emphasis has switched from the Jupiter Icy Moons Orbiter (JIMO) flight system development to more generalized technology development addressing Fission Surface Power (FSP) and Nuclear Thermal Propulsion (NTP). Current NASA budget priorities and the deferred mission need date for nuclear systems prohibit a fully funded reactor Flight Development Program. However, a modestly funded Advanced Technology Program can and should be conducted to reduce the risk and cost of future flight systems. A potential roadmap for FSP technology development leading to possible flight applications could include three elements: 1) Conceptual Design Studies, 2) Advanced Component Technology, and 3) Non-Nuclear System Testing. The Conceptual Design Studies would expand on recent NASA and DOE analyses while increasing the depth of study in areas of greatest uncertainty such as reactor integration and human-rated shielding. The Advanced Component Technology element would address the major technology risks through development and testing of reactor fuels, structural materials, primary loop components, shielding, power conversion, heat rejection, and power management and distribution (PMAD). The Non-Nuclear System Testing would provide a modular, technology testbed to investigate and resolve system integration issues.

Nuclear Theft: Real and Imagined Dangers

DTIC Science & Technology

1976-03-01

are utilized in connection with fossil fuel energy research and development programs and related activities conducted by the Bureau of Mines "energy... development associated with the U.S. nuclear weapons program . Addition- ally, ERDA conducts related programs which include power reactor design... development , nuclear propulsion, and other systems associated with space programs . The military and ERDA enjoy a symbiotic relationship in that nuclear

Flow reversal power limit for the HFBR

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

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

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

Study Neutronic of Small Pb-Bi Cooled Non-Refuelling Nuclear Power Plant Reactor (SPINNOR) with Hexagonal Geometry Calculation

NASA Astrophysics Data System (ADS)

Nur Krisna, Dwita; Su'ud, Zaki

2017-01-01

Nuclear reactor technology is growing rapidly, especially in developing Nuclear Power Plant (NPP). The utilization of nuclear energy in power generation systems has been progressing phase of the first generation to the fourth generation. This final project paper discusses the analysis neutronic one-cooled fast reactor type Pb-Bi, which is capable of operating up to 20 years without refueling. This reactor uses Thorium Uranium Nitride as fuel and operating on power range 100-500MWtNPPs. The method of calculation used a computer simulation program utilizing the SRAC. SPINNOR reactor is designed with the geometry of hexagonal shaped terrace that radially divided into three regions, namely the outermost regions with highest percentage of fuel, the middle regions with medium percentage of fuel, and most in the area with the lowest percentage. SPINNOR fast reactor operated for 20 years with variations in the percentage of Uranium-233 by 7%, 7.75%, and 8.5%. The neutronic calculation and analysis show that the design can be optimized in a fast reactor for thermal power output SPINNOR 300MWt with a fuel fraction 60% and variations of Uranium-233 enrichment of 7%-8.5%.

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

Kupca, L.; Beno, P.

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

Assessment of the high temperature fission chamber technology for the French fast reactor program

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

Jammes, C.; Filliatre, P.; Geslot, B.

2011-07-01

High temperature fission chambers are key instruments for the control and protection of the sodium-cooled fast reactor. First, the developments of those neutron detectors, which are carried out either in France or abroad are reviewed. Second, the French realizations are assessed with the use of the technology readiness levels in order to identify tracks of improvement. (authors)

Early Program Development

NASA Image and Video Library

2004-04-15

This artist's concept illustrates the NERVA (Nuclear Engine for Rocket Vehicle Application) engine's hot bleed cycle in which a small amount of hydrogen gas is diverted from the thrust nozzle, thus eliminating the need for a separate system to drive the turbine. The NERVA engine, based on KIWI nuclear reactor technology, would power a RIFT (Reactor-In-Flight-Test) nuclear stage, for which the Marshall Space Flight Center had development responsibility.

76 FR 17159 - Office of New Reactors; Final Interim Staff Guidance on Standard Review Plan, Section 17.4...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-28

... Interim Staff Guidance on Standard Review Plan, Section 17.4, ``Reliability Assurance Program'' AGENCY... design reliability assurance program (RAP). This ISG updates the guidance provided to the staff in Standard Review Plan (SRP), Section 17.4, ``Reliability Assurance Program,'' of NUREG-0800, ``Standard...

Nonlinear Ultrasonic Measurements in Nuclear Reactor Environments

NASA Astrophysics Data System (ADS)

Reinhardt, Brian T.

Several Department of Energy Office of Nuclear Energy (DOE-NE) programs, such as the Fuel Cycle Research and Development (FCRD), Advanced Reactor Concepts (ARC), Light Water Reactor Sustainability, and Next Generation Nuclear Power Plants (NGNP), are investigating new fuels, materials, and inspection paradigms for advanced and existing reactors. A key objective of such programs is to understand the performance of these fuels and materials during irradiation. In DOE-NE's FCRD program, ultrasonic based technology was identified as a key approach that should be pursued to obtain the high-fidelity, high-accuracy data required to characterize the behavior and performance of new candidate fuels and structural materials during irradiation testing. The radiation, high temperatures, and pressure can limit the available tools and characterization methods. In this thesis, two ultrasonic characterization techniques will be explored. The first, finite amplitude wave propagation has been demonstrated to be sensitive to microstructural material property changes. It is a strong candidate to determine fuel evolution; however, it has not been demonstrated for in-situ reactor applications. In this thesis, finite amplitude wave propagation will be used to measure the microstructural evolution in Al-6061. This is the first demonstration of finite amplitude wave propagation at temperatures in excess of 200 °C and during an irradiation test. Second, a method based on contact nonlinear acoustic theory will be developed to identify compressed cracks. Compressed cracks are typically transparent to ultrasonic wave propagation; however, by measuring harmonic content developed during finite amplitude wave propagation, it is shown that even compressed cracks can be characterized. Lastly, piezoelectric transducers capable of making these measurements are developed. Specifically, three piezoelectric sensors (Bismuth Titanate, Aluminum Nitride, and Zinc Oxide) are tested in the Massachusetts Institute of Technology Research reactor to a fast neutron fluence of 8.65x10 20 n/cm2. It is demonstrated that Bismuth Titanate is capable of transduction up to 5 x1020 n/cm2, Zinc Oxide is capable of transduction up to 6.27 x1020 n/cm 2, and Aluminum Nitride is capable of transduction up to 8.65x x10 20 n/cm2.

Reactor Safety Gap Evaluation of Accident Tolerant Components and Severe Accident Analysis

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

Farmer, Mitchell T.; Bunt, R.; Corradini, M.

The overall objective of this study was to conduct a technology gap evaluation on accident tolerant components and severe accident analysis methodologies with the goal of identifying any data and/or knowledge gaps that may exist, given the current state of light water reactor (LWR) severe accident research, and additionally augmented by insights obtained from the Fukushima accident. The ultimate benefit of this activity is that the results can be used to refine the Department of Energy’s (DOE) Reactor Safety Technology (RST) research and development (R&D) program plan to address key knowledge gaps in severe accident phenomena and analyses that affectmore » reactor safety and that are not currently being addressed by the industry or the Nuclear Regulatory Commission (NRC).« less

Flow reversal power limit for the HFBR

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

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

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

R and D program for core instrumentation improvements devoted for French sodium fast reactors

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

Jeannot, J. P.; Rodriguez, G.; Jammes, C.

2011-07-01

Under the framework of French R and D studies for Generation IV reactors and more specifically for sodium-cooled fast reactors (SFR); the CEA, EDF and AREVA have launched a joint coordinated research programme. This paper deals with the R and D sets out to achieve better inspection, maintenance, availability and decommissioning. In particular the instrumentation requirements for core monitoring and detection in the case of accidental events. Requirements mainly involve diversifying the means of protection and improving instrumentation performance in terms of responsiveness and sensitivity. Operation feedback from the Phenix and Superphenix prototype reactors and studies, carried out within themore » scope of the EFR projects, has been used to define the needs for instrumentation enhancement. (authors)« less

Interim waste storage for the Integral Fast Reactor

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

Benedict, R.W.; Phipps, R.D.; Condiff, D.W.

1991-01-01

The Integral Fast Reactor (IFR), which Argonne National Laboratory is developing, is an innovative liquid metal breeder reactor that uses metallic fuel and has a close coupled fuel recovery process. A pyrochemical process is used to separate the fission products from the actinide elements. These actinides are used to make new fuel for the reactor. As part of the overall IFR development program, Argonne has refurbished an existing Fuel Cycle Facility at ANL-West and is installing new equipment to demonstrate the remote reprocessing and fabrication of fuel for the Experimental Breeder Reactor II (EBR-II). During this demonstration the wastes thatmore » are produced will be treated and packaged to produce waste forms that would be typical of future commercial operations. These future waste forms would, assuming Argonne development goals are fulfilled, be essentially free of long half-life transuranic isotopes. Promising early results indicate that actinide extraction processes can be developed to strip these isotopes from waste stream and return them to the IFR type reactors for fissioning. 1 fig.« less

Thermal-Mechanical Stress Analysis of PWR Pressure Vessel and Nozzles under Grid Load-Following Mode: Interim Report on the Effect of Cyclic Hardening Material Properties and Pre-existing Cracks on Stress Analysis Results

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

Mohanty, Subhasish; Soppet, William; Majumdar, Saurin

This report provides an update on an assessment of environmentally assisted fatigue for light water reactor components under extended service conditions. This report is a deliverable under the work package for environmentally assisted fatigue as part of DOE’s Light Water Reactor Sustainability Program. In a previous report (September 2015), we presented tensile and fatigue test data and related hardening material properties for 508 low-alloys steel base metal and other reactor metals. In this report, we present thermal-mechanical stress analysis of the reactor pressure vessel and its hot-leg and cold-leg nozzles based on estimated material properties. We also present results frommore » thermal and thermal-mechanical stress analysis under reactor heat-up, cool-down, and grid load-following conditions. Analysis results are given with and without the presence of preexisting cracks in the reactor nozzles (axial or circumferential crack). In addition, results from validation stress analysis based on tensile and fatigue experiments are reported.« less

Alternative nuclear technologies

NASA Astrophysics Data System (ADS)

Schubert, E.

1981-10-01

The lead times required to develop a select group of nuclear fission reactor types and fuel cycles to the point of readiness for full commercialization are compared. Along with lead times, fuel material requirements and comparative costs of producing electric power were estimated. A conservative approach and consistent criteria for all systems were used in estimates of the steps required and the times involved in developing each technology. The impact of the inevitable exhaustion of the low- or reasonable-cost uranium reserves in the United States on the desirability of completing the breeder reactor program, with its favorable long-term result on fission fuel supplies, is discussed. The long times projected to bring the most advanced alternative converter reactor technologies the heavy water reactor and the high-temperature gas-cooled reactor into commercial deployment when compared to the time projected to bring the breeder reactor into equivalent status suggest that the country's best choice is to develop the breeder. The perceived diversion-proliferation problems with the uranium plutonium fuel cycle have workable solutions that can be developed which will enable the use of those materials at substantially reduced levels of diversion risk.

Analysis of fluid fuel flow to the neutron kinetics on molten salt reactor FUJI-12

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

Aji, Indarta Kuncoro, E-mail: indartaaji@s.itb.ac.id; Waris, Abdul, E-mail: awaris@fi.itb.ac.id; Permana, Sidik

Molten Salt Reactor is a reactor are operating with molten salt fuel flowing. This condition interpret that the neutron kinetics of this reactor is affected by the flow rate of the fuel. This research analyze effect by the alteration velocity of the fuel by MSR type Fuji-12, with fuel composition LiF-BeF{sub 2}-ThF{sub 4}-{sup 233}UF{sub 4} respectively 71.78%-16%-11.86%-0.36%. Calculation process in this study is performed numerically by SOR and finite difference method use C programming language. Data of reactivity, neutron flux, and the macroscopic fission cross section for calculation process obtain from SRAC-CITATION (Standard thermal Reactor Analysis Code) and JENDL-4.0 datamore » library. SRAC system designed and developed by JAEA (Japan Atomic Energy Agency). This study aims to observe the effect of the velocity of fuel salt to the power generated from neutron precursors at fourth year of reactor operate (last critical condition) with number of multiplication effective; 1.0155.« less

Emissivity of Candidate Materials for VHTR Applicationbs: Role of Oxidation and Surface Modification Treatments

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

Sridharan, Kumar; Allen, Todd; Anderson, Mark

The Generation IV (GEN IV) Nuclear Energy Systems Initiative was instituted by the Department of Energy (DOE) with the goal of researching and developing technologies and materials necessary for various types of future reactors. These GEN IV reactors will employ advanced fuel cycles, passive safety systems, and other innovative systems, leading to significant differences between these future reactors and current water-cooled reactors. The leading candidate for the Next Generation Nuclear Plant (NGNP) to be built at Idaho National Lab (INL) in the United States is the Very High Temperature Reactor (VHTR). Due to the high operating temperatures of the VHTR,more » the Reactor Pressure Vessel (RPV) will partially rely on heat transfer by radiation for cooling. Heat expulsion by radiation will become all the more important during high temperature excursions during off-normal accident scenarios. Radiant power is dictated by emissivity, a material property. The NGNP Materials Research and Development Program Plan [1] has identified emissivity and the effects of high temperature oxide formation on emissivity as an area of research towards the development of the VHTR.« less

Development of ORIGEN Libraries for Mixed Oxide (MOX) Fuel Assembly Designs

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

Mertyurek, Ugur; Gauld, Ian C.

In this research, ORIGEN cross section libraries for reactor-grade mixed oxide (MOX) fuel assembly designs have been developed to provide fast and accurate depletion calculations to predict nuclide inventories, radiation sources and thermal decay heat information needed in safety evaluations and safeguards verification measurements of spent nuclear fuel. These ORIGEN libraries are generated using two-dimensional lattice physics assembly models that include enrichment zoning and cross section data based on ENDF/B-VII.0 evaluations. Using the SCALE depletion sequence, burnup-dependent cross sections are created for selected commercial reactor assembly designs and a representative range of reactor operating conditions, fuel enrichments, and fuel burnup.more » The burnup dependent cross sections are then interpolated to provide problem-dependent cross sections for ORIGEN, avoiding the need for time-consuming lattice physics calculations. The ORIGEN libraries for MOX assembly designs are validated against destructive radiochemical assay measurements of MOX fuel from the MALIBU international experimental program. This program included measurements of MOX fuel from a 15 × 15 pressurized water reactor assembly and a 9 × 9 boiling water reactor assembly. The ORIGEN MOX libraries are also compared against detailed assembly calculations from the Phase IV-B numerical MOX fuel burnup credit benchmark coordinated by the Nuclear Energy Agency within the Organization for Economic Cooperation and Development. Finally, the nuclide compositions calculated by ORIGEN using the MOX libraries are shown to be in good agreement with other physics codes and with experimental data.« less

Space Nuclear Facility test capability at the Baikal-1 and IGR sites Semipalatinsk-21, Kazakhstan

NASA Astrophysics Data System (ADS)

Hill, T. J.; Stanley, M. L.; Martinell, J. S.

1993-01-01

The International Space Technology Assessment Program was established 1/19/92 to take advantage of the availability of Russian space technology and hardware. DOE had two delegations visit CIS and assess its space nuclear power and propulsion technologies. The visit coincided with the Conference on Nuclear Power Engineering in Space Nuclear Rocket Engines at Semipalatinsk-21 (Kurchatov, Kazakhstan) on Sept. 22-25, 1992. Reactor facilities assessed in Semipalatinski-21 included the IVG-1 reactor (a nuclear furnace, which has been modified and now called IVG-1M), the RA reactor, and the Impulse Graphite Reactor (IGR), the CIS version of TREAT. Although the reactor facilities are being maintained satisfactorily, the support infrastructure appears to be degrading. The group assessment is based on two half-day tours of the Baikals-1 test facility and a brief (2 hr) tour of IGR; because of limited time and the large size of the tour group, it was impossible to obtain answers to all prepared questions. Potential benefit is that CIS fuels and facilities may permit USA to conduct a lower priced space nuclear propulsion program while achieving higher performance capability faster, and immediate access to test facilities that cannot be available in this country for 5 years. Information needs to be obtained about available data acquisition capability, accuracy, frequency response, and number of channels. Potential areas of interest with broad application in the U.S. nuclear industry are listed.

Development of ORIGEN Libraries for Mixed Oxide (MOX) Fuel Assembly Designs

DOE PAGES

Mertyurek, Ugur; Gauld, Ian C.

2015-12-24

In this research, ORIGEN cross section libraries for reactor-grade mixed oxide (MOX) fuel assembly designs have been developed to provide fast and accurate depletion calculations to predict nuclide inventories, radiation sources and thermal decay heat information needed in safety evaluations and safeguards verification measurements of spent nuclear fuel. These ORIGEN libraries are generated using two-dimensional lattice physics assembly models that include enrichment zoning and cross section data based on ENDF/B-VII.0 evaluations. Using the SCALE depletion sequence, burnup-dependent cross sections are created for selected commercial reactor assembly designs and a representative range of reactor operating conditions, fuel enrichments, and fuel burnup.more » The burnup dependent cross sections are then interpolated to provide problem-dependent cross sections for ORIGEN, avoiding the need for time-consuming lattice physics calculations. The ORIGEN libraries for MOX assembly designs are validated against destructive radiochemical assay measurements of MOX fuel from the MALIBU international experimental program. This program included measurements of MOX fuel from a 15 × 15 pressurized water reactor assembly and a 9 × 9 boiling water reactor assembly. The ORIGEN MOX libraries are also compared against detailed assembly calculations from the Phase IV-B numerical MOX fuel burnup credit benchmark coordinated by the Nuclear Energy Agency within the Organization for Economic Cooperation and Development. Finally, the nuclide compositions calculated by ORIGEN using the MOX libraries are shown to be in good agreement with other physics codes and with experimental data.« less

Nuclear space power safety and facility guidelines study

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

Mehlman, W.F.

1995-09-11

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

Prospective Activities outlined for Regulatory Approval in Ghana Overview

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

Abrefah, R.G.; Odoi, H.C.; Mo, S.C.

The Ghana Research Reactor-1 (GHARR-1) is one of Chinese’s Miniature Neutron Source Reactor (MNSR) which was purchased under a tripartite agreement between Ghana, China and the IAEA. The reactor was installed in 1994 and has since been in operation without any incident. It has been used chiefly for Neutron Activation Analysis (NAA) and Training of students in the field of Nuclear Engineering. The GHARR-1 has been earmarked for the Conversion of Core from HEU to LEU which is in accordance with the GTRI program and other related and/or associated programs. Over the past few years the National Nuclear Research Institutemore » (NNRI), the Operating Organization of the Research Reactor for the Ghana Atomic Energy Commission (GAEC), has undertaken various tasks in order to implement the replacement of the reactor core. After completion, of the neutronic calculations, results showed that that an LEU fuel of 12.5% enrichment was desirable. However, recent developments have shown that an LEU fuel with 13% enrichment will be fabricated by the manufacturers, which is captured in a fuel specification document sent to NNRI by the CIAE. It is therefore imperative that all neutronic and thermal hydraulic calculation be done again to help acquire regulatory approval. Furthermore, the radiation exposure to personnel involved in the conversion must be estimated to help convince our regulators. This paper outlines the processes and activities that will enable us meet regulatory requirements.« less

(Boiling water reactor (BWR) CORA experiments)

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

Ott, L.J.

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

Operational Philosophy for the Advanced Test Reactor National Scientific User Facility

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

J. Benson; J. Cole; J. Jackson

2013-02-01

In 2007, the Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF). At its core, the ATR NSUF Program combines access to a portion of the available ATR radiation capability, the associated required examination and analysis facilities at the Idaho National Laboratory (INL), and INL staff expertise with novel ideas provided by external contributors (universities, laboratories, and industry). These collaborations define the cutting edge of nuclear technology research in high-temperature and radiation environments, contribute to improved industry performance of current and future light-water reactors (LWRs), and stimulate cooperative research between user groupsmore » conducting basic and applied research. To make possible the broadest access to key national capability, the ATR NSUF formed a partnership program that also makes available access to critical facilities outside of the INL. Finally, the ATR NSUF has established a sample library that allows access to pre-irradiated samples as needed by national research teams.« less

Global threat reduction initiative Russian nuclear material removal progress

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

Cummins, Kelly; Bolshinsky, Igor

2008-07-15

In December 1999 representatives from the United States, the Russian Federation, and the International Atomic Energy Agency (IAEA) started discussing a program to return to Russia Soviet- or Russian-supplied highly enriched uranium (HEU) fuel stored at the Russian-designed research reactors outside Russia. Trilateral discussions among the United States, Russian Federation, and the International Atomic Energy Agency (IAEA) have identified more than 20 research reactors in 17 countries that have Soviet- or Russian-supplied HEU fuel. The Global Threat Reduction Initiative's Russian Research Reactor Fuel Return Program is an important aspect of the U.S. Government's commitment to cooperate with the other nationsmore » to prevent the proliferation of nuclear weapons and weapons-usable proliferation-attractive nuclear materials. To date, 496 kilograms of Russian-origin HEU have been shipped to Russia from Serbia, Latvia, Libya, Uzbekistan, Romania, Bulgaria, Poland, Germany, and the Czech Republic. The pilot spent fuel shipment from Uzbekistan to Russia was completed in April 2006. (author)« less

Fusion Studies in Japan

NASA Astrophysics Data System (ADS)

Ogawa, Yuichi

2016-05-01

A new strategic energy plan decided by the Japanese Cabinet in 2014 strongly supports the steady promotion of nuclear fusion development activities, including the ITER project and the Broader Approach activities from the long-term viewpoint. Atomic Energy Commission (AEC) in Japan formulated the Third Phase Basic Program so as to promote an experimental fusion reactor project. In 2005 AEC has reviewed this Program, and discussed on selection and concentration among many projects of fusion reactor development. In addition to the promotion of ITER project, advanced tokamak research by JT-60SA, helical plasma experiment by LHD, FIREX project in laser fusion research and fusion engineering by IFMIF were highly prioritized. Although the basic concept is quite different between tokamak, helical and laser fusion researches, there exist a lot of common features such as plasma physics on 3-D magnetic geometry, high power heat load on plasma facing component and so on. Therefore, a synergetic scenario on fusion reactor development among various plasma confinement concepts would be important.

Effects of thermal annealing and reirradiation on toughness of reactor pressure vessel steels

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

Nanstad, R.K.; Iskander, S.K.; Sokolov, M.A.

1997-02-01

One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPV) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. This paper summarizes recent experimental results from work performed at the Oak Ridge National Laboratory (ORNL) to study the annealing response, or {open_quotes}recovery,{close_quotes} of several irradiated RPV steels; it also includes recent results from both ORNL and the Russian Research Center-Kurchatov Institute (RRC-KI) on a cooperative program of irradiation, annealing and reirradiation of both U.S. and Russian RPV steels. The cooperative program was conducted under the auspices ofmore » Working Group 3, U.S./Russia Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS). The materials investigated are an RPV plate and various submerged-arc welds, with tensile, Charpy impact toughness, and fracture toughness results variously determined. Experimental results are compared with applicable prediction guidelines, while observed differences in annealing responses and reirradiation rates are discussed.« less

Manned space flight nuclear system safety. Volume 3: Reactor system preliminary nuclear safety analysis. Part 2: Accident Model Document (AMD)

NASA Technical Reports Server (NTRS)

1972-01-01

The Accident Model Document is one of three documents of the Preliminary Safety Analysis Report (PSAR) - Reactor System as applied to a Space Base Program. Potential terrestrial nuclear hazards involving the zirconium hydride reactor-Brayton power module are identified for all phases of the Space Base program. The accidents/events that give rise to the hazards are defined and abort sequence trees are developed to determine the sequence of events leading to the hazard and the associated probabilities of occurence. Source terms are calculated to determine the magnitude of the hazards. The above data is used in the mission accident analysis to determine the most probable and significant accidents/events in each mission phase. The only significant hazards during the prelaunch and launch ascent phases of the mission are those which arise form criticality accidents. Fission product inventories during this time period were found to be very low due to very limited low power acceptance testing.

Exposure calculation code module for reactor core analysis: BURNER

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

Vondy, D.R.; Cunningham, G.W.

1979-02-01

The code module BURNER for nuclear reactor exposure calculations is presented. The computer requirements are shown, as are the reference data and interface data file requirements, and the programmed equations and procedure of calculation are described. The operating history of a reactor is followed over the period between solutions of the space, energy neutronics problem. The end-of-period nuclide concentrations are determined given the necessary information. A steady state, continuous fueling model is treated in addition to the usual fixed fuel model. The control options provide flexibility to select among an unusually wide variety of programmed procedures. The code also providesmore » user option to make a number of auxiliary calculations and print such information as the local gamma source, cumulative exposure, and a fine scale power density distribution in a selected zone. The code is used locally in a system for computation which contains the VENTURE diffusion theory neutronics code and other modules.« less

Nuclear power plant 5,000 to 10,000 kilowatts

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

Not Available

The purpose of this proposal is to present a suggested program for the development of an Aqueous Homogeneous Reactor Power Plant for the production of power in the 5000 to 10,000 kilowatt range under the terms of the Atomic Energy Commission's invitation of September 21, 1955. It envisions a research and development program prior to finalizing fabricating commitments of full scale components for the purpose of proving mechanical and hydraulic operating and chemical processing feasibility with the expectation that such preliminary effort will assure the contruction of the reactor at the lowest cost and successful operation at the earliest date.more » It proposes the construction of a reactor for an eventual net electrical output of ten megawatts but initially in conjunction with a five megawatt turbo-generating unit. This unit would be constructed at the site of the existing Hersey diesel generating plant of the Wolverine Electric Cooperative approximately ten miles north of Big Rapids, Michigan.« less

Evaluation of COBRA III-C and SABRE-I (wire wrap version) computational results by comparison with steady-state data from a 19-pin internally guard heated sodium cooled bundle with a six-channel central blockage (THORS bundle 3C). [LMFBR

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

Dearing, J F; Rose, S D; Nelson, W R

The predicted computational results of two well-known sub-channel analysis codes, COBRA-III-C and SABRE-I (wire wrap version), have been evaluated by comparison with steady state temperature data from the THORS Facility at ORNL. Both codes give good predictions of transverse and axial temperatures when compared with wire wrap thermocouple data. The crossflow velocity profiles predicted by these codes are similar which is encouraging since the wire wrap models are based on different assumptions.

High-Temperature Fluid-Wall Reactor Technology Research, Test and Evaluation Performed at Naval Construction Battalion Center, Gulfport, Mississippi, for the United States Air Force Installation/Restoration Program

DTIC Science & Technology

1988-01-01

the reactor Duties: The Process Engineers rotate with the Lead Operator to monitor the process at the top of the reactor through the site glass...pant cuffs and coverhoods of coveralls, will be attached to gloves, boots and coveralls, using duct tape. * IF AMBIENT WORK STATIONS TEMPERATURE IS...L of the sample fortification solution (Section ýý8) containing 1C 12-2,3,7,8-TCDD at a concentration of 0.5 ng/1,Land C14-2,3,7,8-TCDD at a

DYNAMIC AND STATIC PARAMETERS OF THE AQUEOUS HOMOGENEOUS ARMOUR RESEARCH REACTOR

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

Terrell, C.W.; McElroy, W.N.

1959-06-01

A brief description of the aqueous homogeneous Armour Research Reactor is given. The negative reactivity coefficient resulting from a temperature increase was determined over a fuel temperature range of 37 to 150 deg F. Possession of an accurately calibrated rod and temperature coefficient permitted a direct measurement of the void coefficient. The reactor was taken to different power levels, and from the calibrated rod the total reduction in excess reactivity was obtained. During the power increase program additional U/sup 235/ and water were added to the core to determine the worth of U/sup 235/ and water. (W.D.M.)

Advances in modelling of condensation phenomena

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

Liu, W.S.; Zaltsgendler, E.; Hanna, B.

1997-07-01

The physical parameters in the modelling of condensation phenomena in the CANDU reactor system codes are discussed. The experimental programs used for thermal-hydraulic code validation in the Canadian nuclear industry are briefly described. The modelling of vapour generation and in particular condensation plays a key role in modelling of postulated reactor transients. The condensation models adopted in the current state-of-the-art two-fluid CANDU reactor thermal-hydraulic system codes (CATHENA and TUF) are described. As examples of the modelling challenges faced, the simulation of a cold water injection experiment by CATHENA and the simulation of a condensation induced water hammer experiment by TUFmore » are described.« less

MTR BUILDING, TRA603. EAST SIDE. CAMERA FACING WEST. CORRUGATED IRON ...

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

MTR BUILDING, TRA-603. EAST SIDE. CAMERA FACING WEST. CORRUGATED IRON BUILDING MARKED WITH "X" IS TRA-651. TRA-626, TO ITS RIGHT, HOUSED COMPRESSOR EQUIPMENT FOR THE AIRCRAFT NUCLEAR PROPULSION PROGRAM. LATER, IT WAS USED FOR STORAGE. INL NEGATIVE NO. HD46-42-4. Mike Crane, Photographer, April 2005 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

PWR PRELIMINARY DESIGN FOR PL-3

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

Humphries, G. E.

1962-02-28

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

Advanced In-Pile Instrumentation for Materials Testing Reactors

NASA Astrophysics Data System (ADS)

Rempe, J. L.; Knudson, D. L.; Daw, J. E.; Unruh, T. C.; Chase, B. M.; Davis, K. L.; Palmer, A. J.; Schley, R. S.

2014-08-01

The U.S. Department of Energy sponsors the Advanced Test Reactor (ATR) National Scientific User Facility (NSUF) program to promote U.S. research in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, advancing U.S. energy security needs. A key component of the ATR NSUF effort is to design, develop, and deploy new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. This paper describes the strategy developed by the Idaho National Laboratory (INL) for identifying instrumentation needed for ATR irradiation tests and the program initiated to obtain these sensors. New sensors developed from this effort are identified, and the progress of other development efforts is summarized. As reported in this paper, INL researchers are currently involved in several tasks to deploy real-time length and flux detection sensors, and efforts have been initiated to develop a crack growth test rig. Tasks evaluating `advanced' technologies, such as fiber-optics based length detection and ultrasonic thermometers, are also underway. In addition, specialized sensors for real-time detection of temperature and thermal conductivity are not only being provided to NSUF reactors, but are also being provided to several international test reactors.

Science in Flux: NASA's Nuclear Program at Plum Brook Station 1955-2005

NASA Technical Reports Server (NTRS)

Bowles, Mark D.

2006-01-01

Science in Flux traces the history of one of the most powerful nuclear test reactors in the United States and the only nuclear facility ever built by NASA. In the late 1950's NASA constructed Plum Brook Station on a vast tract of undeveloped land near Sandusky, Ohio. Once fully operational in 1963, it supported basic research for NASA's nuclear rocket program (NERVA). Plum Brook represents a significant, if largely forgotten, story of nuclear research, political change, and the professional culture of the scientists and engineers who devoted their lives to construct and operate the facility. In 1973, after only a decade of research, the government shut Plum Brook down before many of its experiments could be completed. Even the valiant attempt to redefine the reactor as an environmental analysis tool failed, and the facility went silent. The reactors lay in costly, but quiet standby for nearly a quarter-century before the Nuclear Regulatory Commission decided to decommission the reactors and clean up the site. The history of Plum Brook reveals the perils and potentials of that nuclear technology. As NASA, Congress, and space enthusiasts all begin looking once again at the nuclear option for sending humans to Mars, the echoes of Plum Brook's past will resonate with current policy and space initiatives.

The Atomic Energy Commission's Annual Report to Congress for 1962. Major Activities in the Atomic Energy Programs, January - December 1962

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

Seaborg, Glenn T.

1963-01-31

The document represents the 1962 Annual Report of the Atomic Energy Commission (AEC) to Congress. This year's report opens with a section of Highlights of the Atomic Energy Programs of 1962, followed by five parts: Part One, Commission Activities; Part Two, Nuclear Reactor Programs; Part Three, Production and Weapons Programs; Part Four, Other Major Programs; and Part Five, The Regulatory Program. Sixteen appendices are also included.

Lessons Learned From Developing Reactor Pressure Vessel Steel Embrittlement Database

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

Wang, Jy-An John

Materials behaviors caused by neutron irradiation under fission and/or fusion environments can be little understood without practical examination. Easily accessible material information system with large material database using effective computers is necessary for design of nuclear materials and analyses or simulations of the phenomena. The developed Embrittlement Data Base (EDB) at ORNL is this comprehensive collection of data. EDB database contains power reactor pressure vessel surveillance data, the material test reactor data, foreign reactor data (through bilateral agreements authorized by NRC), and the fracture toughness data. The lessons learned from building EDB program and the associated database management activity regardingmore » Material Database Design Methodology, Architecture and the Embedded QA Protocol are described in this report. The development of IAEA International Database on Reactor Pressure Vessel Materials (IDRPVM) and the comparison of EDB database and IAEA IDRPVM database are provided in the report. The recommended database QA protocol and database infrastructure are also stated in the report.« less

Multiscale Evaluation of Catalytic Upgrading of Biomass Pyrolysis Vapors on Ni- and Ga-Modified ZSM-5

DOE PAGES

Yung, Matthew M.; Stanton, Alexander R.; Iisa, Kristiina; ...

2016-10-07

Metal-impregnated (Ni or Ga) ZSM-5 catalysts were studied for biomass pyrolysis vapor upgrading to produce hydrocarbons using three reactors constituting a 100 000x change in the amount of catalyst used in experiments. Catalysts were screened for pyrolysis vapor phase upgrading activity in two small-scale reactors: (i) a Pyroprobe with a 10 mg catalyst in a fixed bed and (ii) a fixed-bed reactor with 500 mg of catalyst. The best performing catalysts were then validated with a larger scale fluidized-bed reactor (using ~1 kg of catalyst) that produced measurable quantities of bio-oil for analysis and evaluation of mass balances. Despite somemore » inherent differences across the reactor systems (such as residence time, reactor type, analytical techniques, mode of catalyst and biomass feed) there was good agreement of reaction results for production of aromatic hydrocarbons, light gases, and coke deposition. Relative to ZSM-5, Ni or Ga addition to ZSM-5 increased production of fully deoxygenated aromatic hydrocarbons and light gases. In the fluidized bed reactor, Ga/ZSM-5 slightly enhanced carbon efficiency to condensed oil, which includes oxygenates in addition to aromatic hydrocarbons, and reduced oil oxygen content compared to ZSM-5. Ni/ZSM-5, while giving the highest yield of fully deoxygenated aromatic hydrocarbons, gave lower overall carbon efficiency to oil but with the lowest oxygen content. Reaction product analysis coupled with fresh and spent catalyst characterization indicated that the improved performance of Ni/ZSM-5 is related to decreasing deactivation by coking, which keeps the active acid sites accessible for the deoxygenation and aromatization reactions that produce fully deoxygenated aromatic hydrocarbons. The addition of Ga enhances the dehydrogenation activity of the catalyst, which leads to enhanced olefin formation and higher fully deoxygenated aromatic hydrocarbon yields compared to unmodified ZSM-5. Catalyst characterization by ammonia temperature programmed desorption, surface area measurements, and postreaction temperature-programmed oxidation (TPO) also showed that the metal-modified zeolites retained a greater percentage of their initial acidity and surface area, which was consistent between the reactor scales. These results demonstrate that the trends observed with smaller (milligram to gram) catalyst reactors are applicable to larger, more industrially relevant (kg) scales to help guide catalyst research toward application.« less

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

Unwin, Stephen D.; Lowry, Peter P.; Layton, Robert F.

This is a working report drafted under the Risk-Informed Safety Margin Characterization pathway of the Light Water Reactor Sustainability Program, describing statistical models of passives component reliabilities.

Evaluation of Nuclear Facility Decommissioning Projects program: a reference test reactor. Project summary report

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

Boing, L.E.; Miller, R.L.

1983-10-01

This document presents, in summary form, generic conceptual information relevant to the decommissioning of a reference test reactor (RTR). All of the data presented were extracted from NUREG/CR-1756 and arranged in a form that will provide a basis for future comparison studies for the Evaluation of Nuclear Facility Decommissioning Projects (ENFDP) program. During the data extraction process no attempt was made to challenge any of the assumptions used in the original studies nor was any attempt made to update assumed methods or processes to state-of-the-art decommissioning techniques. In a few instances obvious errors were corrected after consultation with the studymore » author.« less

Screening of redox couples and electrode materials

NASA Technical Reports Server (NTRS)

Giner, J.; Swette, L.; Cahill, K.

1976-01-01

Electrochemical parameters of selected redox couples that might be potentially promising for application in bulk energy storage systems were investigated. This was carried out in two phases: a broad investigation of the basic characteristics and behavior of various redox couples, followed by a more limited investigation of their electrochemical performance in a redox flow reactor configuration. In the first phase of the program, eight redox couples were evaluated under a variety of conditions in terms of their exchange current densities as measured by the rotating disk electrode procedure. The second phase of the program involved the testing of four couples in a redox reactor under flow conditions with a varity of electrode materials and structures.

Large-break LOCA, in-reactor fuel bundle Materials Test MT-6A

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

Wilson, C.L.; Hesson, G.M.; Pilger, J.P.

1993-09-01

This is a report on one of a series of experiments to simulates a loss-of-coolant accident (LOCA) using full-length fuel rods for pressurized water reactors (PWR). The experiments were conducted by Pacific Northwest Laboratory (PNL) under the LOCA simulation Program sponsored by the US Nuclear Regulatory Commission (NRC). The major objective of this program was causing the maximum possible expansion of the cladding on the fuel rods from a short-term adiabatic temperature transient to 1200 K (1700 F) leading to the rupture of the cladding; and second, by reflooding the fuel rods to determine the rate at which the fuelmore » bundle is cooled.« less

Basic elements of light water reactor fuel rod design. [FUELROD code

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

Weisman, J.; Eckart, R.

1981-06-01

Basic design techniques and equations are presented to allow students to understand and perform preliminary fuel design for normal reactor conditions. Each of the important design considerations is presented and discussed in detail. These include the interaction between fuel pellets and cladding and the changes in fuel and cladding that occur during the operating lifetime of the fuel. A simple, student-oriented, fuel rod design computer program, called FUELROD, is described. The FUELROD program models the in-pile pellet cladding interaction and allows a realistic exploration of the effect of various design parameters. By use of FUELROD, the student can gain anmore » appreciation of the fuel rod design process. 34 refs.« less

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

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

None

2013-09-25

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

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

ScienceCinema

None

2018-01-16

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

RERTR 2009 (Reduced Enrichment for Research and Test Reactors)

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

Totev, T.; Stevens, J.; Kim, Y. S.

2010-03-01

The U.S. Department of Energy/National Nuclear Security Administration's Office of Global Threat Reduction in cooperation with the China Atomic Energy Authority and International Atomic Energy Agency hosted the 'RERTR 2009 International Meeting on Reduced Enrichment for Research and Test Reactors.' The meeting was organized by Argonne National Laboratory, China Institute of Atomic Energy and Idaho National Laboratory and was held in Beijing, China from November 1-5, 2009. This was the 31st annual meeting in a series on the same general subject regarding the conversion of reactors within the Global Threat Reduction Initiative (GTRI). The Reduced Enrichment for Research and Testmore » Reactors (RERTR) Program develops technology necessary to enable the conversion of civilian facilities using high enriched uranium (HEU) to low enriched uranium (LEU) fuels and targets.« less

Improvement of yields and rates during enzymatic hydrolysis of cellulose to glucose. Progress report, March 1, 1979-May 31, 1979

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

Sundstrom, D W; Klei, H E; Coughlin, R W

1979-05-01

The objective of this program is to show that the conversion of cellulose to glucose can be significantly increased by enzymatically removing the inhibitory cellobiose from the reaction system using immobilized ..beta..-glucosidase (..beta..-G). An enzymatic catalyst was prepared and used in a fluidized bed with cellobiose as the substrate, only a 10% loss of activity was observed during a 500 hour period. Cellulose was hydrolyzed in two batch reactors operated side-by-side, with one reactor containing immobilized ..beta..-G and cellulose and the other reactor containing an equal amount of cellulose only. After 30 hours the reactor containing the immobilized ..beta..-G hadmore » 100% more glucose, indicating that the catalytic removal of the cellobiose had a significant effect upon the production of glucose.« less

Development of variable-width ribbon heating elements for liquid-metal and gas-cooled fast breeder reactor fuel-pin simulators

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

McCulloch, R.W.; Post, D.W.; Lovell, R.T.

1981-04-01

Variable-width ribbon heating elements that provide a chopped-cosine variable heat flux profile have been fabricated for fuel pin simulators used in test loops by the Breeder Reactor Program Thermal-Hydraulic Out-of-Reactor Safety test facility and the Gas-Cooled Fast Breeder Reactor-Core Flow Test Loop. Thermal, mechanical, and electrical design considerations are used to derive an analytical expression that precisely describes ribbon contour in terms of the major fabrication parameters. These parameters are used to generate numerical control tapes that control ribbon cutting and winding machines. Infrared scanning techniques are developed to determine the optimum transient thermal profile of the coils and relatemore » this profile to that generated by the coils in completed fuel pin simulators.« less

The Experimental Breeder Reactor II seismic probabilistic risk assessment

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

Roglans, J; Hill, D J

1994-02-01

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

Monolithic catalyst beds for hydrazine reactors

NASA Technical Reports Server (NTRS)

1973-01-01

A monolithic catalyst bed for monopropellant hydrazine decomposition was evaluated. The program involved the evaluation of a new hydrazine catalyst concept wherein open-celled foamed materials are used as supports for the active catalysts. A high-surface-area material is deposited upon the open-celled foamed material and is then coated with an active metal to provide a spontaneous catalyst. Only a fraction of the amount of expensive active metal in currently available catalysts is needed to promote monolithic catalyst. Numerous parameters were evaluated during the program, and the importance of additional parameters became obvious only while the program was in progress. A demonstration firing (using a 2.2-Newton (N)(0.5-lbf) reactor) successfully accumulated 7,700 seconds of firing time and 16 ambient temperature starts without degradation. Based on the excellent results obtained throughout the program and the demonstrated life capability of the monolithic foam, it is recommended that additional studies be conducted to further exploit the advantages of this concept.

The role of nuclear reactors in space exploration and development

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

Lipinski, R.J.

2000-07-01

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

The Advanced Test Reactor National Scientific User Facility Advancing Nuclear Technology

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

T. R. Allen; J. B. Benson; J. A. Foster

2009-05-01

To help ensure the long-term viability of nuclear energy through a robust and sustained research and development effort, the U.S. Department of Energy (DOE) designated the Advanced Test Reactor and associated post-irradiation examination facilities a National Scientific User Facility (ATR NSUF), allowing broader access to nuclear energy researchers. The mission of the ATR NSUF is to provide access to world-class nuclear research facilities, thereby facilitating the advancement of nuclear science and technology. The ATR NSUF seeks to create an engaged academic and industrial user community that routinely conducts reactor-based research. Cost free access to the ATR and PIE facilities ismore » granted based on technical merit to U.S. university-led experiment teams conducting non-proprietary research. Proposals are selected via independent technical peer review and relevance to DOE mission. Extensive publication of research results is expected as a condition for access. During FY 2008, the first full year of ATR NSUF operation, five university-led experiments were awarded access to the ATR and associated post-irradiation examination facilities. The ATR NSUF has awarded four new experiments in early FY 2009, and anticipates awarding additional experiments in the fall of 2009 as the results of the second 2009 proposal call. As the ATR NSUF program mature over the next two years, the capability to perform irradiation research of increasing complexity will become available. These capabilities include instrumented irradiation experiments and post-irradiation examinations on materials previously irradiated in U.S. reactor material test programs. The ATR critical facility will also be made available to researchers. An important component of the ATR NSUF an education program focused on the reactor-based tools available for resolving nuclear science and technology issues. The ATR NSUF provides education programs including a summer short course, internships, faculty-student team projects and faculty/staff exchanges. In June of 2008, the first week-long ATR NSUF Summer Session was attended by 68 students, university faculty and industry representatives. The Summer Session featured presentations by 19 technical experts from across the country and covered topics including irradiation damage mechanisms, degradation of reactor materials, LWR and gas reactor fuels, and non-destructive evaluation. High impact research results from leveraging the entire research infrastructure, including universities, industry, small business, and the national laboratories. To increase overall research capability, ATR NSUF seeks to form strategic partnerships with university facilities that add significant nuclear research capability to the ATR NSUF and are accessible to all ATR NSUF users. Current partner facilities include the MIT Reactor, the University of Michigan Irradiated Materials Testing Laboratory, the University of Wisconsin Characterization Laboratory, and the University of Nevada, Las Vegas transmission Electron Microscope User Facility. Needs for irradiation of material specimens at tightly controlled temperatures are being met by dedication of a large in-pile pressurized water loop facility for use by ATR NSUF users. Several environmental mechanical testing systems are under construction to determine crack growth rates and fracture toughness on irradiated test systems.« less

Unique educational opportunities at the Missouri University research reactor

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

Ketring, A.R.; Ross, F.K.; Spate, V.

1997-12-01

Since the Missouri University Research Reactor (MURR) went critical in 1966, it has been a center where students from many departments conduct their graduate research. In the past three decades, hundreds of graduate students from the MU departments of chemistry, physics, anthropology, nuclear engineering, etc., have received masters and doctoral degrees based on research using neutrons produced at MURR. More recently, the educational opportunities at MURR have been expanded to include undergraduate students and local high school students. Since 1989 MURR has participated in the National Science Foundation-funded Research Experience for Undergraduates (REU) program. As part of this program, undergraduatemore » students from universities and colleges throughout the United States come to MURR and get hands-on research experience during the summer. Another program, started in 1994 by the Nuclear Analysis Program at MURR, allows students from a local high school to conduct a neutron activation analysis (NAA) experiment. We also conduct tours of the center, where we describe the research and educational programs at MURR to groups of elementary school children, high school science teachers, state legislators, professional organizations, and many other groups.« less

Fusion Safety Program annual report, fiscal year 1994

NASA Astrophysics Data System (ADS)

Longhurst, Glen R.; Cadwallader, Lee C.; Dolan, Thomas J.; Herring, J. Stephen; McCarthy, Kathryn A.; Merrill, Brad J.; Motloch, Chester C.; Petti, David A.

1995-03-01

This report summarizes the major activities of the Fusion Safety Program in fiscal year 1994. The Idaho National Engineering Laboratory (INEL) is the designated lead laboratory and Lockheed Idaho Technologies Company is the prime contractor for this program. The Fusion Safety Program was initiated in 1979. Activities are conducted at the INEL, at other DOE laboratories, and at other institutions, including the University of Wisconsin. The technical areas covered in this report include tritium safety, beryllium safety, chemical reactions and activation product release, safety aspects of fusion magnet systems, plasma disruptions, risk assessment failure rate data base development, and thermalhydraulics code development and their application to fusion safety issues. Much of this work has been done in support of the International Thermonuclear Experimental Reactor (ITER). Also included in the report are summaries of the safety and environmental studies performed by the Fusion Safety Program for the Tokamak Physics Experiment and the Tokamak Fusion Test Reactor and of the technical support for commercial fusion facility conceptual design studies. A major activity this year has been work to develop a DOE Technical Standard for the safety of fusion test facilities.

The PROSPECT physics program

NASA Astrophysics Data System (ADS)

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

2016-11-01

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

Metabolic modeling of synthesis gas fermentation in bubble column reactors.

PubMed

Chen, Jin; Gomez, Jose A; Höffner, Kai; Barton, Paul I; Henson, Michael A

2015-01-01

A promising route to renewable liquid fuels and chemicals is the fermentation of synthesis gas (syngas) streams to synthesize desired products such as ethanol and 2,3-butanediol. While commercial development of syngas fermentation technology is underway, an unmet need is the development of integrated metabolic and transport models for industrially relevant syngas bubble column reactors. We developed and evaluated a spatiotemporal metabolic model for bubble column reactors with the syngas fermenting bacterium Clostridium ljungdahlii as the microbial catalyst. Our modeling approach involved combining a genome-scale reconstruction of C. ljungdahlii metabolism with multiphase transport equations that govern convective and dispersive processes within the spatially varying column. The reactor model was spatially discretized to yield a large set of ordinary differential equations (ODEs) in time with embedded linear programs (LPs) and solved using the MATLAB based code DFBAlab. Simulations were performed to analyze the effects of important process and cellular parameters on key measures of reactor performance including ethanol titer, ethanol-to-acetate ratio, and CO and H2 conversions. Our computational study demonstrated that mathematical modeling provides a complementary tool to experimentation for understanding, predicting, and optimizing syngas fermentation reactors. These model predictions could guide future cellular and process engineering efforts aimed at alleviating bottlenecks to biochemical production in syngas bubble column reactors.

White paper report on using nuclear reactors to search for a value of theta13

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

Anderson, K.; Anjos, J.C.; Ayres, D.

2004-02-26

There has been superb progress in understanding the neutrino sector of elementary particle physics in the past few years. It is now widely recognized that the possibility exists for a rich program of measuring CP violation and matter effects in future accelerator {nu} experiments, which has led to intense efforts to consider new programs at neutrino superbeams, off-axis detectors, neutrino factories and beta beams. However, the possibility of measuring CP violation can be fulfilled only if the value of the neutrino mixing parameter {theta}{sub 13} is such that sin{sup 2} (2{theta}{sub 13}) greater than or equal to on the ordermore » of 0.01. The authors of this white paper are an International Working Group of physicists who believe that a timely new experiment at a nuclear reactor sensitive to the neutrino mixing parameter {theta}{sub 13} in this range has a great opportunity for an exciting discovery, a non-zero value to {theta}{sub 13}. This would be a compelling next step of this program. We are studying possible new reactor experiments at a variety of sites around the world, and we have collaborated to prepare this document to advocate this idea and describe some of the issues that are involved.« less

Program for studying fundamental interactions at the PIK reactor facilities

NASA Astrophysics Data System (ADS)

Serebrov, A. P.; Vassiljev, A. V.; Varlamov, V. E.; Geltenbort, P.; Gridnev, K. A.; Dmitriev, S. P.; Dovator, N. A.; Egorov, A. I.; Ezhov, V. F.; Zherebtsov, O. M.; Zinoviev, V. G.; Ivochkin, V. G.; Ivanov, S. N.; Ivanov, S. A.; Kolomensky, E. A.; Konoplev, K. A.; Krasnoschekova, I. A.; Lasakov, M. S.; Lyamkin, V. A.; Martemyanov, V. P.; Murashkin, A. N.; Neustroev, P. V.; Onegin, M. S.; Petelin, A. L.; Pirozhkov, A. N.; Polyushkin, A. O.; Prudnikov, D. V.; Ryabov, V. L.; Samoylov, R. M.; Sbitnev, S. V.; Fomin, A. K.; Fomichev, A. V.; Zimmer, O.; Cherniy, A. V.; Shoka, I. V.

2016-05-01

A research program aimed at studying fundamental interactions by means of ultracold and polarized cold neutrons at the GEK-4-4' channel of the PIK reactor is presented. The apparatus to be used includes a source of cold neutrons in the heavy-water reflector of the reactor, a source of ultracold neutrons based on superfluid helium and installed in a cold-neutron beam extracted from the GEK-4 channel, and a number of experimental facilities in neutron beams. An experiment devoted to searches for the neutron electric dipole moment and an experiment aimed at a measurement the neutron lifetime with the aid of a large gravitational trap are planned to be performed in a beam of ultracold neutrons. An experiment devoted to measuring neutron-decay asymmetries with the aid of a superconducting solenoid is planned in a beam of cold polarized neutrons from the GEK-4' channel. The second ultracold-neutron source and an experiment aimed at measuring the neutron lifetime with the aid of a magnetic trap are planned in the neutron-guide system of the GEK-3 channel. In the realms of neutrino physics, an experiment intended for sterile-neutrino searches is designed. The state of affairs around the preparation of the experimental equipment for this program is discussed.

A preliminary assessment of the effects of heat flux distribution and penetration on the creep rupture of a reactor vessel lower head

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

Chu, T.Y.; Bentz, J.; Simpson, R.

1997-02-01

The objective of the Lower Head Failure (LHF) Experiment Program is to experimentally investigate and characterize the failure of the reactor vessel lower head due to thermal and pressure loads under severe accident conditions. The experiment is performed using 1/5-scale models of a typical PWR pressure vessel. Experiments are performed for various internal pressure and imposed heat flux distributions with and without instrumentation guide tube penetrations. The experimental program is complemented by a modest modeling program based on the application of vessel creep rupture codes developed in the TMI Vessel Investigation Project. The first three experiments under the LHF programmore » investigated the creep rupture of simulated reactor pressure vessels without penetrations. The heat flux distributions for the three experiments are uniform (LHF-1), center-peaked (LHF-2), and side-peaked (LHF-3), respectively. For all the experiments, appreciable vessel deformation was observed to initiate at vessel wall temperatures above 900K and the vessel typically failed at approximately 1000K. The size of failure was always observed to be smaller than the heated region. For experiments with non-uniform heat flux distributions, failure typically occurs in the region of peak temperature. A brief discussion of the effect of penetration is also presented.« less

Nonproliferation and Threat Reduction Assistance: U.S. Programs in the Former Soviet Union

DTIC Science & Technology

2009-07-31

seeks to help Russia reconfigure its large - scale former BW-related facilities so that they can perform peaceful research issues such as infectious...opting instead for the construction of fast breeder reactors that could burn plutonium directly for energy production. The United States might not fund...this effort, as many in the United States argue that breeder reactors , which produce more plutonium than they consume, would undermine

Dual benefit robotics programs at Sandia National Laboratories

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

Jones, A.T.

Sandia National Laboratories has one of the largest integrated robotics laboratories in the United States. Projects include research, development, and application of one-of-a-kind systems, primarily for the Department of Energy (DOE) complex. This work has been underway for more than 10 years. It began with on-site activities that required remote operation, such as reactor and nuclear waste handling. Special purpose robot systems were developed using existing commercial manipulators and fixtures and programs designed in-house. These systems were used in applications such as servicing the Sandia pulsed reactor and inspecting remote roof bolts in an underground radioactive waste disposal facility. Inmore » the beginning, robotics was a small effort, but with increasing attention to the use of robots for hazardous operations, efforts now involve a staff of more than 100 people working in a broad robotics research, development, and applications program that has access to more than 30 robotics systems.« less

A Monte Carlo program to calculate the exposure rate from airborne radioactive gases inside a nuclear reactor containment building.

PubMed

Sherbini, S; Tamasanis, D; Sykes, J; Porter, S W

1986-12-01

A program was developed to calculate the exposure rate resulting from airborne gases inside a reactor containment building. The calculations were performed at the location of a wall-mounted area radiation monitor. The program uses Monte Carlo techniques and accounts for both the direct and scattered components of the radiation field at the detector. The scattered component was found to contribute about 30% of the total exposure rate at 50 keV and dropped to about 7% at 2000 keV. The results of the calculations were normalized to unit activity per unit volume of air in the containment. This allows the exposure rate readings of the area monitor to be used to estimate the airborne activity in containment in the early phases of an accident. Such estimates, coupled with containment leak rates, provide a method to obtain a release rate for use in offsite dose projection calculations.

DMM: A MULTIGROUP, MULTIREGION ONE-SPACE-DIMENSIONAL COMPUTER PROGRAM USING NEUTRON DIFFUSION THEORY. PART II. DMM PROGRAM DESCRIPTION

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

Kavanagh, D.L.; Antchagno, M.J.; Egawa, E.K.

1960-12-31

Operating instructions are presented for DMM, a Remington Rand 1103A program using one-space-dimensional multigroup diffusion theory to calculate the reactivity or critical conditions and flux distribution of a multiregion reactor. Complete descriptions of the routines and problem input and output specifications are also included. (D.L.C.)

Small Reactor Designs Suitable for Direct Nuclear Thermal Propulsion: Interim Report

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

Bruce G. Schnitzler

Advancement of U.S. scientific, security, and economic interests requires high performance propulsion systems to support missions beyond low Earth orbit. A robust space exploration program will include robotic outer planet and crewed missions to a variety of destinations including the moon, near Earth objects, and eventually Mars. Past studies, in particular those in support of both the Strategic Defense Initiative (SDI) and the Space Exploration Initiative (SEI), have shown nuclear thermal propulsion systems provide superior performance for high mass high propulsive delta-V missions. In NASA's recent Mars Design Reference Architecture (DRA) 5.0 study, nuclear thermal propulsion (NTP) was again selectedmore » over chemical propulsion as the preferred in-space transportation system option for the human exploration of Mars because of its high thrust and high specific impulse ({approx}900 s) capability, increased tolerance to payload mass growth and architecture changes, and lower total initial mass in low Earth orbit. The recently announced national space policy2 supports the development and use of space nuclear power systems where such systems safely enable or significantly enhance space exploration or operational capabilities. An extensive nuclear thermal rocket technology development effort was conducted under the Rover/NERVA, GE-710 and ANL nuclear rocket programs (1955-1973). Both graphite and refractory metal alloy fuel types were pursued. The primary and significantly larger Rover/NERVA program focused on graphite type fuels. Research, development, and testing of high temperature graphite fuels was conducted. Reactors and engines employing these fuels were designed, built, and ground tested. The GE-710 and ANL programs focused on an alternative ceramic-metallic 'cermet' fuel type consisting of UO2 (or UN) fuel embedded in a refractory metal matrix such as tungsten. The General Electric program examined closed loop concepts for space or terrestrial applications as well as open loop systems for direct nuclear thermal propulsion. Although a number of fast spectrum reactor and engine designs suitable for direct nuclear thermal propulsion were proposed and designed, none were built. This report summarizes status results of evaluations of small nuclear reactor designs suitable for direct nuclear thermal propulsion.« less

A Unique Master's Program in Combined Nuclear Technology and Nuclear Chemistry at Chalmers University of Technology, Sweden

NASA Astrophysics Data System (ADS)

Skarnemark, Gunnar; Allard, Stefan; Ekberg, Christian; Nordlund, Anders

2009-08-01

The need for engineers and scientists who can ensure safe and secure use of nuclear energy is large in Sweden and internationally. Chalmers University of Technology is therefore launching a new 2-year master's program in Nuclear Engineering, with start from the autumn of 2009. The program is open to Swedish and foreign students. The program starts with compulsory courses dealing with the basics of nuclear chemistry and physics, radiation protection, nuclear power and reactors, nuclear fuel supply, nuclear waste management and nuclear safety and security. There are also compulsory courses in nuclear industry applications and sustainable energy futures. The subsequent elective courses can be chosen freely but there is also a possibility to choose informal tracks that concentrate on nuclear chemistry or reactor technology and physics. The nuclear chemistry track comprises courses in e.g. chemistry of lanthanides, actinides and transactinides, solvent extraction, radioecology and radioanalytical chemistry and radiopharmaceuticals. The program is finished with a one semester thesis project. This is probably a unique master program in the sense of its combination of deep courses in both nuclear technology and nuclear chemistry.

Opportunities for the LWR ATF materials development program to contribute to the LBE-cooled ADS materials qualification program

NASA Astrophysics Data System (ADS)

Gong, Xing; Li, Rui; Sun, Maozhou; Ren, Qisen; Liu, Tong; Short, Michael P.

2016-12-01

Accelerator-driven systems (ADS) are a promising approach for nuclear waste disposal. Nevertheless, the principal candidate materials proposed for ADS construction, such as the ferritic/martensitic steel, T91, and austenitic stainless steels, 316L and 15-15Ti, are not fully compatible with the liquid lead-bismuth eutectic (LBE) coolant. Under some operating conditions, liquid metal embrittlement (LME) or liquid metal corrosion (LMC) may occur in these steels when exposed to LBE. These environmentally-induced material degradation effects pose a threat to ADS reactor safety, as failure of the materials could initiate a severe accident, in which fission products are released into the coolant. Meanwhile, parallel efforts to develop accident-tolerant fuels (ATF) in light water reactors (LWRs) could provide both general materials design philosophies and specific material solutions to the ADS program. In this paper, the potential contributions of the ATF materials development program to the ADS materials qualification program are evaluated and discussed in terms of service conditions and materials performance requirements. Several specific areas where coordinated development may benefit both programs, including composite materials and selected coatings, are discussed.

AGC 2 Irradiated Material Properties Analysis

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

Rohrbaugh, David Thomas

2017-05-01

The Advanced Reactor Technologies Graphite Research and Development Program is conducting an extensive graphite irradiation experiment to provide data for licensing of a high temperature reactor (HTR) design. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor designs. , Nuclear graphite H 451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphite grades have been developed and are considered suitable candidates for new HTR reactor designs. To support the design and licensing of HTR core componentsmore » within a commercial reactor, a complete properties database must be developed for these current grades of graphite. Quantitative data on in service material performance are required for the physical, mechanical, and thermal properties of each graphite grade, with a specific emphasis on data accounting for the life limiting effects of irradiation creep on key physical properties of the HTR candidate graphite grades. Further details on the research and development activities and associated rationale required to qualify nuclear grade graphite for use within the HTR are documented in the graphite technology research and development plan.« less

AGC 2 Irradiation Creep Strain Data Analysis

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

Windes, William E.; Rohrbaugh, David T.; Swank, W. David

2016-08-01

The Advanced Reactor Technologies Graphite Research and Development Program is conducting an extensive graphite irradiation experiment to provide data for licensing of a high temperature reactor (HTR) design. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor designs. Nuclear graphite H-451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphite grades have been developed and are considered suitable candidates for new HTR reactor designs. To support the design and licensing of HTR core components within amore » commercial reactor, a complete properties database must be developed for these current grades of graphite. Quantitative data on in service material performance are required for the physical, mechanical, and thermal properties of each graphite grade, with a specific emphasis on data accounting for the life limiting effects of irradiation creep on key physical properties of the HTR candidate graphite grades. Further details on the research and development activities and associated rationale required to qualify nuclear grade graphite for use within the HTR are documented in the graphite technology research and development plan.« less

Design of conduction cooling system for a high current HTS DC reactor

NASA Astrophysics Data System (ADS)

Dao, Van Quan; Kim, Taekue; Le Tat, Thang; Sung, Haejin; Choi, Jongho; Kim, Kwangmin; Hwang, Chul-Sang; Park, Minwon; Yu, In-Keun

2017-07-01

A DC reactor using a high temperature superconducting (HTS) magnet reduces the reactor’s size, weight, flux leakage, and electrical losses. An HTS magnet needs cryogenic cooling to achieve and maintain its superconducting state. There are two methods for doing this: one is pool boiling and the other is conduction cooling. The conduction cooling method is more effective than the pool boiling method in terms of smaller size and lighter weight. This paper discusses a design of conduction cooling system for a high current, high temperature superconducting DC reactor. Dimensions of the conduction cooling system parts including HTS magnets, bobbin structures, current leads, support bars, and thermal exchangers were calculated and drawn using a 3D CAD program. A finite element method model was built for determining the optimal design parameters and analyzing the thermo-mechanical characteristics. The operating current and inductance of the reactor magnet were 1,500 A, 400 mH, respectively. The thermal load of the HTS DC reactor was analyzed for determining the cooling capacity of the cryo-cooler. The study results can be effectively utilized for the design and fabrication of a commercial HTS DC reactor.

Evaluation of the use of nodal methods for MTR neutronic analysis

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

Reitsma, F.; Mueller, E.Z.

1997-08-01

Although modern nodal methods are used extensively in the nuclear power industry, their use for research reactor analysis has been very limited. The suitability of nodal methods for material testing reactor analysis is investigated with the emphasis on the modelling of the core region (fuel assemblies). The nodal approach`s performance is compared with that of the traditional finite-difference fine mesh approach. The advantages of using nodal methods coupled with integrated cross section generation systems are highlighted, especially with respect to data preparation, simplicity of use and the possibility of performing a great variety of reactor calculations subject to strict timemore » limitations such as are required for the RERTR program.« less

Post impact behavior of mobile reactor core containment systems

NASA Technical Reports Server (NTRS)

Puthoff, R. L.; Parker, W. G.; Vanbibber, L. E.

1972-01-01

The reactor core containment vessel temperatures after impact, and the design variables that affect the post impact survival of the system are analyzed. The heat transfer analysis includes conduction, radiation, and convection in addition to the core material heats of fusion and vaporization under partially burial conditions. Also, included is the fact that fission products vaporize and transport radially outward and condense outward and condense on cooler surfaces, resulting in a moving heat source. A computer program entitled Executive Subroutines for Afterheat Temperature Analysis (ESATA) was written to consider this complex heat transfer analysis. Seven cases were calculated of a reactor power system capable of delivering up to 300 MW of thermal power to a nuclear airplane.

ORNL-TNS/PEPR overall heating requirements

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

Peng, Y. K.M.; Rome, J. A.

1977-01-01

The ORNL TNS/PEPR studies have the objectives of (1) leading to a system that demonstrates the fusion reactor core in the mid-to-late 1980's and extrapolates to an economic tokamak power reactor, and (2) providing a near-term focus for the scientific and technological programs toward the power reactor. This discussion of the overall heating requirements for the ORNL TNS/PEPR is concerned with the neutral beams as the primary heating method, the electron-cyclotron resonance (ECR) heating at a lower power level for profile control, and the upper hybrid resonance (UHR) initiation and preheating of currentless plasmas to reduce current start-up loop voltagemore » (V/sub l/) requirements.« less

Assessment of Sensor Technologies for Advanced Reactors

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

Korsah, Kofi; Ramuhalli, Pradeep; Vlim, R.

2016-10-01

Sensors and measurement technologies provide information on processes, support operations and provide indications of component health. They are therefore crucial to plant operations and to commercialization of advanced reactors (AdvRx). This report, developed by a three-laboratory team consisting of Argonne National Laboratory (ANL), Oak Ridge National Laboratory (ORNL) and Pacific Northwest National Laboratory (PNNL), provides an assessment of sensor technologies and a determination of measurement needs for AdvRx. It provides the technical basis for identifying and prioritizing research targets within the instrumentation and control (I&C) Technology Area under the Department of Energy’s (DOE’s) Advanced Reactor Technology (ART) program and contributesmore » to the design and implementation of AdvRx concepts.« less

Energy Security Threats

DTIC Science & Technology

2011-06-01

continued  although  there  were  equally  ambitious  programs  to  develop nuclear powered submarines at the same time. The first  reactor  capable  of...new plants.13 The  construction  of  new  reactors  has continued periodically since  that  time such  that at  the end of  2009  total  global  nuclear...electrical  generation  capacity  totalled  370  giga watts  (electric)  (GW(e))  spread  amongst  437  operational  reactors .14  According  to

Proceedings of the 1988 International Meeting on Reduced Enrichment for Research and Test Reactors

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

Not Available

1993-07-01

The international effort to develop and implement new research reactor fuels utilizing low-enriched uranium, instead of highly- enriched uranium, continues to make solid progress. This effort is the cornerstone of a widely shared policy aimed at reducing, and possibly eliminating, international traffic in highly-enriched uranium and the nuclear weapon proliferation concerns associated with this traffic. To foster direct communication and exchange of ideas among the specialists in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the eleventh of a series which began 1978. Individual papers presented at the meetingmore » have been cataloged separately.« less

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

NASA Technical Reports Server (NTRS)

1972-01-01

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

National Environmental Policy Act Hazards Assessment for the TREAT Alternative

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

Boyd D. Christensen; Annette L. Schafer

2013-11-01

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

National Environmental Policy Act Hazards Assessment for the TREAT Alternative

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

Christensen, Boyd D.; Schafer, Annette L.

2014-02-01

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

RELAP-7 Software Verification and Validation Plan

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

Smith, Curtis L.; Choi, Yong-Joon; Zou, Ling

This INL plan comprehensively describes the software for RELAP-7 and documents the software, interface, and software design requirements for the application. The plan also describes the testing-based software verification and validation (SV&V) process—a set of specially designed software models used to test RELAP-7. The RELAP-7 (Reactor Excursion and Leak Analysis Program) code is a nuclear reactor system safety analysis code being developed at Idaho National Laboratory (INL). The code is based on the INL’s modern scientific software development framework – MOOSE (Multi-Physics Object-Oriented Simulation Environment). The overall design goal of RELAP-7 is to take advantage of the previous thirty yearsmore » of advancements in computer architecture, software design, numerical integration methods, and physical models. The end result will be a reactor systems analysis capability that retains and improves upon RELAP5’s capability and extends the analysis capability for all reactor system simulation scenarios.« less