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Status of the design concepts for a high fluence fast pulse reactor (HFFPR)

Description: The report describes progress that has been made on the design of a High Fluence Fast Pulse Reactor (HFFPR) through the end of calendar year 1977. The purpose of this study is to present design concepts for a test reactor capable of accommodating large scale reactor safety tests. These concepts for reactor safety tests are adaptations of reactor concepts developed earlier for DOE/OMA for the conduct of weapon effects tests. The preferred driver core uses fuel similar to that developed for Sandia's ACPR upgrade. It is a BeO/UO/sub 2/ fuel that is gas cooled and has a high volumetric heat capacity. The present version of the design can drive large (217) pin bundles of prototypically enriched mixed oxide fuel well beyond the fuel's boiling point. Applicability to specific reactor safety accident scenarios and subsequent design improvements will be presented in future reports on this subject.
Date: October 1, 1978
Creator: Philbin, J.S.; Nelson, W.E. & Rosenstroch, B.
Partner: UNT Libraries Government Documents Department

Two-Dimensional Cell Calculations for Critical Assembly Analysis

Description: A pseudo two-dimensional collision probability method for calculating plate and pin cell flux distributions has been developed, tested against Monte Carlo calculations, and has been found to yield collision probabilities which are in error less than 2%. In both the plate and pin cell cases, the pseudo 2-D collision probability methods have been shown to be as accurate as direct collision probability methods for the geometries of interest in ZPPR critical assembly analysis. The computational efficiency of the pseudo 2-D method is approximately the same as standard 1-D methods.
Date: 1984~
Creator: Smith, K. S.
Partner: UNT Libraries Government Documents Department

Inspection methods for physical protection Task III review of other agencies' physical security activities for research reactors

Description: In Task I of this project, the current Nuclear Regulatory Commission (NRC) position-on physical security practices and procedures at research reactors were reviewed. In the second task, a sampling of the physical security plans was presented and the three actual reactor sites described in the security plans were visited. The purpose of Task III is to review other agencies' physical security activities for research reactors. During this phase, the actions, procedures and policies of two domestic and two foreign agencies other than the NRC that relate to the research reactor community were examined. The agencies examined were: International Atomic Energy Agency; Canadian Atomic Energy Control Board; Department of Energy; and American Nuclear Insurers.
Date: unknown
Partner: UNT Libraries Government Documents Department

Treatment of measurement uncertainties at the power burst facility

Description: The treatment of measurement uncertainty at the Power Burst Facility provides a means of improving data integrity as well as meeting standard practice reporting requirements. This is accomplished by performing the uncertainty analysis in two parts, test independent uncertainty analysis and test dependent uncertainty analysis. The test independent uncertainty analysis is performed on instrumentation used repeatedly from one test to the next, and does not have to be repeated for each test except for improved or new types of instruments. A test dependent uncertainty analysis is performed on each test based on the test independent uncertainties modified as required by test specifications, experiment fixture design, and historical performance of instruments on similar tests. The methodology for performing uncertainty analysis based on the National Bureau of Standards method is reviewed with examples applied to nuclear instrumentation.
Date: January 1, 1980
Creator: Meyer, L.C.
Partner: UNT Libraries Government Documents Department

Fabrication of internally instrumented reactor fuel rods

Description: Procedures are outlined for fabricating internally instrumented reactor fuel rods while maintaining the original quality assurance level of the rods. Instrumented fuel rods described contain fuel centerline thermocouples, ultrasonic thermometers, and pressure tubes for internal rod gas pressure measurements. Descriptions of the thermocouples and ultrasonic thermometers are also contained. (auth)
Date: January 1, 1975
Creator: Schmutz, J.D. & Meservey, R.H.
Partner: UNT Libraries Government Documents Department

Fast neutron spectrum and dosimetry studies in the Coupled Fast Reactivity Measurements Facility

Description: The fast neutron spectrum of the Coupled Fast Reactivity Measurements Facility (CFRMF) at the Idaho National Engineering Laboratory (INEL) is being used to study and standardize fast reactor neutron dosimetry materials and methods. The CFRMF has been designated a ''benchmark experiment'' to test the cross section data of dosimetry materials as well as other materials used and produced in fast reactors. Information about the neutron energy spectrum of the CFRMF is presented. (auth)
Date: January 1, 1975
Creator: Rogers, J.W.; Harker, Y.D. & Millsap, D.A.
Partner: UNT Libraries Government Documents Department

Designing CNR, a very high thermal neutron flux facility

Description: According to a recent study (Eastman-Seitz Committee, National Academy of Science) there is a need for a new generation of steady neutron sources with a thermal neutron flux peak between 5 to 10 times 10/sup 15//cm/sup 2/ sec. Ideally the neutron source would have to operate continuously for several days (two weeks at least) with minimum time (2 to 3 days) for refueling and/or maintenance and it would also be used to irradiate materials and produce isotopes. This paper describes the preliminary design of the nuclear reactor for the proposed Center for Neutron Research (CNR). A duplication of existing designs (HFIR, (ORNL), ILL (Grenoble, France)) would imply high total power and small core life; the necessity of higher efficiencies (in terms of peak-flux-per-unit source or power) then becomes apparent. We have found analytical expressions for the efficiency in terms of a few parameters such as the volume of the source and the Fermi age and diffusion length of thermal neutrons in both the source and reflector regions. A single analytical expression can then be used for scoping the design and to intercompare radically different designs. Higher efficiencies can be achieved by reducing the volume and the moderation of a core immersed in a very low absorbing reflector; on the contrary a very long core life has a negative effect on the efficiency at beginning of life. Consequently, and after detailed calculations, we have found a candidate design with the following characteristics: core, U/sub 3/Si/sub 2/, 93% enriched, 18.1-kg /sup 235/U, metal fraction 50%, Al cladding, and 35-L volume; reflector and moderator, D/sub 2/O; efficiency at end of life (EOL) with respect to the ILL reactor, 1.29; flux at EOL, 10 x 10/sup 15//cm/sup 2/ sec (power in core 270. MW); core life, 14 days; burnup 28.4%.
Date: January 1, 1986
Creator: Difilippo, F.C.
Partner: UNT Libraries Government Documents Department

Critical experiments in support of the CNPS (Compact Nuclear Power Source) program

Description: Zero-power static and kinetic measurements have been made on a mock-up of the Compact Nuclear Power Source (CNPS), a graphite moderated, graphite reflected, U(19.9% /sup 235/U) fueled reactor design. Critical configurations were tracked from a first clean configuration (184 most central fuel channels filled and all control rod and heat pipe channels empty) to a fully loaded configuration (all 492 fuel channels filled, core-length stainless steel pipe in the twelve heat-pipe channels, and approximately half-core-length boron carbide in the outer 4 control rod channels. Reactor physics data such as material worths and neutron lifetime are presented only for the clean and fully loaded configurations.
Date: January 1, 1988
Creator: Hansen, G.E.; Audas, J.H.; Martin, E.R.; Pederson, R.A.; Spriggs, G.D. & White, R.H.
Partner: UNT Libraries Government Documents Department

Reckoning THOR

Description: Theoretical computation of the Los Alamos National Laboratory's critical assembly THOR (a thorium-reflected plutonium sphere) yields a high eigenvalue when compared to the experimentally measured eigenvalue. Several calculational improvements are investigated in an effort to reduce the discrepancy. Finally, the experimental procedure of reducing the raw configuration to clean specifications is reviewed.
Date: May 1, 1981
Creator: Kidman, R.B.
Partner: UNT Libraries Government Documents Department

Neutron radiographic facility at the 3-mw Livermore pool-type reactor

Description: A description is presented of the neutron experimental radiographic facility at the Livermore Pool-Type Reactor. This facility was installed in 1974 to assist Lawrence Livermore Laboratory research programs. Some of the testing techniques used to modify the neutron beam and the present radiographic parameters are also discussed. (auth)
Date: September 10, 1975
Creator: Richards, W.J.; Peterson, R.T. & Prindle, J.A.
Partner: UNT Libraries Government Documents Department

Data acquisition and control system for the K/sub 1C/-HSST experiments at the ORR

Description: Major components and primary functions of the process control system for the K/sub 1C/-HSST irradiation experiments at the Oak Ridge Research (ORR) are described. Information relative to methodology for integrating unique features of the Digital Equipment Corporation's RSX-11M Operating System with analog-to-digital and digital-to-analog hardware is presented. In particular, data flow among various real-time applications programs relative to system hardware is presented. General features of the temperature control algorithm are presented, and results that illustrate the spatial temperature distribution in the capsule achieved by the control system are included.
Date: January 1, 1984
Creator: Miller, L.F. & Hobbs, R.W.
Partner: UNT Libraries Government Documents Department

Status report of irradiated NPR fuel element rupture studies in the IRP

Description: The Irradiated Rupture Prototype (IRP) has been used for rupture testing irradiated NPR prototype fuel elements. Most of the tests have been made to determine the rupture effect of different reactor exposures, fuel element geometries and water cooldown rates following the start of the rupture. This report summarizes the results obtained to date, mentions where information is lacking and gives further tests scheduled for the IRP.
Date: September 9, 1963
Creator: Hayden, K.D.
Partner: UNT Libraries Government Documents Department

FAST FLUX TEST FACILITY. Design and development quality assurance requirements for the FFTF

Description: The document is presented to provide general management requirements for Pacific Northwest Laboratory (PNL) and contractor design and development quality assurance programs to assure the required quality level of the various items required for the FFTF. The document is applicable as imposed by the contract to FFTF contractors and subcontractors. The document is also applicable to PNL design and development activities related to the FFTF.
Date: October 23, 1968
Creator: Albert, W.G.
Partner: UNT Libraries Government Documents Department

In-place testing of off-gas iodine filters

Description: At the Idaho National Engineering Laboratory, both charcoal and silver zeolite (AgX) filters are used for radioactive iodine off-gas cleanup of reactor systems. These filters are used in facilities which are conducting research in the areas of reactor fuel failure, reactor fuel inspection, and loss of fluids from reactor vessels. Iodine retention efficiency testing of these filters is dictated by prudent safety practices and regulatory guidelines. A procedure for determining iodine off-gas filter efficiency in-place has been developed and tested on both AgX and charcoal filters. The procedure involves establishing sample points upstream and downstream of the filter to be tested. A step-by-step approach for filter efficiency testing is presented.
Date: January 1, 1980
Creator: Duce, S.W.; Tkachyk, J.W. & Motes, B.G.
Partner: UNT Libraries Government Documents Department

Oak Ridge Research Reactor quarterly report, July, August, and September of 1977

Description: The ORR operated at an average power level of 29.5 MW for 23.4% of the time during July, August, and September of 1977. Three fuel elements were declared spent (60.7% burnup) during the quarter, while five new elements were placed in service. The reactor was shut down on seven occasions, none of which were unscheduled. Reactor downtime needed for refueling, maintenance and checks was quite low, with the reactor remaining available for operation 86.5% of the time. Maintenance activities, both mechanical and instrument were essentially routine in nature. In-service inspections completed during the quarter included inspection of the ORR primary heat exchanger No. 2.
Date: January 1, 1978
Creator: Hurt, S.S. III & Lance, E.D.
Partner: UNT Libraries Government Documents Department

Operation of a nuclear test gage at low multiplications

Description: The Nuclear Test Gage (NTG) at the Savannah River Plant is a subcritical multiplying facility (low k) with H/sub 2/O moderator and 2.54-cm-diameter fuel slugs of 5 wt percent /sup 235/U in aluminum alloy at a 4.285-cm triangular pitch. The core of the facility is 61-cm long with a normal diameter of 27 cm. The NTG is used for quality control of reactor components, such as /sup 235/U-Al fuel tubes, Li--Al target tubes, control and safety rods, and miscellaneous special irradiation elements. A component is tested by passing it through an axial test port 11.63 cm in diameter. The ion chamber response from the resultant change in neutron source multiplication is then compared with corresponding responses from known standards.
Date: January 1, 1977
Creator: Baumann, N.P.
Partner: UNT Libraries Government Documents Department

Los Alamos Omega West Reactor

Description: A description is given of the Omega West Reactor and associated experimental facilities, followed by a brief discussion of recent usage, new experiments, and future prospects.
Date: January 1, 1983
Creator: Lyle, A.R.; Williams, H.T. & Bunker, M.E.
Partner: UNT Libraries Government Documents Department

Component design for LMFBR's

Description: Just as FFTF has prototype components to confirm their design, FFTF is serving as a prototype for the design of the commercial LMFBR's. Design and manufacture of critical components for the FFTF system have been accomplished primarily using vendors with little or no previous experience in supplying components for high temperature sodium systems. The exposure of these suppliers, and through them a multitude of subcontractors, to the requirements of this program has been a necessary and significant step in preparing American industry for the task of supplying the large mechanical components required for commercial LMFBR's. (auth)
Date: January 1, 1975
Creator: Fillnow, R.H.; France, L.L.; Zerinvary, M.C. & Fox, R.O.
Partner: UNT Libraries Government Documents Department

Oak Ridge research reactor. Quarterly report, April-June 1979

Description: The ORR operated at an average power level of 29.9 MW for 77.9% of the time during April, May, and June of 1979. The reactor was shut down on seven occasions, two of which were unscheduled. Reactor downtime needed for refueling, maintenance and checks was normal, with the reactor remaining available for operation 86.5% of the time. Maintenance activities, both mechanical and instrument, were essentially routine in nature.
Date: December 1, 1979
Creator: Hurt, S.S. III & Lance, E.D.
Partner: UNT Libraries Government Documents Department

Oak Ridge Research Reactor quarterly report October, November and December of 1978

Description: The ORR operated at an average power level of 29.0 MW for 80.9% of the time during October, November, and December of 1978. The reactor was shut down on fourteen occasions, four of which were unscheduled. Reactor downtime needed for refueling, maintenance and checks was normal, with the reactor remaining available for operation 92.2% of the time. Maintenance activities, both mechanical and instrument, were essentially routine in nature.
Date: May 1, 1979
Creator: Hurt, S.S. III & Lance, E.D.
Partner: UNT Libraries Government Documents Department

Oak Ridge Research Reactor quarterly report, January-March, 1979

Description: The ORR operated at an average power level of 28.9 MW for 84.2% of the time during January, February and March of 1979. The reactor was shut down on seventeen occasions, three of which were unscheduled. Reactor downtime needed for refueling, maintenance and checks was normal, with the reactor remaining available for operation 93.6% of the time. Maintenance activities, both mechanical and instrument, were essentially routine in nature with the exception of one mechanical design change. Special tests or measurements completed during the quarter included: (1) preliminary measurements for the LWR Pressure Vessel Irradiation Program; (2) flux mapping for Core No. 148-A; and (3) calibration of all six shim rods.
Date: October 1, 1979
Creator: Hurt, S.S. III & Lance, E.D.
Partner: UNT Libraries Government Documents Department

Fuel plate stability experiments and analysis for the Advanced Neutron Source

Description: The planned reactor for the Advanced Neutron Source (ANS) will use closely spaced arrays of involute-shaped fuel plates that will be cooled by water flowing through the channels between the plates. There is concern that at certain coolant flow velocities, adjacent plates may deflect and touch, with resulting failure of the plates. Experiments have been conducted at the Oak Ridge National Laboratory to examine this potential phenomenon. Results of the experiments and comparison with analytical predictions are reported. The tests were conducted using full-scale epoxy plate models of the aluminum/uranium silicide ANS involute-shaped fuel plates. Use of epoxy plates and model theory allowed lower flow velocities and pressures to explore the potential failure mechanism. Plate deflections and channel pressures as functions of the flow velocity are examined. Comparisons with mathematical models are noted.
Date: May 1, 1993
Creator: Swinson, W.F.; Battiste, R.L.; Luttrell, C.R. & Yahr, G.T.
Partner: UNT Libraries Government Documents Department