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Model for evaluating nuclear strategies with proliferation resistance

Description: A model was developed at HEDL to specifically analyze proliferation resistant strategies. The model was not designed to predict the future, but rather to provide a method for estimating the consequences of decisions affecting proliferation resistance in a rational and plausible manner. The characteristics of the model are described.
Date: March 1, 1979
Creator: Shay, M. R.; Hardie, R. W. & Omberg, R. P.
Partner: UNT Libraries Government Documents Department

2200/sup 0/C fuel centerline thermocouples for the LOFT program

Description: The technology as well as commercial suppliers have been developed for high temperature thermocouples for the Loss-of-Fluid-Test (LOFT) program. Two types of thermocouples were developed and tested. Model B units contained a 1/16-inch OD 24-inch long Mo/Re sheath probe and were capable of temperature measurement to 1550/sup 0/C. Model A units contained a 1/16-inch OD 41-inch long W/Re-augmented sheath probe and were capable of temperature measurement to 2200/sup 0/C.
Date: July 1, 1979
Creator: Cannon, C.P. & Lunghofer, J.
Partner: UNT Libraries Government Documents Department

Acoustic emission weld monitoring of nuclear components

Description: Acoustic emission monitoring augments other nondestructive testing methods and is sometimes applicable when other tests cannot be applied. This is, in part, due to the high sensitivity of acoustic emission monitoring. Acoustic emission monitoring is only sensitive to active flaw-growth, however, and will not detect a flaw in equilibrium. This paper describes the application of acoustic emission monitoring to nuclear reactor fuel pin end closure welds and other weldments of the reactor piping.
Date: January 25, 1972
Creator: Romrell, D.M.
Partner: UNT Libraries Government Documents Department

Advances in SSTR techniques for dosimetry and radiation damage measurements

Description: Solid state track recorders (SSTR) have been applied in the diverse nuclear reactor research. Two recent advances are described which possess outstanding relevance for reactor research, namely the evolvement of SSTR radiation damage monitors and the development of CR-39, a new plastic SSTR of extremely high sensitivity. Results from high fluence irradiations of natural quartz crystal SSTR are used to illustrate the concept of the SSTR radiation damage monitor. Response characteristics of CR-39 are presented with emphasis on the remarkable proton sensitivity of this new SSTR.
Date: January 1, 1979
Creator: Gold, R.; Roberts, J. H. & Ruddy, F. H.
Partner: UNT Libraries Government Documents Department

Aerosol measurement techniques and accuracy in the CSTF. [LMFBR]

Description: The Containment Systems Test Facility (CSTF) provides the capability of performing large-scale aerosol behavior experiments at a scale factor of approximately 0.5 in height for a typical reactor containment building. The containment height is 20.3 m, the volume is 850 m/sup 3/, the design pressure is 5 bar, and quantities of sodium up to 1250 kg can be sprayed or spilled for sodium combustion product aerosol sources. Instrumentation is provided for characterization of the aerosol and the containment atmosphere. This paper describes the aerosol sampling techniques and instruments used in the CSTF and discusses their accuracy and reproducibility.
Date: November 1, 1979
Creator: McCormack, J.D. & Hilliard, R. K.
Partner: UNT Libraries Government Documents Department

Analysis of internal fuel motion during PINEX-2 experiment

Description: This paper describes the analyses performed for the PINEX-2 experiment to calculate the ejection of molten fuel into the reflector and fission gas plenum for an internally-vented fuel pin during a simulated 5$/s transient overpower excursion. The LAFM code was used to predict the transient fuel melting and fission gas release, and the HOTPIM and FUMO-T codes were used to predict the fuel ejection. The analytical results were compared with initial data from both the Pinhole-TV Imaging System and the fast-neutron hodoscope, as well as post-transient examinations of the fuel pin.
Date: September 1, 1978
Creator: Padilla, A Jr; Baars, R E; Porten, D R & Randklev, E H
Partner: UNT Libraries Government Documents Department

Analytical study of the dilation of fast reactor fuel assembly ducts

Description: An analytic method is presented for determining the dilation of fast reactor fuel assembly ducts. For temperatures where creep is linearly dependent on stress, the method is rigorous in satisfying equilibrium, compatibility and stress-strain equations. Solutions are presented for two cases: (1) a duct with constant pressure differential, (2) a duct with varying pressure differential. Results are in close agreement with finite element results of the MARC-CDC program. The method is used to predict the dilation of the Fast Test Reactor (FTR) ducts under different operating conditions. Presented are the stress, strain and dilation predictions along the duct wall, and the duct dilation variations with its geometric and loading parameters.
Date: November 15, 1978
Creator: Chan, D. P. & Jackson, R. J.
Partner: UNT Libraries Government Documents Department

Application of optimal estimation techniques to FFTF decay heat removal analysis

Description: The verification and adjustment of plant models for decay heat removal analysis using a mix of engineering judgment and formal techniques from control theory are discussed. The formal techniques facilitate dealing with typical test data which are noisy, redundant and do not measure all of the plant model state variables directly. Two pretest examples are presented.
Date: July 20, 1979
Creator: Nutt, W. T.; Additon, S. L. & Parziale, E. A.
Partner: UNT Libraries Government Documents Department

Applications of solid state track recorders in United States Nuclear Reactor Energy Programs

Description: The domain of Solid State Track Recorder (SSTR) applications in United States nuclear reactor energy programs extends from the harsh high temperature environments found in high power reactor cores to very low flux environments arising in out-of-core locations, critical assemblies, or away from reactors (AFR) experiments. The neutron energy region arising in these applications is very broad, covering more than eight decades from thermal up to fusion energies. The range of neutron flux/fluence intensity is even greater, extending over more than thirteen decades. As a consequence, use of a variety of SSTR is entailed in US Fast Breeder Reactor (FBR), Light Water Reactor (LWR), and Magnetic Fusion Energy Reactor (MFER) programs. A summary status is presented of selected SSTR experiments undertaken in these programs at the Hanford Engineering Development Laboratory (HEDL).
Date: June 1, 1979
Creator: Gold, R.; Ruddy, F. H. & Roberts, J. H.
Partner: UNT Libraries Government Documents Department

ASTM standard recommended guide on application of ENDF/A cross section and uncertainty file: establishment of the file

Description: A new ASTM Standard Recommended Guide on Application of ENDF/A Cross Section and Uncertainty File is in preparation by ASTM Committee E10 on Nuclear Technology and Applications. This ASTM Standard is being prepared in support of the standardization of physics-dosimetry procedures and data needed for Light Water Reactor (LWR) power plant pressure vessel and support structure materials surveillance and test reactor development programs. The main subject of this paper is the estabilishment of the ENDF/A Cross Section and Uncertainty File. The development of evaluated cross section files such as the evaluated nuclear data file, ENDF/B, has occurred mainly to meet the needs of physics calculators. These files are tested by calculations of well-measured benchmark problems such as reactivity or critical mass measurements. Data in the files have then been re-evaluated where disagreements with the benchmark measurements indicate data to be deficient.
Date: October 1, 1981
Creator: Lippincott, E.P. & McElroy, W.N.
Partner: UNT Libraries Government Documents Department

Computer code for the atomistic simulation of lattice defects and dynamics. [COMENT code]

Description: This document has been prepared to satisfy the need for a detailed, up-to-date description of a computer code that can be used to simulate phenomena on an atomistic level. COMENT was written in FORTRAN IV and COMPASS (CDC assembly language) to solve the classical equations of motion for a large number of atoms interacting according to a given force law, and to perform the desired ancillary analysis of the resulting data. COMENT is a dual-purpose intended to describe static defect configurations as well as the detailed motion of atoms in a crystal lattice. It can be used to simulate the effect of temperature, impurities, and pre-existing defects on radiation-induced defect production mechanisms, defect migration, and defect stability.
Date: April 1, 1980
Creator: Schiffgens, J.O.; Graves, N.J. & Oster, C.A.
Partner: UNT Libraries Government Documents Department

Computer simulation of natural circulation in FFTF secondary loops

Description: A thermal/hydraulic model of the FFTF secondary heat transport loop has been calibrated against transient natural circulation test data collected March to May 1979. The tests verified that the transition to natural convective flow could be effected from near isothermal conditions without excessive cooling at the air dump heat exchangers. Key empirical parameters of pressure drop and heat loss were found to be at 88% and 81% of pretest estimates, respectively. Pretest piping thermal transport and flow calculational models required no further revision to produce good agreement with test data.
Date: July 1, 1979
Creator: Beaver, T.R.; Turner, D.M. & Additon, S.L.
Partner: UNT Libraries Government Documents Department

Conceptual design of the high-flux VTA-2 test assembly for FMIT

Description: This report describes the conceptual design of the test module for the high neutron flux vertical test assembly (VTA-2). The description emphasizes the thermal control systems available for monitoring test specimen temperatures at any desired temperature within the range of 100 to 650/sup 0/C. VTA-2 will be located in the Fusion Materials Irradiation Test Facility (FMIT) test cell directly behind VTA-1.
Date: August 1, 1983
Creator: Vogel, M.A.
Partner: UNT Libraries Government Documents Department

Control of beryllium-7 in liquid lithium

Description: Radiation fields created by the production of /sup 7/Be in lithium of the Fusion Materials Irradiation Test (FMIT) Facility can be sufficiently high to prevent contact maintenance of system components. Preliminary experiments have shown that /sup 7/Be will adhere strongly to the FMIT piping and components and a good control method for /sup 7/Be must be developed. The initial experiments have been conducted in static stainless steel capsules and a Modified Thermal Convection Loop (MTCL). The average lithium film thickness on stainless steel was found to be 11 ..mu..m in the temperature range 495/sup 0/ to 571/sup 0/K from the capsule experiments. The diffusion coefficient for /sup 7/Be in stainless steel at 543/sup 0/K was calculated to be 5.31 x 10/sup -15/ cm/sup 2//sec. The cold leg of the MTCL picked up much of the /sup 7/Be activity released into the loop. The diffusion trap, located in the cold leg of the MTCL, was ineffective in removing /sup 7/Be from lithium, at the very slow flow rates (< 3.79 x 10/sup -4/ m/sup 3//s) used in the MTCL. Pure iron has been shown to be superior to coblat and nickel as a getter material for /sup 7/Be.
Date: December 1, 1978
Creator: Anantatmula, R. P.; Brehm, W. F.; Baldwin, D. L. & Bevan, J. L.
Partner: UNT Libraries Government Documents Department

Coupled channels model for radiative capture of nucleons by /sup 12/C. [1 to 10 MeV]

Description: A simple model based upon coupled-channels scattering calculations for nucleons on /sup 12/C was applied to the corresponding radiative capture reactions. It includes only electric dipole transitions via direct capture plus capture occurring via intermediate states consisting of only the 2/sup +/ first excited state of /sup 12/C coupled to a nucleon in the (s,d) shell. It is shown that the shape and magnitude of measured excitation functions of the /sup 12/C(p,..gamma../sub 0/) and /sup 13/C(..gamma..,n/sub 0/) reactions are largely reproduced for excitation energies up to about 10 MeV. Furthermore, it is shown that the excitation functions are strongly affected by competition and interference between direct capture and the indirect modes. Angular distribution data are also fairly well reproduced by the model. Implications of the success of the model are discussed. 9 figures.
Date: January 1, 1979
Creator: Johnson, D.L.
Partner: UNT Libraries Government Documents Department

Creep relaxation of fuel pin bending and ovalling stresses

Description: Analytical methods for calculating fuel pin cladding bending and ovalling stresses due to pin bundle-duct mechanical interaction taking into account nonlinear creep are presented. Calculated results are in close agreement with finite element results by MARC-CDC program. The methods are used to investigate the effect of creep on the FTR fuel cladding bending and ovalling stresses. It is concluded that the cladding of 316 SS 20% CW and reference design has high creep rates in the FTR core region to keep the bending and ovalling stresses to low levels.
Date: June 1, 1979
Creator: Chan, D.P. & Jackson, R.J.
Partner: UNT Libraries Government Documents Department

DALIS: a computer-assisted document retrieval system for the FFTF

Description: The FFTF (Fast Flux Test Facility) is a liquid sodium cooled, fast flux reactor designed specifically for irradiation testing of fuels and components for liquid metal fast breeder reactors. The Department of Energy and the Nuclear Regulatory Commission require that all pertinent documentation for maintenance, operation, and safety of the FFTF be readily accessible and retrievable, both during initial startup and for the lifetime of the plant. That amounts to a lot of information which has to be retrievable. The indexing system finally developed is called the DALIS system, short for Document and Location Indexing System. This system was designed by an engineer (Michael Theo) for use by engineers. DALIS uses descriptiors and keywords to identify each document in the system. The descriptors give such information as document number, date of issuance of the document, the title, the originating organization, and the microfilm or hardcopy location of the document. The keywords are words or phrases that describe the content of the document and permit retrieval by means of a computer search for documents with the stated keywords.
Date: May 12, 1981
Creator: Harves, W G
Partner: UNT Libraries Government Documents Department

Data report for the nondestructive examination of Turkey Point spent fuel assemblies B02, B03, B17, B41, and B43

Description: The fuel assembly sip test on all rods from assembly B17 were of sound intergrity. Visual examination showed three general regions of rod surface appearance: a dark adherent crud layer at the bottom of the rod, black spalling crud layers in the middle, and a spotty gray loose powdery coating with dark crud or oxide underneath towards the top. Rod lengths varied from 152.395 inches to 152.707 inches. Maximum rod bow measurements for rods G9, I9, and G7 were 0.038 to 040 inch and 0.022 inch for rods H6 and J8. Results from profilometry scans showed the maximum ovality for the five rods was 0.0066 inch with average diameters ranging from 0.4187 to 0.4198 inch. Extensive ridging from pellet-clad interaction over the entire length was evident on all rods. Gamma scan results showed no cesium peaking and no pellet gaps greater than 0.025 inch. No unusual areas, other than ridging and possible fuel-cladding bonding, were seen in the eddy current results.
Date: April 1, 1980
Creator: Davis, R.B.
Partner: UNT Libraries Government Documents Department

Deposition and control of /sup 7/Be in liquid lithium

Description: Preferential beryllium-7 deposition has been found in the higher temperature region of non-isothermal flowing lithium between 270/sup 0/C to 200/sup 0/C. Various methods for controlling /sup 7/Be distribution for application to the Fusion Materials Irradiation Test (FMIT) facility are examined. Flushing a loop with 425/sup 0/C lithium decreased /sup 7/Be activity on pipe walls by 60 to 80%. Yttrium has been found to be more effective than the other materials tested for removal of /sup 7/Be from lithium. Preliminary results on diffusion traps indicate effectiveness for /sup 7/Be removal also.
Date: January 1, 1980
Creator: Bechtold, R.A.
Partner: UNT Libraries Government Documents Department

Design and construction of the Fuels and Materials Examination Facility

Description: Final design is more than 85 percent complete on the Fuels and Materials Examination Facility, the facility for post-irradiation examination of the fuels and materials tests irradiated in the FFTF and for fuel process development, experimental test pin fabrication and supporting storage, assay, and analytical chemistry functions. The overall facility is generally described with specific information given on some of the design features. Construction has been initiated and more than 10% of the construction contracts have been awarded on a fixed price basis.
Date: November 15, 1979
Creator: Burgess, C.A.
Partner: UNT Libraries Government Documents Department

Deposition and removal of radioactive isotopes from LMFBR components

Description: The development of an analytical model to describe the production, transport and eventual removal of radioactive materials in the primary sodium of LMFBR's is a continuing Sodium Technology activity sponsored by the Department of Energy. This paper describes studies directed toward obtaining an understanding of the deposition from sodium of fuel cladding activated corrosion products onto stainless steel alloys and the effect of their diffusion into the base metal on the process required to decontaminate it. The objective of the decontamination operation is to reduce the activity to a level allowing hands on maintenance without causing unacceptable damage to the component.
Date: January 1, 1980
Creator: Hill, E.F.; Lutton, J.M. & Maffei, H.P.
Partner: UNT Libraries Government Documents Department

Design considerations for mechanical snubbers

Description: The use of mechanical snubbers to restrain piping during an earthquake event is becoming more common in design of nuclear power plants. The design considerations and qualification procedures for mechanical snubbers used on the Fast Flux Test Facility will be presented. Design precautions and requirements for both normal operation and seismic operation are necessary. Effects of environmental vibration (nonseismic) induced through the piping by pump shaft imbalance and fluid flow oscillations will be addressed. Also, the snubber dynamic characteristics of interest to design and snubber design application considerations will be discussed.
Date: March 1, 1980
Creator: Severud, L.K. & Summers, G.D.
Partner: UNT Libraries Government Documents Department