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Assessment of light water reactor fuel damage during a reactivity initiated accident

Description: This paper presents an assessment of LWR fuel damage during a reactivity initiated accident and comments on the adequacy of the present USNRC design requirements. Results from early SPERT tests are reviewed and compared with results from recent computer simulations and PBF tests. A progression of fuel rod and cladding damage events is presented. High strain rate deformation of relatively cool irradiated cladding early in the transient may result in fracture at a radial average peak fuel enthalp… more
Date: January 1, 1980
Creator: MacDonald, P. E.; Seiffert, S. L.; Martinson, Z. R.; McCardell, R. K.; Owen, D. E. & Fukuda, S. K.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Progress report, December 1, 1979-February 29, 1980

Description: Information is presented concerning bundle geometry with wrapped and bare rods; LMFBR outlet plenum flow mixing; and theoretical determination of local temperature fields in LMFBR fuel rod bundles.
Date: January 1, 1980
Creator: Todreas, N.E.; Golay, M.W. & Wolf, L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Design configuration of GCFR core assemblies

Description: The current design configurations of the core assemblies for the gas-cooled fast reactor (GCFR) demonstration plant reactor core conceptual design are described. Primary emphasis is placed upon the design innovations that have been incorporated in the design of the core assemblies since the establishment of the initial design of an upflow GCFR core. A major feature of the design configurations is that they are prototypical of core assemblies for use in commercial plants; a larger number of the … more
Date: May 1, 1980
Creator: LaBar, M.P.; Lee, G.E. & Meyer, R.J.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Development of an extended-burnup Mark B design. First semi-annual progress report, July-December 1978. Report BAW-1532-1. [PWR]

Description: The primary objective of this program is to develop and demonstrate an improved PWR fuel assembly design capable of batch average burnups of 45,000-50,000 MWd/mtU. To accomplish this, a number of technical areas must be investigated to verify acceptable extended-burnup fuel performance. This report is the first semi-annual progress report for the program, and it describes work performed during the July-December 1978 time period. Efforts during this period included the definition of a preliminar… more
Date: October 1, 1979
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Measurement and analysis of flow wall shear stress in an interior subchannel of triangular array rods. [LMFBR]

Description: A simulated model of triangular array rods with pitch to diameter ratio of 1.10 (as a test section) and air as the fluid flow was used to study the LMFBR hydraulic parameters. The wall shear stress distribution around the rod periphery, friction factors, static pressure distributions and turbulence intensity corresponding to various Reynolds numbers ranging from 4140 to 36170 in the central subchannel were measured. Various approaches for measurement of wall shear stress were compared. The meas… more
Date: August 1, 1977
Creator: Fakori-Monazah, M.R. & Todreas, N.E.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Fission product release from high gap-inventory LWR fuel under LOCA conditions

Description: Fission product release tests were performed with light water reactor (LWR) fuel rod segments containing large amounts of cesium and iodine in the pellet-to-cladding gap space in order to check the validity of the previously published Source Term Model for this type of fuel. The model describes the release of fission product cesium and iodine from LWR fuel rods for controlled loss-of-coolant accident (LOCA) transients in the temperature range 500 to 1200/sup 0/C. The basis for the model was tes… more
Date: January 1, 1980
Creator: Lorenz, R. A.; Collins, J. L.; Osborne, M. F. & Malinauskas, A. P.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

RODCON: a finite difference heat conduction computer code in cylindrical coordinates

Description: RODCON, a finite difference computer code, was developed to calculate the internal temperature distribution of the fuel rod simulator (FRS) for the Core Flow Test Loop (CFTL). RODCON solves the implicit, time-dependent forward-differencing heat transfer equation in 2-dimensional (Rtheta) cylindrical coordinates at an axial plane with user specified radial material zones and surface conditions at the FRS periphery. Symmetry of the boundary conditions of coolant bulk temperatures and film coeffic… more
Date: September 16, 1980
Creator: Conklin, J. C.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

End of life fission product distributions in F-1 experiment fuel rods

Description: Fission product migration and end-of-life distributions were examined in the F-1 (X094) series of sealed, mixed-oxide fuel rods irradiated in the fast flux of EBR-II. Cesium, rubidium, iodine, and strontium data obtained from axial gamma scanning, mass spectrometry, and radiochemical analyses are presented. The results show significant migration of cesium, probably as both a volatile species and as the noble gas precursor. Cesium metal species leaving the fuel region accumulate predominately at… more
Date: May 1, 1980
Creator: Goodin, D. T.; Langer, S. & Bell, W. E.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Test PCM-5 rod bowing and bow direction reversal. [PWR]

Description: Test PCM-5 was the first bundle test in the PCM Test Series being conducted by the Thermal Fuels Behavior Program of EG and G Idaho, Inc. as part of the Nuclear Regulatory Commission's Fuel Behavior Program. The experiment was performed in the Power Burst Facility (PBF) reactor at the Idaho National Engineering Laboratory. The bundle consisted of nine previously unirradiated PWR-type fuel rods, arranged in a 3 x 3 array within a square cross section flow shroud, with rod-to-rod spacing typical … more
Date: January 1, 1980
Creator: Kerwin, D. K.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Argonne Code Center: benchmark problem book

Description: This report is a supplement to the original report, published in 1968, as revised. The Benchmark Problem Book is intended to serve as a source book of solutions to mathematically well-defined problems for which either analytical or very accurate approximate solutions are known. This supplement contains problems in eight new areas: two-dimensional (R-z) reactor model; multidimensional (Hex-z) HTGR model; PWR thermal hydraulics--flow between two channels with different heat fluxes; multidimension… more
Date: June 1, 1977
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Gamma densitometer for measuring Pu density in fuel tubes

Description: A fuel-gamma-densitometer (FGD) has been developed to examine nondestructively the uniformity of plutonium in aluminum-clad fuel tubes at the Savannah River Plant (SRP). The monitoring technique is ..gamma..-ray spectroscopy with a lead-collimated Ge(Li) detector. Plutonium density is correlated with the measured intensity of the 208 keV ..gamma..-ray from /sup 237/U (7d) of the /sup 241/Pu (15y) decay chain. The FGD measures the plutonium density within 0.125- or 0.25-inch-diameter areas of th… more
Date: January 1, 1982
Creator: Winn, Willard G.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Investigation of stainless steel clad fuel rod failures and fuel performance in the Connecticut Yankee Reactor. Final report

Description: Significant levels of fuel rod failures were observed in the batch 8 fuel assemblies of the Connecticut Yankee reactor. Failure of 304 stainless steel cladding in a PWR environment was not expected. Therefore a detailed poolside and hot cell examination program was conducted to determine the cause of failure and identify differences between batch 8 fuel and previous batches which had operated without failures. Hot cell work conducted consisted of detailed nondestructive and destructive examinat… more
Date: November 1, 1981
Creator: Pasupathi, V. & Klingensmith, R. W.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Leaching studies using PNL 76-68 glass beads and UO/sub 2/ rods with Umtanum basalt and Nugget sandstone

Description: We have performed a 440-day leaching experiment, Bead Leach II, using PNL 76-68 glass beads and simulated uranium fuel rods in the presence of repository host rocks. The experiment was conducted in a single pass continuous-flow apparatus consisting of 72 channels. The experimental conditions were: 25/sup 0/C and 75/sup 0/C, flow rates of 1, 10, and 300 m1/d, and leachant solutions consisting of simulated basalt groundwater, brine, and sodium bicarbonate solution. The two host rocks studied were… more
Date: February 1, 1984
Creator: Bazan, F.; Rego, J.; Failor, R. & Coles, D.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Analysis of fuel relocation for the NRC/PNL Halden assemblies IFA-431, IFA-432, and IFA-513

Description: The effects of the thermally-induced cracking and subsequent relocation of UO/sub 2/ fuel pellets on the thermal and mechanical behavior of light-water reactor fuel rods during irradiation are quantified in this report. Data from the Nuclear Regulatory Commission/Pacific Northwest Laboratory Halden experiments on instrumented fuel assemblies (IFA) IFA-431, IFA-432, and IFA-513 are analyzed. Beginning-of-life in-reactor measurements of fuel center temperatures, linear heat ratings, and cladding … more
Date: April 1, 1980
Creator: Williford, R. E.; Mohr, C. L.; Lanning, D. D.; Cunningham, M. E.; Rausch, W. N. & Bradley, E. R.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Irradiation test OF-2: high-temperature irradiation behavior of LASL-made fuel rods and LASL-made coated particles. [ZrC coated particles]

Description: Three LASL-made, substoichiometric ZrC-coated particles with inert kernels, and two high-density molded graphite fuel rods that contained LASL-made, ZrC-coated fissile particles were irradiated in the Oak Ridge Research Reactor test OF-2. The severest test conditions were 8.36 x 10/sup 21/ nvt (E greater than 0.18 MeV) at 1350/sup 0/C. The graphite matrix showed no effect of the irradiation. There was no interaction between the matrix and any of the particle coats. The loose ZrC coated particle… more
Date: October 1, 1977
Creator: Wagner, P.; Reiswig, R. D.; Hollabaugh, C. M.; White, R. W.; O'Rourke, J. A.; Davidson, K. V. et al.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Progress report, September 1, 1979-November 30, 1979

Description: Information is presented concerning bundle geometry with wrapped and bare rods; subchannel geometry with bare rods; LMFBR outlet plenum flow mixing; and theoretical determination of local temperature fields in LMFBR fuel rod bundles.
Date: January 1, 1979
Creator: Todreas, N.E.; Golay, M.W. & Wolf, L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Development of annular-coated-pressurized and sphere-pac LWR fuels

Description: Annular-coated (graphite)-pressurized and sphere-pac fuel rod designs, which are expected to exhibit improved PCI-failure resistance, and, thus, more reliable extended burnup performance, are being developed. Data sufficient to provide the technical bases needed to license lead test assemblies of the improved designs for irradiation in commercial LWRs are being obtained. Out-of-reactor experiments, in-reactor instrumented experiments, in-reactor power-ramp tests, and lead-rod demonstration irra… more
Date: December 1, 1981
Creator: Buckman, F. W.; Crouthamel, C. E.; Freshley, M. D. & Barner, J. O.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Effects of fill gas composition and pellet eccentricity. [BWR]

Description: Data and an analysis are presented showing that when the operating pellet-cladding gap size of contemporary UO/sub 2/ fuel rods is carefully considered, the gap conductances are closely proportional to the thermal conductivities of the fill gases. Pellet-cladding gap eccentricity is shown to raise the gap conductance appreciably in cases of high thermal gradients across the gap. Ignoring the azimuthal heat flow can lead to an underestimation of the thermal time constant of the rod, resulting in… more
Date: July 1, 1977
Creator: Williford, R.E. & Hann, C.R.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Estimation and Control in HTGR Fuel Rod Fabrication

Description: A control algorithm has been derived for a HTGR Fuel Rod Fabrication Process utilizing the method of Box and Jenkins. The estimator is a Kalman filter and is compared with a Least Square estimator and a standard control chart. The effects of system delays are presented.
Date: January 1, 1980
Creator: Downing, D. J. & Bailey, M. J.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Assessment of degradation concerns for spent fuel, high-level wastes, and transuranic wastes in monitored retrievalbe storage

Description: It has been concluded that there are no significant degradation mechanisms that could prevent the design, construction, and safe operation of monitored retrievable storage (MRS) facilities. However, there are some long-term degradation mechanisms that could affect the ability to maintain or readily retrieve spent fuel (SF), high-level wastes (HLW), and transuranic wastes (TRUW) several decades after emplacement. Although catastrophic failures are not anticipated, long-term degradation mechanism… more
Date: January 1, 1984
Creator: Guenther, R. J.; Gilbert, E. R.; Slate, S. C.; Partain, W. L.; Divine, J. R. & Kreid, D. K.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Demonstration of fuel resistant to pellet-cladding interaction: Phase 2. Second semiannual report, July-December 1979

Description: This program has as its ultimate objective the demonstration of an advanced fuel design that is resistant to the failure mechanism known as fuel pellet-cladding interaction (PCI). Two fuel concepts are being developed for possible demonstration within this program: (a) Cu-barrier fuel and (b) Zr-liner fuel. In the current report period the nuclear design of the demonstration was begun. The design calls for 132 bundles of barrier fuel to be inserted into the core of Quad Cities Unit 2 at the beg… more
Date: March 1, 1980
Creator: Rosenbaum, H.S. (comp.)
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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