1,363 Matching Results

Search Results

Advanced search parameters have been applied.

in subjects: "Fuel Cans"

open access

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

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: Article

Partner: UNT Libraries Government Documents Department

open access

Natural convection phenomena in a nuclear power plant during a postulated TMLB' accident

After the TMI (Three Mile Island) accident, there has been significant interest in analyzing and understanding the phenomena that may occur in a PWR (Pressurized Water Reactor) accident which may lead to partial or total core meltdown and degradation. Natural convection is one of the important phenomena. In the present paper the results of two numerical simulations of (1) four-loop PWR and (2) three-loop PWR are presented. The simulations were performed with the COMMIX(2) computer code. Our ana… more

Date: January 1, 1987

Creator: Domanus, H. M.; Schmitt, R. C.; Sha, W. T.; Shah, V. L. & Han, J. T.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Integral fast reactor safety features

The Integral Fast Reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFR development effort are improved economics and enhanced safety. In addition to liquid metal cooling, the principal design features that distinguish the IFR are: (1) a pool-type primary system, (2) an advanced ternary alloy metallic fuel, and (3) an integral fuel cycle with on-site fuel reprocessing and fabrication. This paper focuses on… more

Date: January 1, 1988

Creator: Cahalan, J. E.; Kramer, J. M.; Marchaterre, J. F.; Mueller, C. J.; Pedersen, D. R.; Sevy, R. H. et al.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

POWER REACTOR HIGH PERFORMANCE UO$sub 2$ PROGRAM. Progress Report Number 9, October 1, 1968--March 31, 1969.

No Description Available.

Date: January 1, 1969

Creator: Ogawa, S. Y.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Foreign experience on effects of extended dry storage on the integrity of spent nuclear fuel

This report summarizes the results of a survey of foreign experience in dry storage of spent fuel from nuclear power reactors that was carried out for the US Department of Energy's (DOE) Office of Civilian Radioactive Waste Management (OCRWM). The report reviews the mechanisms for degradation of spent fuel cladding and fuel materials in dry storage, identifies the status and plans of world-wide experience and applications, and documents the available information on the expected long-term integr… more

Date: April 1, 1992

Creator: Schneider, K.J. & Mitchell, S.J.

Item Type: 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

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: Report

Partner: UNT Libraries Government Documents Department

open access

Postirradiation cladding strength under biaxial loading with an increasing temperature ramp. [LMFBR]

The flow behavior of unirradiated 20% cold worked AISI 316 tubing during constant pressure, increasing temperature tests was modeled with a constitutive relation approach; strain below approximately 0.2% came predominantly from an anelastic portion of the model while higher strains were predominantly plastic. The flow of cladding sections from irradiated fuel pins was largely restricted to the strain region attributed to anelastic deformation due to reduced ductility compared to unirradiated tu… more

Date: April 1, 1980

Creator: Duncan, D. R. & Hunter, C. W.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Acoustic Emission Weld Monitoring of Nuclear Components

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.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Management of waste cladding hulls. Part II. An assessment of zirconium pyrophoricity and recommendations for handling waste hulls

This report reviews experience and research related to the pyrophoricity of zirconium and zirconium alloys. The results of recent investigations of the behavior of Zircaloy and some observations of industrial handling and treatment of Zircaloy tubing and scrap are also discussed. A model for the management of waste Zircaloy cladding hulls from light water reactor fuel reprocessing is offered, based on an evaluation of the reviewed information. It is concluded that waste Zircaloy cladding hulls … more

Date: November 1, 1977

Creator: Kullen, B J; Levitz, N M & Steindler, M J

Item Type: 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

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: Report

Partner: UNT Libraries Government Documents Department

open access

Cladding creepdown under compression. [BWR; PWR]

Light-water power reactors use Zircaloy tubing as cladding to contain the UO/sub 2/ fuel pellets. In-service operating conditions impose an external hydrostatic force on the cladding, causing it to creep down into eventual contact with the fuel. Knowledge of the rate of such creepdown is of great importance to modelers of fuel element performance. An experimental system was devised for studying creepdown that meets several severe requirements by providing (1) correct stress state, (2) multiple … more

Date: November 9, 1977

Creator: Hobson, D.O.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Effect of water chemistry on the erosion-corrosion of aluminum in high temperature high velocity water

This paper reports on a laboratory study of erosion-corrosion on aluminum surfaces in high temperature water. It is essentially a continuation of a similar previous study with refinement in testing procedure and the addition of electrochemical measurements to study the phenomenon. The electrochemical procedures are of intrinsic worth, because such measurements have never before been conducted with such an unusual cell geometry as imposed by the erosion-corrosion testing apparatus.

Date: January 30, 1970

Creator: Jones, D. A.

Item Type: 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

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: Report

Partner: UNT Libraries Government Documents Department

open access

Design and safety report for HOBBIE capsules. A series of Zircaloy creepdown irradiation tests

The design and safety features of a series of irradiation experiments designed to study the in-reactor creepdown behavior of Zircaloy tubing typical of LWR fuel cladding are presented.

Date: December 1, 1977

Creator: Thoms, K.R. & van der Kaa, T.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Boundary effects on Zircaloy-4 cladding deformation in LOCA simulation tests. [PWR; BWR]

Deformation behavior of Zircaloy-4 cladding under simulated loss-of-coolant accident (LOCA) conditions is being investigated in the Multirod Burst Test (MRBT) program in single rod and multirod tests. In these tests, internally-pressurized unirradiated Zircaloy-4 tubes containing internal electrical heaters are heated to failure in a low-pressure, superheated-steam environment (200 < Re < 800). The results provide a data base for evaluating deformation and blockage models employed with design-b… more

Date: January 1, 1982

Creator: Longest, A.W.; Chapman, R.H. & Crowley, J.L.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Characterization of the sodium corrosion behavior of commercial austenitic steels

During the course of an on-going evaluation of austenitic alloys for potential liquid metal fast breeder reactor (LMFBR) fuel pin cladding application, a series of commercial alloys was selected for study. The data obtained led to the recognition of an underlying pattern of behavior and enabled the prediction of surface chemistry changes. The changes in surface topographical development from alloy to alloy are shown and the important role played by the element molybdenum in this development is … more

Date: January 1, 1980

Creator: Shiels, S. A.; Bagnall, C.; Keeton, A. R.; Witkowski, R. E. & Anantatmula, R. P.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

In-sodium creep behavior of alloys M-813 and Nimonic PE16

The in-sodium biaxial creep deformation of internally pressurized tube specimens of alloys M-813 and Nimonic PE16 was measured at 650/sup 0/C under constant stress conditions after 4000 hours of sodium exposure. Each alloy had specimens at two different stress levels, viz., 0 and 165 MPa (24,000 psi). The data showed negative diameter changes at zero stress, which were attributed to material densification associated with precipitation. Although material densification was also seen in comparable… more

Date: April 1, 1980

Creator: Anantatmula, R.P. & Gilbert, E.R.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Reference fuel studies. Seventh quarterly report May-July 1976. [LMFBR]

Task 3 of Contract AT03-76SF78003 consists of the following programs: fuel rod chemistry and thermodynamics; fuel rod engineering; fuel irradiations testing and analysis; reference structural materials. The four parts are closely interrelated and in combination are aimed at providing a sound basis for the design and performance evaluation of LMFBR mixed oxide fuel rods.

Date: August 1, 1976

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Thermal-hydraulics of the PFB/LOFT lead rod loss-of-coolant experiments. [PWR]

Results of the four PBF/LOFT Lead Rod sequential blowdown tests conducted in the Power Burst Facility (PBF) are presented. The primary objective of the test series was to evaluate the extent of mechanical deformation that would be expected to occur to low pressure (0.1 MPa), light water reactor design fuel rods subjected to a series of nuclear blowdown tests, and to determine if subjecting deformed fuel rods to subsequent testing would result in rod failure. The extent of mechanical deformation… more

Date: January 1, 1980

Creator: Varacalle, D. J. Jr.; Garner, R. W.; MacDonald, P. E. & Cox, W. R.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Analysis of the ballooning deformation of an internally pressurized thin-wall tube during fast thermal transients

A large-strain time-dependent thermoplastic analysis has been developed for the ballooning deformation of a thin-wall tube subjected to internal pressure, axial loading, and fast thermal transients. This deformation initiates with the onset of plastic instability in the material, the onset being determined by a plastic-instability criterion for strain-rate sensitive materials. The interaction among the local ballooning geometry, the state of stress, and the plastic flow process was considered, … more

Date: January 1, 1977

Creator: Lin, E.I.H.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

SSMS near surface analysis of B in irradiated Zircaloy-2: ion implantation standards as a calibration technique

Purpose of this study was to determine the amount of /sup 10/B contamination on the surface of Zircaloy-2 clad irradiated fuel elements that had been stored in an aqueous solution containing 5000 wt. ppM enriched B. SMSS indicated that the contamination was less than 0.06 ..mu..g/cm/sup 2/. (DLC)

Date: January 1, 1980

Creator: Christie, W. H.; Carter, J. A.; Eby, R. E.; Landau, L. & Musick, W. R.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Fuel rod mechanical deformation during the PBF/LOFT lead rod loss-of-coolant experiments

Results of four PBF/LOFT Lead Rod (LLR) sequential blowdown tests conducted in the Power Burst Facility (PBF) are presented. Each test employed four separately shrouded fuel rods. The primary objective of the test series was to evaluate the extent of mechanical deformation that would be expected to occur to low pressure (0.1 MPa), light water reactor design fuel rods when subjected to a series of double ended cold leg break loss-of-coolant accident (LOCA) tests, and to determine whether subject… more

Date: January 1, 1980

Creator: Varacalle, D. J. Jr.; MacDonald, P. E.; Shiozawa, S. & Driskell, W. E.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

General-purpose heat source development: Safety Verification Test Program. Titanium bullet/fragment test series

The radioisotope thermoelectric generator (RTG) that will provide power for the Galileo and Ulysses space missions contains 18 General-Purpose Heat Source (GPHS) modules. Each module contains four /sup 238/PuO/sub 2/-fueled clads and generates 250 W(t). Because the possibility of launch-pad or postlaunch explosion exists and because any explosion would generate a field of high-energy fragments, the fueled clads within each GPHS module must be able to survive fragment impact. In this test series… more

Date: June 1, 1986

Creator: George, T.G.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Acoustic-emission monitoring of LOFT fuel-cladding-burst tests

Experiments at the Loss-of-Fluid Test Facility (LOFT), beginning with experiment L6-8 (Anticipated Transient Experiment), will use a core equipped with several pressurized fuel rods. Because some of the tests may produce temperature and pressure conditions which could conceivably burst a number of rods, a nondestructive method for burst detection is needed. Acoustic emission monitoring of a number of tests of small zircaloy tubing samples, each with internal gas volume similar to that of an act… more

Date: February 1, 1982

Creator: Reinhardt, W. W.

Item Type: Report

Partner: UNT Libraries Government Documents Department