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Elimination of TOA corrosion limits

Description: In 1958, planned large scale use of the new I & E slug geometry at more severe operating conditions than had been generally experienced suggested a possible compromise in reactor life and safety if a reasonable degree of rupture control with the new type of element was not maintained. The formalized slug corrosion limit (Top-of-Annulus limit) was issued as a Process Standard at the time of the full-scale loading of I & E geometry fuel elements to provide this limit for reactor operation. The loading of I & E slugs at all reactors has been accomplished and initial power level increases have been made. To date, 67 I & E ruptures have been sustained including both `hole` and `annulus` failures. The type and behavior of ruptures to be expected with I & E geometry are now characterized. Recent studies have indicated that the I & E failure experience is consistent with the general mathematical rupture model formulated from analysis of solid slug experience. Increased confidence in the use of this model in combination with Optimization Studies permits greater emphasis to be placed on the rupture model as a guide for reactor operation. It is the purpose of this report to present the basis for substituting the rupture model for the TOA corrosion limits for rupture control purposes.
Date: May 12, 1959
Creator: Graves, S.M.
Partner: UNT Libraries Government Documents Department

FFTF transient overpower accident: a perspective

Description: This paper is a reflection on the current understanding of the unprotected transient overpower (TOP) accident, in order to place it in perspective with regard to FFTF core energetics. The experimental data base is addressed, wtih particular emphasis on the E and H-series data, and its relevance to axial failure location and hydraulic fuel sweepout is considered. It is shown that the only way in which TOP could lead to a sizeable energy release is if either total plugging takes place or a plug at the radial center of the subassemblies propagates to the hex can walls prior to neutronics shutdown. (DLC)
Date: February 1, 1975
Creator: Waltar, A.E.
Partner: UNT Libraries Government Documents Department

Effects of burnup on fuel failure. Power burst tests on fuel rods with 13,000 and 32,000 MWd/MTU burnup

Description: Results are presented from preliminary tests designed to investigate the behavior of preirradiated fuel rods under reactivity initiated accident (RIA) conditions. The tests were conducted in 1970 as part of the SPERT/Capsule Driver Core (CDC) program. The report was intended to be published in a series of Idaho Nuclear Corporation Interim Technical Reports (IN-ITRs); however, the CDC program was terminated before the report could be released. In September 1975, the Nuclear Regulatory Commission concluded that the data contained in the report could be a valuable reference in planning future water reactor safety program tests and requested its release.
Date: January 1, 1976
Creator: Miller, R.W.
Partner: UNT Libraries Government Documents Department

Quarterly progress report on the creepdown and collapse of Zircaloy fuel cladding program sponsored by the NRC Division of Reactor Safety Research for October--December 1975

Description: The creepdown and collapse study on Zircaloy fuel cladding is concerned with the deformation behavior of cladding under normal and near-normal reactor operating conditions. Progress is reported on two phases of the program: (1) collapse testing and (2) development and fabrication of deformation-monitoring equipment. Sufficient collapse testing has been done to begin the analysis of the test data. It is apparent that all three of the independent variables used in this study (pellet-to-pellet gap, pellet-to-cladding gap, and temperature) play major roles in the collapse phenomenon. Tentative trends are presented but they constitute an insufficient basis for the quantitative formulation of a collapse model at this time. The instrumentation for deformation monitoring has been developed and provision has been made for automatic control of the data gathering system and for protection of the monitoring coils from damage caused by collapse of the specimen tube.
Date: February 1, 1976
Creator: Hobson, D. O.
Partner: UNT Libraries Government Documents Department

Computation of initial stage of RBMK reactor fuel channel vessel rupture

Description: Objective of this work is estimation of temperature and time characteristics for rupture of the zirconium pipe which is the RBMK reactor fuel channel (FC) vessel under emergencies. As an emergency the zirconium pipe temperature rise process is considered which results in loss of pipe material strength properties and pipe rupture under the action of internal pressure P=80MPa. The work was carried out under Task Order 007 of University of California - VNIIEF Subcontract No. 0002P0004-95. The problem formulation is stated in Protocol (Task 3, Appendix 3) of the Russian-American Workshop which was held in December, 1994 in Los Alamos. Physical-mechanical and geometry characteristics of structure elements (FC vessel with graphite ring and graphite slug) are presented by NIKIET. The temperature mode of the structure is taken in conformity with the NIKIET data obtained with the RELAP5/MOD3 code. Numerical simulation of structure element behavior in an emergency is performed using the DRAKON program comlex oriented to solving strength problems for complex spatial structures at intense dynamic loading. The {open_quotes}DRAKON{close_quotes} program complex is described and compared with similar western codes in its capabilities.
Date: December 31, 1995
Creator: Pevnitsky, A.V.; Solovyev, V.P. & Abakumov, A.I.
Partner: UNT Libraries Government Documents Department

Net return course - operational severity index formuli

Description: This document presents a nomograph from which the relationship between reactor operating parameters, tube power, and outlet temperature can be correlated with rupture rate. The index indicates the severity of the reactor climate during irradiation and does not include the metal quality parameters defined in the rupture rate equation. The general form of the Operational Severity Index Equation is OSI=P{sup 3.3}/1000{times}t{sub 0}{sup 8.7}/100, where OSI, is the unitless Operational Severity Index, P is the tube power in kW, and t{sub 0} is the tube outlet temperature, in degrees C.
Date: December 28, 1959
Partner: UNT Libraries Government Documents Department

Mark 60B blanket tube failure

Description: A Mark 60B Li-Al target element jammed during discharge from 105-K reactor on March 17, 1975. This failure is described. Fragments from it were analyzed to determine the most probable failure mechanism: The failed target element was single extruded. When a piece of Li-rich dross fell into the mold during casting, it remained intact until coextrusion, when it penetrated or greatly thinned the cladding. The inclusion reacted with moderator during irradiation to corrode the target element and cause swelling, which ultimately caused jamming. All targets should be double extruded. (DLC)
Date: August 1, 1975
Creator: Carlson, M. K.
Partner: UNT Libraries Government Documents Department

Design of production test IP-423-A-FP evaluation of uranium fuel cores having virus heat treatments

Description: Fuel element warp occurring during the irradiation period is considered to be one of the major fuel element dimensional stability problems. Warp has been shown to correlate with accelerated corrosion attack. and also can contribute to stuck fuel charges, particularly in bumpered or self-supported charges where the annular clearances are reduced due to the presence of the projection rails. Thus, any process which offers a potential for reducing the average warp should be evaluated. Preliminary tests offsite have indicated that the use of a commercially available oil for a quench medium following beta heat treatment produces a fuel core with less residual stresses and a slightly finer and more uniform grain size than that produced by the present HAPO method of water quench. Thus, adoption of an oil quench nay offer a means whereby warp can be reduced without incurring costly revisions to equipment or fabrication processes. This report presents an irradiation testing program to evaluate the performance of oil quenched cores and to determine the optimum core heat treatment.
Date: January 5, 1962
Creator: Hodgson, W. H. & Clinton, M. A.
Partner: UNT Libraries Government Documents Department

C reactor overbore fuel examination

Description: On April 16, 1962, the fuel charge in overbore tube 3062-C sustained a failure, and upon examination after discharge was found to contain three split failures and three ``worm tracked`` elements (depression in the aluminum cladding apparently the result of uranium cleavage and subsequent yielding of the cladding). These failures occurred approximately ten days following a period of reactor neutron flux cycling, and during a second cycle at C Reactor. In addition to the failures, a total of 17 elements, from nine separate fuel charges, contained worm tracks. Four of these elements were sent to Radiometallurgy Laboratory for destructive examination, to determine the mechanism of the suspected uranium cleavage.
Date: April 18, 1963
Creator: Hladek, K. L.; Teats, R. & Weakley, E. A.
Partner: UNT Libraries Government Documents Department

Examination of enriched ruptured element from 2955C (RM 452)

Description: A CIIIE (enriched) element, one of two elements which failed April 5, 1962 in tube 2955C, was shipped to the Radiometallurgy Laboratory for detailed examination. The element was 21st from the downsteam end. A more severe rupture which occurred in the 19th piece from the downsteam end was believed to have caused the initial rupture indication. After a hot startup a heat cycle developed in the part of the reactor containing tube 2955 increasing the bulk water temperature about 20 per cent. At about the same time the temperature rise was noted, the first rupture indication was seen. The reactor was shutdown 40--45 minutes later, tube 2955 was pushed, and the two ruptures were found. The element was classed as a ``hot spot`` failure. Detailed examination was requested to determine the cause of failure and characterize the nature of any observed corrosion.
Date: August 21, 1962
Creator: Gruber, W. J.
Partner: UNT Libraries Government Documents Department

Interim goal exposure plans for O-III-NB and O-III-EB material for B, D, DR and F reactors

Description: The purpose of this report is to recommend variable goal plans for natural and enriched bumper fuel elements, specifically for 0-III-NB and 0-III-EB materials, to be irradiated at the B, D, DR, and F Reactors. The average goal exposure for all bumper fuel elements at D Reactor was specified to be 900 MWD/T, with provisions being made for revision by normal procedures. Exposures for enriched bumper material at the B, M. and F Reactors were not specified in the PITA supplement authorizing charging of this material.
Date: February 13, 1961
Creator: Bloomstrand, R. R.
Partner: UNT Libraries Government Documents Department

Slug jacket failures, January 1952

Description: There were twenty slug jacket failures during the month of January 1952. Of these, fourteen were end cap failures, five were split slugs and one has not been removed. A total of 311.4 hours of outage time was required for removal of these ruptured slugs from the reactors. The detection, removal and radiation aspects, along with the slug data, are shown on the attached sheets.
Date: February 13, 1952
Creator: Lewis, C. G.
Partner: UNT Libraries Government Documents Department

Fuel element performance

Description: For some time, it has been considered that cycling natural uranium fuel elements through the {alpha}-{beta} transformation point (approx. 662 C) is a contributing factor in split type ruptures. This hypothesis is based upon the fact that the transition temperature, a one percent increase in volume occurs. It is reasoned that as the uranium hardens under irradiation, it becomes progressively less able to adjust to the severe stresses imposed by the volume change as the material cycles through the {alpha}-{beta} transformation point. Failure finally occurs by a splitting of the core. The IPD is now interested in evaluating some of the effects of reactor operation upon fuel performance. It appears that the life of solid fuel elements may be prolonged if sudden changes in reactor operation can be minimized. It is the purpose of this report to bring to the attention of FPD management the new ideas being developed in the IPD on split type failures and to point out the effect such thinking may have on future fuel element development activities. Based on a discussion with personnel in PID, there is reasonable hope that the performance of solid fuel elements will not be limited by the {alpha}-{beta} transformation, and that the solid geometry can be used to advantage in all HAPO production reactors.
Date: December 31, 1957
Creator: Hagie, L. T.
Partner: UNT Libraries Government Documents Department

Development and evaluation of spire pulse for AlSi lead-dip canning

Description: In the AlSi process, the I&E fuel element is assembled in a two-piece aluminum container consisting of an outer cylindrical shell and an inner tube containing an integrally impacted cap wafer which forms the top end and cap during canning. Both autoclave failures and reactor failures have been attributed to porosity in the AlSi and non-wetting of the aluminum wafer with AlSi. Porosity and non-wetting in the area of the cap closure provides a pathway for water to penetrate through to the uranium from small defects in the weld which are not detected by visual weld inspection or radiography. These quality deficiencies are caused from two apparent fuel problems, (1) the disparity in mass between the spire and cap wafer which results in uneven pre-heating rates for spire and cap wafer and freezing of gas bubbles under the cap wafer, and (2) the heavyoxide and lubricant contamination on the underside and side of the wafer which causes non-wetting and outgassing. In December, 1958, an additional cleaning step was introduced in the 313 Manufacturing process to improve cap wetting. This change involved degreasing followed by a caustic etch to remove imbedded lubricants in the cap wafer. It was effective in reducing cap wetting, although it did not completely resolve the problem.
Date: January 17, 1962
Creator: Hanson, G. R.
Partner: UNT Libraries Government Documents Department

Corrosion test of irradiated uranium in monoisopropylbiphenyl (RM-171)

Description: The use of organic cooling media for nuclear reactors operating at high power levels predicates the use of a coolant which will not react violently with metallic uranium in the event of a fuel element failure. This report describes the testing, and subsequent examination, of two pieces of irradiated uranium which were immersed in monoisopropylbiphenyl (MIPB) at high temperatures and pressures for periods of time up to twenty-five days. The uranium samples had different irradiation histories and cooling times. Similar experiments had been performed with unirradiated uranium by the Corrosion and Coatings Operation, and it was wished to determine whether irradiated uranium would react with MIPB in a different manner.
Date: November 11, 1958
Creator: Brandt, R. L.
Partner: UNT Libraries Government Documents Department

Rupture Potential and Axial Power Distribution

Description: This report gives results of a study of the effect of changes in axial power distribution on rupture potential. Possible interrelationships between this effect and the effects of other reactor variables were investigated.
Date: August 11, 1959
Creator: Neef, W. I.
Partner: UNT Libraries Government Documents Department

Production Test IP-237-A, irradiation of enriched seven-rod cluster elements for ETR testing

Description: Two Zircaloy-2 jacketed seven-rod cluster elements will be irradiated in the 3674 KE front-to-rear test hole to an exposure of 1000 MWD/T and two elements will be irradiated in the 3674 KW front-to-rear test hole to an exposure of 2000 MWD/T. After irradiation, the elements will be sent to the ETR where they will be ruptured during reactor operation to determine the failure characteristics of co-extruded Zircaloy-2 jacketed cluster elements.
Date: February 23, 1959
Creator: Kratzer, W. K.
Partner: UNT Libraries Government Documents Department

KER Loop I operating report -- A report on the irradiation of zircaloy-2 jacketed tube-and-tube elements in KER Loop I

Description: The objective of this test was to determine the in-reactor performance of co-extruded Zr-2 jacketed tube-and-tube elements. The effect of power, surface temperature, and uranium alloying on dimensional stability and uranium swelling and cracking was to be measured. The charge consisted of two unalloyed tube-and-tube elements, which were Zr-2 jacketed and were 36 inches in length. The rest of the charge consisted of spacer material and coupon holders. The test started on 6-29-59 and ran for a period of 32 days. On 7-30-59, the delayed neutron monitor reading and loop activities increased to a point that required reactor shutdown scram. The reactor was shut down, and the fuel elements were discharged. Loop water samples were analyzed for fission products. Evidence was found of high activity of Iodine isotopes of very short half-life. There was also some indication of other fission products. Visual inspection of the fuel elements after discharge disclosed small wrinkles on the jacket surface of the two unalloyed elements.
Date: September 1, 1959
Creator: Eikum, L. M.
Partner: UNT Libraries Government Documents Department