10,499 Matching Results

Search Results

Advanced search parameters have been applied.


Description: Spert IV is a large pool-type experimental facility for reactor kinetic studies. These studies will include power excursion and instability tests for a variety of reactor designs. Since the Spert IV experimental program requires the performance of tests which will approach, and may exceed the threshold of reactor destruction, the probability of occurrence of the maximum possible accident is not negligible compared with that of other possible accidents. The maximum possible accident for this facility is considered to be a severe nuclear excursion which results in the destruction of the reactor building and the release of 100% of the accumulated fission product inventory of the atmosphere in a steam cloud. The fission product source assumed in the analysis of this accident is an upper limit in view of the nature of the tests to be performed and the heat removal capacity of the system. This postulated accident is independent of the details of core and control system design and is valid for all cores anticipated for use in the experimental program. The major hazards present in the operation of this facility, the precautions to be taken to reduce the probability of an accident, and the consequences of the maximum possible accident are discussed. It is concluded that the proposed method of operation will minimize the hazard to operating personnel, and that the site location will make possible the operation of the Spent IV facility without hazard to the general public. (auth)
Date: July 1, 1961
Creator: Bentzen, F. L. & Crocker, J. G.
Partner: UNT Libraries Government Documents Department


Description: An attempt was made to predict the operating performance of EBWR at various power levels, and the results are compared with the experimental data obtained in raising the EBWR to a power of 100 Mw. An IBM-704 code, RECHOP, was used to calculate the recirculating velocity; experimental subcooling data were fed into RECHOP at high power levels. It is concluded that the prediction shows the general trend of the major parameters but requires further work for producing accurate answers. (D.L.C.)
Date: October 1, 1963
Creator: Nentwich, A.A.
Partner: UNT Libraries Government Documents Department


Description: The objectives of an oxide core destructive test program at the Spert I reactor facility are reviewed. The proposed experimental program of destructive tests on a low-enriched oxide core, the experimental results of nondestructive transient tests that were obtained on the test core and the extrapolation of these results to the destructive case, an analysis of the hazards involved in performing such destructive tests, and a detailed description of the reactor facility and environmental conditions are presented. The supervision and control of personnel during and after each destructive test, and of the plans for reentry, cleanup, and restoration of the facility are discussed. The water- moderated core that will be used for these experiments is mounted in the Spert I open-type reactor vessel, which has no provision for pressurization or forced coolant flow. The core is comprised of approximately 600 4%-enriched UO/sub 2/ fuel rods clad with stainless steel, and four blade-type, gang-operated control rods for reactor control. Reactor transients are initiated at ambient temperature by step-insertions of reactivity, using for this purpose a special control rod that can be quickly ejected from the core. On the basis of the results obtained from previous nondestructivelkinetic tests, an analysis was made to determine the nature of the results to be expected for an assumed 1.8- sec- period test in which tetal core destruction occurs. An evaluation of hazards involved in conducting the 1.8-msec test, based on conservative assumptions regarding fission product release and weather conditions, indicated that with the procedural controls normally exercised in the conduct of any transient test at Spert and the special controls to be in effect during the destructive test series, no significant hazard to personnel or to the general public will be incurred. (auth)
Date: August 30, 1963
Creator: Grund, J.E. & Norton, B.E.
Partner: UNT Libraries Government Documents Department

XENON TRANSIENT TESTS. CORE I, SEED 2, EFPH 1565.4. Section 1. Test Results T-612081

Description: Tests were carried out to determine if there was sufficient excess reactivity present in the core to override a pesk xenon transient and to obtain data for rod worth calculations. Upon completion of 1565.4 EFPH of plant operation of Core I, Seed 2, there was sufficient reactivity to override the peak xenon transient imposed by a rapid shutdown from an average reactor power level of 102.65%. The override occurred 8 hr and 46 min after shutdown with Group I control rods withdrawn to 69 in. and Group II control rods controlling, withdrawn to 44.75 in. Rod Group III and Group IV were fully inserted. (M.C.G.)
Date: January 30, 1961
Partner: UNT Libraries Government Documents Department


Description: Pressure-drop measurements were made across a mockup of a Hallam prototype fuel element in a test section installed in the Hallam Hydraulic Loop. The flow channel was identical to an SRE fuel channel and included simulated upper and lower plenums. The fuel element mockup was equipped with a Hallam-type variable orifice at the channel exit and a fixed orifice in the strainer basket at the bottom of the element. Tests were performed to determine the optimum size for the fixed orifice and the temperature adjustment capability of the variable orifice using this optimum fixed orifice. To obtain the predicted 4.1 lb/sec sodium coolant requirement at a core pressure drop of 1.85 psi, a 3/4 in. fixed orifice was determined to be the optimum. With this fixed orifice size the variable orifice will be approximately 1 in. withdrawn during full-power operation. Adjusting the orifice over its entire range of 3 in. from fully inserted to fully withdrawn covers a temperature range from 875 to 1040 deg F which is approximately plus or minus 80'F about the normal outlet temperature of 960 deg F. Curves are presented for use in determination of operating characteristics of the element with other fixed orifice sizes should the core pressure drop or required flow rate of coolant be changed. (auth)
Date: March 13, 1961
Creator: Begley, R.J.
Partner: UNT Libraries Government Documents Department