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Electromagnetic self-excitation effects in liquid metal cooled reactors

Description: For Westinghouse Hanford Co., Richland, Wash. Electromagnetic self- excitation effect is a posaible problem in LMFBR's. Geometry and flow parameters may rule it out; the magnetic Reynolds number criterion does not. The components in which the electromagnetic self-excitation effect may occur in liquid metal loops are analyzed. The conditions required for the occurrence of the effect and its consequences are described. (auth)
Date: December 1, 1973
Creator: Lessor, D.L.
Partner: UNT Libraries Government Documents Department

Precise Measurement of Process Temperature Differences

Description: Measurement of power in a nuclear reactor system is comparable to measurement of yield in a chemical plant or to measurement of throughput in a paper mill process. In most reactor systems power is determined by measurement of heat transferred to the coolant. In this study reactor coolant heat-rise was determined by the differential-temperature measuring circuitry of a power calculator which computed and recorded reactor power. This paper presents measurement techniques involved in determining the differential temperature and may be of parallel interest to instrument engineers in other process fields.
Date: January 16, 2003
Creator: Kitchen, B.G.
Partner: UNT Libraries Government Documents Department

MOV motor and gearbox performance under design basis loads

Description: This paper describes the results of valve testing sponsored by the US Nuclear Regulatory Commission, Office of Nuclear Regulatory Research and conducted at the Idaho National Engineering and Environmental Laboratory. The research objective was to evaluate the capabilities of specific actuator motor and gearbox assemblies under various design basis loading conditions. The testing was performed using the motor-operated valve load simulator, a test fixture that simulates the stem load profiles a valve actuator would experience when closing a valve against flow and pressure loadings. The authors tested five typical motors (four ac motors and one dc motor) with three gearbox assemblies at conditions a motor might experience in a power plant, including such off-normal conditions as operation at high temperature and reduced voltage. The authors also determined the efficiency of the actuator gearbox. The testing produced the following significant results: all five motors operated at or above their rated torque during tests at full voltage and ambient temperature; for all five motors (dc as well as ac), the actual torque loss due to voltage degradation was greater than the torque loss predicted using common methods; startup torques in locked rotor tests compared well with stall torques in dynamometer-type tests; the methods commonly used to predict torque losses due to elevated operating temperatures sometimes bounded the actual losses, but not in all cases; the greatest discrepancy involved the prediction for the dc motor; running efficiencies published by the manufacturer for actuator gearboxes were higher than the actual efficiencies determined from testing, in some instances, the published pullout efficiencies were also higher than the actual values; operation of the gearbox at elevated temperature did not affect the operating efficiency.
Date: June 1, 1998
Creator: DeWall, K.G.; Watkins, J.C. & Weidenhamer, G.H.
Partner: UNT Libraries Government Documents Department

Comparison of ASME Code NB-3200 and NB-3600 results for fatigue analysis of B31.1 branch nozzles

Description: Fatigue analyses wre conducted on two reactor coolant system branch nozzles in an operating PWR designed to the B31.1 Code, for which no explicit fatigue analysis was required by the licensing basis. These analyses were performed as part of resolving issues connected with NRC`s Fatigue Action Plan to determine if the cumulative usage factor (CUF) for these nozzles, using the 1992 ASME Code and representative PWR transients, were comparable to nozzles designed and analyzed to the ASME Code. Both NB-3200 and NB-3600 ASME Code methods were used. NB-3200 analyses included the development of finite element models for each nozzle. Although detailed thermal transients were not available for the plant analyzed, representative transients from similar PWRs were applied in each method. CUFs calculated using NB-3200 methods were significantly less than using NB-3600. The paper points out differences in analysis methods and highlights difficulties and unknowns in performing more detailed analyses to reduce conservative assumptions.
Date: June 1, 1996
Creator: Nitzel, M.E.; Ware, A.G. & Morton, D.K.
Partner: UNT Libraries Government Documents Department

Two-dimensional fluid-hammer analysis by the method of nearcharacteristics

Description: A numerical technique based on the method of nearcharacteristics is considered for solving propagation of fluid-hammer waves in a two-dimensional geometry. The solution is constructed by relating flow conditions by compatibility equations along lines called nearcharacteristics. Three choices are considered in the numerical scheme that are accurate within an error of the order of magnitude of the time step. Since the nearcharacteristics lie in the coordinate planes, the technique provides an efficient method requiring only simple interpolations in the initial plane. On the other hand, the nearcharacteristics fall outside the characteristics cone. Thus the solution procedure directly refers to conditions outside the true domain of dependence. The effect of this is studied through numerical calculation of a simple example problem and comparison with results obtained by a bicharacteristic method. Comparison is also made with existing analytical solutions and experiments. Furthermore, the three solution schemes considered are examined for numerical stability by the vonNeumann test. Two of the schemes were found to be unstable; the third yielded a stability criterion equivalent to that of the bicharacteristic formulation. The stability-analysis results were confirmed by numerical experimentation. (auth)
Date: May 1, 1975
Creator: Shin, Y.W. & Kot, C.A.
Partner: UNT Libraries Government Documents Department

Circulating water subsystem design description: 4 x 350 MW(t) Modular HTGR [High-Temperature Gas-Cooled Reactor] Plant

Description: The Circulating Water System is a subsystem within the Heat Rejection Group (HRG). The Circulating Water System consists of two independent loops to remove waste heat from the turbine building closed cooling water system and from the condensers associated with each turbine generator set. In normal plant operation circulating water is pumped from the cooling tower basin through the condensers and heat exchangers and back to the cooling tower where the waste heat is released to the atmosphere via mechanical draft cooling towers. The system consists of two flow paths with two half-size, vertical pumps associated with each path.
Date: June 1, 1986
Partner: UNT Libraries Government Documents Department

Statistical methods for detecting ichthyoplankton density patterns that influence entrainment mortality

Description: Samples of drifting American shad eggs were collected at two transects in the Savannah River near industrial water intakes. At each transect the river was divided into four sectors that were sampled at two hour intervals over a 24 hour period. The actual risk of entrainment was approximately 35-50% lower that if the shad eggs were uniformly distributed, and the risk of entrainment was lower at one intake than the other.
Date: December 31, 1995
Creator: Paller, M.H.; Tuckfield, R.C. & Starkel, W.M.
Partner: UNT Libraries Government Documents Department

Measurements of interfacial area concentration in two-phase flow with two-point conductivity probe. Brief communication

Description: Kataoka, Ishii and Serizawa analyzed the measurements of the local time-averaged interfacial area concentration in two-phase flow with a two-point conductivity probe. They considered the influence of the bubble velocity fluctuation on the measurement and directly transferred the mathematics concept of the local time-averaged interfacial area concentration into the measurable parameters. In the end of the derivation, however, the expression of the interfacial area concentration was inappropriate due to the over-simplification to the integration limits of the probability distributions. Consequently, the resultant interfacial area concentration may be significantly lower than the actual value. Since the formula is very important for the interpretation of experimental data, we feel it is necessary to provide a correction to the original work.
Date: February 1, 1997
Creator: Wu, Q.; Zheng, D.; Ishii, M. & Beus, S.G.
Partner: UNT Libraries Government Documents Department

Pump system characterization and reliability enhancement

Description: Pump characterization studies were performed at the Oak Ridge National Laboratory (ORNL) to review and analyze six years (1990 to 1995) of data from pump systems at domestic nuclear plants. The studies considered not only pumps and pump motors but also pump related circuit breakers and turbine drives (i.e., the pump system). One significant finding was that the number of significant failures of the pump circuit breaker exceeds the number of significant failures of the pump itself. The study also shows how regulatory code testing was designed for the pump only and therefore did not lead to the discovery of other significant pump system failures. Potential diagnostic technologies both experimental and mature, suitable for on-line and off-line pump testing were identified. The study does not select or recommend technologies but proposes diagnostic technologies and monitoring techniques that should be further evaluated/developed for making meaningful and critically needed improvements in the reliability of the pump system.
Date: September 1, 1997
Creator: Staunton, R.H.
Partner: UNT Libraries Government Documents Department

Aging management of major LWR components with nondestructive evaluation

Description: Nondestructive evaluation of material damage can contribute to continued safe, reliable, and economical operation of nuclear power plants through their current and renewed license period. The aging mechanisms active in the major light water reactor components are radiation embrittlement, thermal aging, stress corrosion cracking, flow-accelerated corrosion, and fatigue, which reduce fracture toughness, structural strength, or fatigue resistance of the components and challenge structural integrity of the pressure boundary. This paper reviews four nondestructive evaluation methods with the potential for in situ assessment of damage caused by these mechanisms: stress-strain microprobe for determining mechanical properties of reactor pressure vessel and cast stainless materials, magnetic methods for estimating thermal aging damage in cast stainless steel, positron annihilation measurements for estimating early fatigue damage in reactor coolant system piping, and ultrasonic guided wave technique for detecting cracks and wall thinning in tubes and pipes and corrosion damage to embedded portion of metal containments.
Date: December 31, 1997
Creator: Shah, V.N.; MacDonald, P.E.; Akers, D.W.; Sellers, C.; Murty, K.L.; Miraglia, P.Q. et al.
Partner: UNT Libraries Government Documents Department

Buoyancy-driven flow excursions in fuel assemblies

Description: A power limit criterion was developed for a postulated Loss of Pumping Accident (LOPA) in one of the recently shut down heavy water production reactors at the Savannah River Site. These reactors were cooled by recirculating moderator downward through channels in cylindrical fuel tubes. Powers were limited to prevent a flow excursion from occurring in one or more of these parallel channels. During full-power operation, limits prevented a boiling flow excursion from taking place. At low flow rates, during the addition of emergency cooling water, buoyant forces reverse the flow in one of the coolant channels before boiling occurs. As power increases beyond the point of flow reversal, the maximum wall temperature approaches the fluid saturation temperature, and a thermal excursion occurs. The power limit criterion for low flow rates was the onset of flow reversal. To determine conditions for flow reversal, tests were performed in a mock-up of a fuel assembly that contained two electrically heated concentric tubes surrounded by three flow channels. These tests were modeled using a finite difference thermal-hydraulic code. According to code calculations, flow reversed in the outer flow channel before the maximum wall temperature reached the local fluid saturation temperature. Thermal excursions occurred when the maximum wall temperature approximately equaled the saturation temperature. For a postulated LOPA, the flow reversal criterion for emergency cooling water addition was more limiting than the boiling excursion criterion for full power operation. This criterion limited powers to 37% of historical levels.
Date: December 31, 1995
Creator: Laurinat, J.E.; Paul, P.K. & Menna, J.D.
Partner: UNT Libraries Government Documents Department

Analysis of panthers full-scale heat transfer tests with RELAP5

Description: The RELAP5 code is being assessed on the full-scale Passive Containment Cooling System (PCCS) in the Performance ANalysis and Testing of HEat Removal Systems (PANTHERS) facility at Societa Informazioni Termoidrauliche (SIET) in Italy. PANTHERS is a test facility with fall-size prototype beat exchangers for the PCCS in support of the General Electric`s (GE) Simplified Boiling Water Reactor (SBWR) program. PANTHERS tests with a low noncondensable gas concentration and with a high noncondensable gas concentration were analyzed with RELAP5. The results showed that beat transfer rate decreases significantly along the PCCS tubes. In the test case with a higher inlet noncondensable gas fraction, the PCCS removed 35% less heat than in the test case with the lower noncondensable gas fraction. The dominant resistance to the overall heat transfer is the condensation beat transfer resistance inside the tubes. This resistance increased by about 5-fold between the inlet and exit of the tube due to the build up of noncondensable gases along the tube. The RELAP5 calculations also predicted that 4% to 5% of the heat removed to the PCCS pool occurs in the inlet steam piping and PCCS upper and lower headers. These piping needs to be modeled for other tests systems. The full-scale PANTHERS predictions are also compared against 1/400 scale GIRAFFE tests. GIRAFFE has 33% larger heat surface area, but its efficiency is only 15% and 23% higher than PANTHERS for the two cases analyzed This was explained by the high heat transfer resistance inside the tubes near the exit.
Date: January 1, 1996
Creator: Parlatan, Y.; Boyer, B.D.; Jo, J. & Rohatgi, S.
Partner: UNT Libraries Government Documents Department

The effects of parameter variation on MSET models of the Crystal River-3 feedwater flow system.

Description: In this paper we develop further the results reported in Reference 1 to include a systematic study of the effects of varying MSET models and model parameters for the Crystal River-3 (CR) feedwater flow system The study used archived CR process computer files from November 1-December 15, 1993 that were provided by Florida Power Corporation engineers Fairman Bockhorst and Brook Julias. The results support the conclusion that an optimal MSET model, properly trained and deriving its inputs in real-time from no more than 25 of the sensor signals normally provided to a PWR plant process computer, should be able to reliably detect anomalous variations in the feedwater flow venturis of less than 0.1% and in the absence of a venturi sensor signal should be able to generate a virtual signal that will be within 0.1% of the correct value of the missing signal.
Date: April 1, 1998
Creator: Miron, A.
Partner: UNT Libraries Government Documents Department

Gas evolution in reactor coolant

Description: In Testing and Irradiation Service requested that the reactor coolant be sampled in an attempt to determine the amount of air evolution in the vicinity of the K downcomer. This letter includes the data which were obtained and conclusions drawn from analysis of the data.
Date: April 21, 1966
Creator: Zimmerman, P.J.
Partner: UNT Libraries Government Documents Department

K Reactor low alum feed test

Description: The production reactors operated by Douglas United Nuclear, Inc., use treated Columbia River water as the coolant on a once through basis. Thus, radionuclides formed largely by the neutron activation of river salts are discharged to the river. One method of reducing the quality of radionuclides in the effluent is to increase the efficiency of parent isotope removal during the water treatment process. Prior to 1961 the water treatment process for preparing reactor coolant had been improved to the point that reactor quality coolant could be produced using an average alum flocculent feed rate of 6 ppM. Laboratory experiments carried out in 1959 and 1960 demonstrated that a markedly increased removal of parent isotopes resulted when alum feed rates in the neighborhood of 20 ppM were used. The results were confirmed by two half-plant tests of short duration in July, 1961, all water treatment plants began to use alum at a somewhat arbitrarily selected rate of 18 ppm. The practice Continues to date at all plants except at the K Reactors. The K Reactor alum feed has been limited to a nominal 15 ppM because of the high filtered water requirements. The use of the high alum feed rate did reduce the quantity of radionuclides in the reactor effluent. However, early in 1965 it as requested that a test be carried out at the K Reactors to determine the magnitude of the improvement. This report summarizes the results of that test.
Date: May 24, 1967
Creator: Geier, R.G.
Partner: UNT Libraries Government Documents Department

Nomogram for TAI equation

Description: A nomogram is attached as Figure 1 which may be used to determine TAI limits. The nomogram was developed to meet the need for a rapid and accurate method for calculating these limits. The technical basis for determining individual tube outlet temperature limits is given in HW-65729.
Date: December 12, 1960
Creator: Carlson, P. A.
Partner: UNT Libraries Government Documents Department