Showing papers in "Nuclear Technology in 1994"

A new fuel material for once-through weapons plutonium burning

Abstract: For the burning of plutonium derived from nuclear warheads, once-through type oxide fuels have been studied by considering their proliferation resistance and environmental safety as well as their technological backgrounds of fuel fabrication and reactors.From phase relations of ceramic materials and their chemical properties, it seems that a two-phase mixture of a fluorite-type phase and alumina has favorable characteristics as a once-through-type fuel of plutonium burning. It also seems that the fluorite-type phases such as thoria and fully stabilized zirconia are acceptable as host phases of plutonium because of high solid solubility of the actinide elements and fission products, irradiation stability, and chemical stability. The spent fuels finally obtained will become mineral-like waste forms, which could be buried under deep geological formations without further processing.From reactor burnup calculations with the use of the fuels, light water reactors (LWRs) with the larger volume ratio of m...

Chinese experience in the removal of actinides from highly active waste by trialkylphosphine-oxide extraction

Abstract: Mixed trialkylphosphine oxide (TRPO) (alkyl is C[sub 6]-C[sub 8]) was chosen as the extractant for the removal of uranium, neptunium, plutonium, and americium from highly active waste (HAW) in China. Composition and properties of the extractant and process chemistry are based on 30 vol% TRPO-kerosene as solvent. Hexa- and tetravalent actinides are highly extractable in 30 vol% TRPO extraction from acidic HAW, and trivalent americium (curium) can be extracted effectively from HAW with a nitric acid concentration of [approximately]1 mol/[ell]. Actinides extracted can be stripped successively by 5.5 mol/[ell] HNO[sub 3], 0.6 mol/[ell] H[sub 2]C[sub 2]O[sub 4], and 5% Na[sub 2]CO[sub 3] into americium + rare earth, neptunium + plutonium, and uranium fractions, respectively. The loading capacity of TRPO solvent is higher than that of bifunctional organophosphorus extractants, and the radiolytic stability of TRPO is higher than that of tributyl phosphate (TBP) and bis(2-ethyl hexyl)phosphoric acid. The extraction and stripping rate of TRPO is high enough to be compatible with the centrifugal contactors. Optimized process parameters of multistage countercurrent extraction and stripping and results of experimental verification are established. In both a batch experiment with simulated nuclear power plant (NPP) spent-fuel Purex HAW and a continuous experiment with real NPPmore » spent-fuel Purex HAW, 99.9% recovery of actinides was achieved. The modification of the solvent system with TBP to fit the conditions in the chemical pretreatment of defense HAW is considered.« less

Industrial scale removal of cesium with hexacyanoferrate exchanger ― process development

Abstract: An industrial scale process utilizing hexacyanoferrate-based ion exchangers was developed for the selective separation of radioactive cesium from nuclear waste solutions. This process was put into ...

Theoretical modeling of forced convection condensation of steam in a vertical tube in the presence of a noncondensable gas

Abstract: An analytical study was conducted to characterize the local condensation heat transfer coefficient of a vapor in the presence of a noncondensable gas, where the gas mixture is flowing downward insi...

Methods for the Determination of the In-Phase and Out-of-Phase Stability Characteristics of a Boiling Water Reactor

Abstract: Measurements, taken in the Ringhals-1 boiling water reactor after revision in 1990, showed that instability occurred at high power and low core flow. Measurements in several points of the power-flow map showed that the decay ratio (DR), obtained by conventional methods, jumps from a moderate value directly to unity. This was valid for DR values calculated from both average power range monitor (APRM) and local power range monitor (LPRM) signals. Thus, the conventional DR cannot be used as a measure of the margin to instability. It was found that both global (in-phase) and regional (out-of-phase) oscillations occur, the global with low DR but large signal amplitude, and the regional with high DR but low signal amplitude. The former dominates the DR calculated from both APRMs and LPRMs, except when the instability is fully developed and impedes detection of the actual margin to instability. Methods for obtaining the stability characteristics of both modes separately from neutron noise signals were developed. The DR of the out-of-phase mode appears to be a good indicator of the margin to instability.

Monitoring Feedwater Flow Rate and Component Thermal Performance of Pressurized Water Reactors by Means of Artificial Neural Networks

Abstract: The fouling of venturi meters, used for steam generator feedwater flow rate measurement in pressurized water reactors (PWRs), may result in unnecessary plant power derating On-line monitoring of these important instrument channels and the thermal efficiencies of the balance-of-plant components are addressed The steam generator feedwater flow rate and thermal efficiencies of critical components in a PWR are estimated by means of artificial neural networks The physics of these systems and appropriate plant measurements are combined to establish robust neural network models for on-line prediction of feedwater flow rate and thermal efficiency of feedwater heaters in PWRs A statistical sensitivity analysis technique was developed to establish the performance of this methodology

Effective Thermal Conductivity and Edge Conductance Model for a Spent-Fuel Assembly

Abstract: In a typical transportation or storage cask, each spent-fuel assembly resides in a square enclosure that is backfilled with a nonoxidizing gas. For design purposes, it is desirable to have a simple yet accurate method to predict the maximum fuel rod temperature in a spent-fuel assembly in these casks. An effective thermal conductivity (k[sub eff]) and an edge thermal conductance (h[sub edge]) model are developed for the interior and edge regions of a spent-fuel assembly residing in an enclosure. The model includes conductive and radiative modes of heat transfer. Predictions using the proposed k[sub eff]/h[sub edge] model are compared with five sets of experimental data for validation. The model is compared with predictions generated by the engine maintenance, assembly, and disassembly (E-MAD) and Wooten-Epstein correlations, which represent the state of the art in this field. The model is applied to a typical pressurized water reactor and a typical boiling water reactor spent-fuel assembly, and a set of both nonlinear and linear formulations of the model are derived. The proposed model is based on rigorous models of the governing heat transfer mechanisms and can be applied to a large range of assembly and enclosure types, enclosure temperatures, and assembly decay heat more » values. The proposed model is more accurate than comparable lumped correlations and is more amenable for simple, repetitive design applications than other detailed numerical models. « less

Industrial Scale Removal of Cesium with Hexacyanoferrate Exchanger—Process Realization and Test Run

Abstract: At the Loviisa Nuclear Power Station (NPS) all liquid waste, i.e., spent resins and evaporator concentrates, have been stored in a large tank storage facility. Dominating radionuclides in the evaporator concentrates have been [sup 134]Cs and [sup 137]Cs. By removing cesium from the waste, purified liquid can be released within licensed release limits, and cobalt as a second dominating nuclide is left in a small waste volume on the bottom of the tank. Since 1985, the use of inorganic hexacyanoferrate-based materials for purification of cesium has been studied. A full-scale system for cesium removal, called the IVO-CsTreat System, was constructed in 1990 to 1991. A method to produce the ion exchanger in granular form in industrial scale was developed, and the facility to produce it was constructed. The ion exchange material was produced in 1991, and the full-scale purification facility was commissioned at the Loviisa NPS in October 1991. In the test run, 253 m[sup 3] of concentrate was purified between october 31, 1991 and June 11, 1992 with three ion exchange columns, each with a volume of 8 liters. A volume reduction factor of over 10,000 was achieved as the ratio of liquid and ion exchanger volume. The decontaminationmore » factor for cesium was [approximately] 2,000.« less

Incipient fault detection and identification in process systems using accelerated neural network learning

Abstract: The objective of this paper is to present the development and numerical testing of a robust fault detection and identification (FDI) system using artificial neural networks (ANNs), for incipient (slowly developing) faults occurring in process systems. The challenge in using ANNs in FDI systems arises because of one's desire to detect faults of varying severity, faults from noisy sensors, and multiple simultaneous faults. To address these issues, it becomes essential to have a learning algorithm that ensures quick convergence to a high level of accuracy. A recently developed accelerated learning algorithm, namely a form of an adaptive back propagation (ABP) algorithm, is used for this purpose. The ABP algorithm is used for the development of an FDI system for a process composed of a direct current motor, a centrifugal pump, and the associated piping system. Simulation studies indicate that the FDI system has significantly high sensitivity to incipient fault severity, while exhibiting insensitivity to sensor noise. For multiple simultaneous faults, the FDI system detects the fault with the predominant signature. The major limitation of the developed FDI system is encountered when it is subjected to simultaneous faults with similar signatures. During such faults, the inherent limitation of pattern-recognition-based FDI methodsmore » becomes apparent. Thus, alternate, more sophisticated FDI methods become necessary to address such problems. Even though the effectiveness of pattern-recognition-based FDI methods using ANNs has been demonstrated, further testing using real-world data is necessary.« less

TRIGA Mark II benchmark experiment; Part I: Steady-state operation

Abstract: The experimental results of startup tests after reconstruction and modification of the TRIGA Mark II reactor in Ljubljana are presented. The experiments were performed with a completely fresh, compact, and uniform core. The operating conditions were well defined and controlled, so that the results can be used as a benchmark test case for TRIGA reactor calculations. Both steady-state and pulse mode operation were tested. In this paper, the following steady-state experiments are treated: critical core and excess reactivity, control rod worths, fuel element reactivity worth distribution, fuel temperature distribution, and fuel temperature reactivity coefficient.

A study on boiling water reactor regional stability from the viewpoint of higher harmonics

Abstract: A quantitative study on a mechanism for boiling water reactor regional stability has been carried out from the viewpoint of higher harmonics. In the mechanism, the gain decrease in the void-to-power transfer function can be explained by the higher harmonics mode subcriticality. It is shown that the thermal-hydraulic feedback effect can compensate for the gain decrease, and regional oscillation can be sustained that way. For quantitative evaluations, a three-dimensional higher harmonics analysis model has been developed. The results show that the first azimuthal harmonics subcriticality has a relatively small value under a regionally unstable condition. Comparing the subcriticality and the steady-state power distribution, it is shown that the distribution exists whose first azimuthal harmonics subcriticality takes a small value. A method of decomposition for the oscillated power responses into the harmonics modes is presented. The results show that the corewide oscillation power response consists almost entirely of the fundamental mode, and the regional oscillation power response consists almost entirely of the first azimuthal harmonics mode. This indicates that regional oscillation is a phenomenon in which the first azimuthal harmonics mode oscillates on the basis of the fundamental mode.

Mixed-oxide fuel fabrication technology and experience at the BELGONUCLEAIRE and CFCa plants and further developments for the MELOX plant

Abstract: Plutonium recycling in light water reactors (LWRs) has progressively become a fact. Over 200 t of mixed-oxide (MOX) fuel have been produced in the COGEMA and BELGONUCLEAIRE plants in the last 7 yr. Fuel loaded in European reactors -- mainly MIMAS fuel -- is presenting satisfactory in-core behavior and performance. Fuel fabrication technology and operation experience of the BELGONUCLEAIRE Po Dessel Plant (35 ton HM/yr) are reviewed. Backfitting of the Complex de Fabrication de Cadarache (CFCa) plant to MOX fabrication and recent fabrication progress are also addressed. The MELOX plant erected by COGEMA in Marcoule (South France) is a major commitment to provide the utilities with important additional plutonium recycling by the mid-1990s. This second generation plant has been designed to currently produce high plutonium-content MOX fuel and recycle degraded plutonium originating from high-burnup LWR UO[sub 2] fuels.

The HEVA experimental program

Abstract: Between 1983 and 1989, the Fuel Behavior Studies Branch of the Commissariat a l'Energie Atomique-Grenoble performed eight tests in the HEVA (helium and vapor) program. This program, which is a part of the general French Institute for Nuclear Protection and Safety program concerning severe accident studies, is devoted to the measurement of fission product (FP) release rates under severe accident conditions. Each test was performed with a small section (three pellets) of a standard pressurized water reactor fuel rod in its original cladding, heated in a high frequency furnace, at temperatures up to 2,300 K, in a stream and hydrogen environment. The volatile FP release rates were measured by gamma spectrometry. Posttest examinations supplied further information about the behavior of the FP, mainly concerning the aerosol sizing and the chemical speciation of the deposits. The results were compared with those obtained by other laboratories and with the calculated values. The measured release rates are generally lower than those calculated using the CORSOR model. A large influence of the environment is evidenced. The aerosol mean aerodynamic diameter is [approximately] 0.3 [mu]m. The HEVA program is extended by the VERCORS program mainly devoted to low volatile FP release rates and kinetics.

TRACG Transient Analysis Code—Three-Dimensional Kinetics Model Implementation and Applicability for Space-Dependent Analysis

Abstract: Space- and time-dependent phenomena, mostly related to neutron flux oscillations, have been observed in several boiling water reactor plants A time-dependent three-dimensional transient analysis code is indispensable for simulating such phenomena In a joint effort between the General Electric Company and the Toshiba Corporation, a three-dimensional neutron kinetics model has been implemented into the best-estimate thermal-hydraulics code, TRACG A neutronics model implementation and the applicability of the modified TRACG code for analyzing space-dependent phenomena are discussed To verify the code, startup tests with selected rod insertions, where control rods are locally inserted, are simulated Both corewide, spatially in-phase neutron flux oscillations and regional, spatially out-of-phase oscillations are modeled The results show that the modified TRACG code has sufficient capability to simulate space-dependent transients and is also a useful tool for investigating the fundamental mechanisms behind such transients

Accelerating Learning of Neural Networks with Conjugate Gradients for Nuclear Power Plant Applications

Abstract: The method of conjugate gradients is used to expedite the learning process of feedforward multilayer artificial neural networks and to systematically update both the learning parameter and the momentum parameter at each training cycle. The mechanism for the occurrence of premature saturation of the network nodes observed with the back propagation algorithm is described, suggestions are made to eliminate this undesirable phenomenon, and the reason by which this phenomenon is precluded in the method of conjugate gradients is presented. The proposed method is compared with the standard back propagation algorithm in the training of neural networks to classify transient events in neural power plants simulated by the Midland Nuclear Power Plant Unit 2 simulator. The comparison results indicate that the rate of convergence of the proposed method is much greater than the standard back propagation, that it reduces both the number of training cycles and the CPU time, and that it is less sensitive to the choice of initial weights. The advantages of the method are more noticeable and important for problems where the network architecture consists of a large number of nodes, the training database is large, and a tight convergence criterion is desired.

Monte Carlo Simulation of the Massachusetts Institute of Technology Research Reactor

Abstract: The three-dimensional continuous-energy MCNP Monte Carlo code is used to develop a versatile and accurate reactor physics model of the Massachusetts Institute of Technology research reactor II (MITR-II). The validation of the model against existing experimental data is presented. Core multiplication factors as well as fast neutron in-core flux measurements were used in the validation process. The agreement between the MCNP predictions and the experimentally determined values is very good, which indicates that the Monte Carlo model is correctly simulating the MITR-II

An intelligent man-machine system for future nuclear power plants

Abstract: The objective of the development of an intelligent man-machine system for future nuclear power plants is enhancement of operational reliability by applying recent advances in cognitive science, artificial intelligence, and computer technologies. To realize this objective, the intelligent man-machine system, aiming to support a knowledge-based decision making process in an operator's supervisory plant control tasks, consists of three main functions, i.e., a cognitive model-based advisor, a robust automatic sequence controller, and an ecological interface. These three functions have been integrated into a console-type nuclear power plant monitoring and control system as a validation test bed. The validation tests in which experienced operator crews participated were carried out in 1991 and 1992. The test results show the usefulness of the support functions and the validity of the system design approach.

Development and Comparison of a TOPAZ-II System Model with Experimental Data

Abstract: An integrated model of the TOPAZ-II space nuclear reactor system is developed and compared with measurements from the TOPAZ-II, V-71 unit tests. For a given reactor thermal power, the model calculates the coolant flow rate, temperature, and pressure throughout the system; load electric power; and overall system efficiency. Model predictions showed good agreement with the experimental data. The calculated coolant temperatures and pressure are within 15 K (< 2%) and 12% of the measurements, respectively. Analysis showed that at the nominal operating thermal power of the system (115 kW), and NaK coolant is highly subcooled. The largest subcooling of 365 K occurs at the exit of the electromagnetic pump, where coolant pressure is highest, and the lowest subcooling of 275 K occurs at the exit of the reactor core, where coolant temperature is highest.

Error Prediction for a Nuclear Power Plant Fault-Diagnostic Advisor Using Neural Networks

Abstract: The objective of this research is to develop a fault-diagnostic advisor for nuclear power plant transients that is based on artificial neural networks. A method is described that provides an error bound and therefore a figure of merit for the diagnosis provided by this advisor. The data used in the development of the advisor contain ten simulated anomalies for the San Onofre Nuclear Power Generating Station. The stacked generalization approach is used with two different partitioning schemes. The results of these partitioning schemes are compared. It is shown that the advisor is capable of recognizing all ten anomalies while providing estimated error bounds on each of its diagnoses.

Fuzzy Power Control Algorithm for a Pressurized Water Reactor

Abstract: A fuzzy power control algorithm is presented for automatic reactor power control in a pressurized water reactor (PWR). Automatic power shape control is complicated by the use of control rod...

Boiling Water Reactor Fuel Behavior at Burnup of 26 GWd/tonne U Under Reactivity-Initiated Accident Conditions

Abstract: Irradiated boiling water reactor (BWR) fuel behavior under reactivity-initiated accident (RIA) conditions was investigated in the Nuclear Safety Research Reactor (NSRR) of the Japan Atomic Energy Research Institute. Short test fuel rods, refabricated from a commercial 7 x 7 type BWR fuel rod at a burnup of 26 GWd/ tonne U, were pulse irradiated in the NSRR under simulated cooled startup RIA conditions of the BWRs. Thermal energy from 230 J/g fuel (55 cal/g fuel) to 410 J/g fuel (98 cal/g fuel) was promptly subjected to the test fuel rods by pulse irradiation within [approximately] 10 ms. The peak fuel enthalpies are believed to be the same as the prompt energy depositions. The test fuel rods demonstrated characteristic behavior of the irradiated fuel rods under the accident conditions, such as enhanced pellet cladding mechanical interaction (PCMI) and fission gas release. However, all the fuel rods survived the accident conditions with considerable margins. Simulations by the FRAP-T6 code and fresh fuel rod tests under the same RIA conditions highlighted the burnup effects on the accident fuel performance. The tests and the simulation suggested that the BWR fuel would possibly fail by a cladding burst due to fission gas release during themore » cladding temperature escalation rather than the PCMI under the cold startup RIA conditions of a severe power burst.« less

Internal flaws in the silicon carbide coating of fuel particles for high-temperature gas cooled reactors

Abstract: Internal flaws in the silicon carbide (SiC) coating of fuel particles have been characterized. The internal flaws of the SiC coating were seen as external discolored spots. The porous flaws formed circumferentially during SiC deposition. These flaws may have a harmful effect on the mechanical integrity and the diffusion barrier of the particle. The SiC coating experiments were performed under systematically selected conditions to study the mechanism of flaw formation. The most important factor influencing flaw formation was found to be the mode of particle fluidization. Internal flaws were eliminated from the particles fabricated in a mass-production coater by controlling particle fluidization.

Core design of a direct-cycle, supercritical-water-cooled fast breeder reactor

Abstract: The conceptual design of a direct-cycle fast breeder reactor (FBR) core cooled by supercritical water is carried out as a step toward a low-cost FBR plant. The supercritical water does not exhibit change of phase. The turbines are directly driven by the core outlet coolant. In comparison with a boiling water reactor (BWR), the recirculation systems, steam separators, and dryers are eliminated. The reactor system is much simpler than the conventional steam-cooled FBRs, which adopted Loeffler boilers and complicated coolant loops for generating steam and separating it from water. Negative complete and partial coolant void reactivity are provided without much deterioration in the breeding performances by inserting thin zirconium-hydride layers between the seeds and blankets in a radially heterogeneous core. The net electric power is 1245 MW (electric). The estimated compound system doubling time is 25 yr. The discharge burnup is 77.7 GWd/t, and the refueling period is 15 months with a 73% load factor. The thermal efficiency is high (41.5%), an improvement of 24% relative to a BWR's. The pressure vessel is not thick at 30.3 cm.

Numerical Analysis for Microstructure Change of a Light Water Reactor Fuel Pellet at High Burnup

Abstract: The peripheral region of a high burnup light water reactor (LWR) fuel pellet shows a microstructure that is different from the as-fabricated microstructure. The region where the microstructure change occurs (the rim region) is highly porous, and the original grains in the rim region are divided into much smaller subgrains. The electron probe microanalysis data of high burnup fuels indicate fission gas depletion in the rim region as well as in the central region. The burnup in the rim region is enhanced by built-up plutonium derived from a 238 U self-shielding effect, which is called a rim effect. The rim effect accelerates microstructure change in the peripheral region. We developed a detailed burnup analysis code ANRB computing the rim effect in LWR fuels

Transient analysis and startup simulation of a thermionic space nuclear reactor system

Abstract: The thermionic transient analysis model is used to simulate the startup of the TOPAZ-II space nuclear power system in orbit The simulated startup procedures are assumed for the purpose of demonstr

Development of software for the microsimulator for the KO-RI nuclear power plant unit 2

Abstract: A workstation-based real-time simulator for two-loop pressurized water reactor plants is developed for classroom training in support of a full-scale simulator, on-site transient analysis, and engineering studies. The present simulator consists of three functional modules: plant module, graphic module, and man-machine interaction module. The plant module includes models for the core kinetics, reactor coolant system, steam generator, main steam line, balance of plant, and control and protection system. Each of the models is optimized to obtain the capability of real-time simulation. The graphic module is designed to provide the user with more information at a glance by dynamically displaying schematic diagrams of the systems, symbols indicating the operating status of each component, trend curves, and the main control room. As tools for the man-machine interface, the man-machine interaction model uses a color cathode ray tube monitor, a standard keyboard, and the mouse. The interactive communication module receives parameters from the user via the keyboard and mouse, and transfers them to the plant module so as to enable the user to perform his specific actions. This module provides the user with various initiating events (malfunctions and manual controls) through SYSTEM, CONTROL ROOM, and ACCIDENTS menus, and thus a wide range ofmore » nuclear steam supply system transients can be easily simulated. The FISA-2/WS is verified through comparisons with analytical solutions, separated tests and integral tests, and predictions by RETRAN-2 and RELAP5/MOD3.« less

Measurement of neutron radiation exposure of commercial airline pilots using bubble detectors

Abstract: Neutron bubble detectors have been used over a 1-yr period by commercial airline pilots from Air Canada and Air France to measure the high-altitude neutron radiation exposure produced by galactic c...

An Integrated Model of the TOPAZ-II Electromagnetic Pump

Abstract: A detailed model of the electromagnetic pump of the TOPAZ-II space nuclear reactor power system is developed and compared with experimental data. The magnetic field strength in the pump depends not only on the current supplied by the pump thermionic fuel elements in the reactor core but also on the temperature of the coolant, the magnetic coil, and the pump structure. All electric and thermal properties of the coolant, wall material of the pump ducts, and electric leads are taken to be temperature dependent. The model predictions are in good agreement with experimental data.

Conceptual Design Loss-of-Coolant Accident Analysis for the Advanced Neutron Source Reactor

Abstract: A RELAP5 system model for the Advanced Neutron Source Reactor has been developed for performing conceptual safety analysis report calculations To better represent thermal-hydraulic behavior of the core, three specific changes in the RELAP5 computer code were implemented: a turbulent forced-convection heat transfer correlation, a critical heat flux (CHF) correlation, and an interfacial drag correlation The model consists of the core region, the heat exchanger loop region, and the pressurizing/letdown system region Results for three loss-of-coolant accident analyses are presented: (1) an instantaneous double-ended guillotine (DEG) core outlet break with a cavitating venturi installed downstream of the core, (b) a core pressure boundary tube outer wall rupture, and (c) a DEG core inlet break with a finite break-formation time The results show that the core can survive without exceeding the flow excursion of CHF thermal limits at a 95% probability level if the proper mitigation options are provided

TRIGA Mark II Benchmark Experiment, Part II: Pulse Operation

Abstract: Experimental results of pulse parameters and control rod worth measurements at TRIGA Mark II reactor in Ljubljana are presented. The measurements were performed with a completely fresh, uniform, an...