Core Flow Rate Research Articles

Control of a High Temperature Supercritical Pressure Light Water Cooled and Moderated Reactor with Water Rods

The plant system of a supercritical pressure light water reactor (SCR) is once-through direct cycle. The whole coolant from the feedwater pumps is driven to the turbines. The core flow rate is less than 1/7 of that of a boiling water reactor. In the present design of the high temperature thermal reactor (SCLWR-H), the fuel assemblies contain many water rods in which the coolant flows downward. The stepwise responses of the SCLWR-H are analyzed against perturbations without a control system. Based on these analyses, a control system of the SCLWR-H is designed. The pressure is controlled by the turbine control valves. The main steam temperature is controlled by the feedwater pumps. The reactor power is controlled by the control rods. The control parameters are optimized by the test calculations to satisfy the criteria of both fast convergence and stability. The reactor is controlled stably with the designed control systems against various perturbations, such as setpoint change of the pressure, the main steam temperature and the core power, decrease in the feedwater temperature, and decrease in the feedwater flow rate.

Development of advanced core noise monitoring system for BWRS

A BWR core noise monitoring system is developed for addressing core anomaly problems in future advanced core operation. In order to monitor in-core status from a limited number of signals, various up-to-date signal processing algorithms are introduced to compensate for a lack of information. These algorithms, such as independent component analysis, factor analysis and model based parameter estimation, are demonstrated to be effective through real plant data analysis to evaluate core and regional stability index, reactivity coefficients and core flow rate. Through these practices, we demonstrate that the core noise monitoring system is an effective general platform for providing a variety of monitoring tools to meet the requirements in future advanced core operation.

Control of a High Temperature Supercritical Pressure Light Water Cooled and Moderated Reactor with Water Rods

The plant system of a supercritical pressure light water reactor (SCR) is once-through direct cycle. The whole coolant from the feedwater pumps is driven to the turbines. The core flow rate is less than 1/7 of that of a boiling water reactor. In the present design of the high temperature thermal reactor (SCLWR-H), the fuel assemblies contain many water rods in which the coolant flows downward. The stepwise responses of the SCLWR-H are analyzed against perturbations without a control system. Based on these analyses, a control system of the SCLWR-H is designed. The pressure is controlled by the turbine control valves. The main steam temperature is controlled by the feedwater pumps. The reactor power is controlled by the control rods. The control parameters are optimized by the test calculations to satisfy the criteria of both fast convergence and stability. The reactor is controlled stably with the designed control systems against various perturbations, such as setpoint change of the pressure, the main steam temperature and the core power, decrease in the feedwater temperature, and decrease in the feedwater flow rate.

Core–annular flow in a periodically constricted circular tube. Part 2. Nonlinear dynamics

Nonlinear dynamics of the concentric, two-phase flow of two immiscible fluids in a circular tube of variable cross-section is studied for parameter values where the steady core–annular flow (CAF) is linearly unstable. The simulations are based on a pseudo-spectral numerical method. They are carried out assuming axial symmetry, that the total flow rate remains constant and that all dependent variables are periodic in the axial direction, which includes the minimum necessary number of repeated units so that the obtained solution is independent of this number. The time integration originates with the numerically computed steady CAF or the steady CAF seeded with either the most unstable mode or random small disturbances. Only a limited number of the most interesting cases are presented. For the most part, the values of the majority of the dimensionless parameters are such that oil flows in the centre of the tube driven by an applied pressure gradient against gravity, whereas water is flowing in the annulus. It is shown that, whereas the steady (unstable) solution may indicate that the heavier water flows countercurrently with respect to the oil, the time periodic (observable) solution may indicate the same, albeit at a much smaller core flow rate or that concurrent flow occurs. This is due to the water being trapped between the large-amplitude interfacial waves that are generated and being convected by the oil. It is also shown that increasing the inverse Weber number increases the wave amplitude to the point that the flow of the core fluid may become discontinuous with a mechanism that depends on the viscosity ratio between the two fluids. Increasing the amplitude of the sinusoidal variation of the tube leads to a combination of travelling and standing waves, which interact to produce a time periodic solution with a long period associated with the time it takes the travelling wave to travel through the computational domain and a second much shorter period that is related to their interaction time. Qualitative agreement has been obtained upon comparing our numerical simulations with limited experimental reports, even though the experimental conditions were not identical to those in our model.

Natural circulation characteristics of a marine reactor in rolling motion and heat transfer in the core

A series of single-phase natural circulation tests in a model reactor with rolling motion was performed in order to investigate effects of the rolling motion on its thermal-hydraulic behavior. The loop flow rate in each leg changes cyclically with the rolling angle due to the inertial force of the rolling motion. As the rolling period becomes shorter, the amplitude of the loop flow rate oscillations becomes larger, and the phase lag between the rolling angle and loop flow rate oscillations becomes larger. On the other hand, the core flow rate does not oscillate, though its value changes with the rolling period. Its change correlates well with both the Reynolds number for rolling motion and the Rayleigh number, and it is considered to be caused by the change of the thermal driving head and the pressure loss through the loop. In order to simulate core flow rate change with the rolling period, a simple one-dimensional analytical model was proposed, and its verification was demonstrated. Heat transfer in the core is enhanced by the rolling motion and the enhancement is thought to be caused by the internal flow due to the rolling motion. Heat transfer coefficient in the core is well correlated with the Richardson number for rolling motion and is classified into three regimes.

Refinement of Transient Criteria and Safety Analysis for a High-Temperature Reactor Cooled by Supercritical Water

In past designs of supercritical water-cooled reactors, the core flow rate has had to be kept high enough to satisfy the minimum deterioration heat flux ratio criterion where the deterioration heat flux is a function of the core flow rate. Refinement of transient criteria related to the fuel rod design is undertaken, and new dominant transient criteria are proposed in which the cladding temperature is <610°C for Type 316 stainless steel and <840°C for Inconel 700 to reduce the balance of the plant and improve the thermal efficiency. The safety analysis for the high-temperature core using these new criteria is carried out. All the analyzed events satisfy the new criteria. A new formula for the heat transfer correlation at supercritical pressure is proposed based on numerical simulation.

Control of inner core rotation by electromagnetic, gravitational and mechanical torques

We compute the polar rotation of the Earth's inner core in response to the combined effects of electromagnetic, gravitational and mechanical torques. We obtain electromagnetic torques between 10 18 and 10 20 Nm, compared to gravitational torques between 10 19 and 10 21 Nm estimated by Xu et al. [Xu, S., Crossley D., Szeto, A.M.K., 1999. Variations in length of day and inner core differential rotation from gravitational coupling. Phys. Earth Planet. Inter., 116, 95–110] and Buffett [Buffett, B.A., 1996b. A mechanism for decade fluctuations in the length of day. Geophys. Res. Lett., 23, 3803–3806.], respectively. We obtain mechanical torques on the order of 10 15 Nm. Gravitationally dominated cases produce inner core oscillations with periods between 1 and 10 years. In electromagnetically dominated cases, the inner core rotates close to the imposed rotation rate of the outer core fluid. Modulations in the rotation rate are produced as the inner core passes through successive gravitational wells. These modulations occur over roughly 90 year timescales for outer core flow rates of 1°/year. When the electromagnetic torques are only marginally stronger than the gravitational torques, the inner core slowly rotates prograde by 90° relative to the mantle, escaping a gravitational well in roughly 100 years, then falls into the next gravitational well, rotating through 90° in just 4 years. Finite inner core viscosity is modeled using a relaxation time for the inner core topography. With relaxation, the gravitational torque reduces but does not eliminate the anomalous inner core rotation. Because of the short timescales of many of the irregularities in inner core rotation, it may be possible to observe them in presently-available seismic data.

10.2172/1062632

This is the second in a series of reports on the response of a BWR/5 boiling water reactor to anticipated transients without reactor scram (ATWS) when operating in the expanded operating domain MELLLA+. In this report ATWS events initiated by a turbine trip are considered at two points in the fuel cycle: beginning of cycle (BOC) and peak excess reactivity (PHE, close to the middle of the cycle). The effect of the gap (between the fuel pellet and clad) conductance modeling, turbine bypass fractions, and initial core flow rates are evaluated.. Two limiting values of gas-gap conductance, the lowest value of 5,000 W/m2-K at BOC and the highest value of 161,000 W/m2-K at PHE, are compared with corresponding base cases which utilized the dynamic gas-gap model. Turbine bypass fractions analyzed are 10% and 25% at PHE and these two sensitivity cases are compared with the 100% bypass base case. The reduced core flow case at 75% of nominal core flow is analyzed at PHE and compared with the base case of 85% core flow. The simulations were carried out using the TRACE/PARCS code system and models developed for a previous study at three different times during a fuel cycle with all relevant BWR/5 systems. The modeling in the core is particularly detailed (four fuel rod types included in each fuel assembly and 382 thermal-hydraulic channels to represent all assemblies taking into account half-core symmetry) in order to capture the complex neutronic-thermal-hydraulic coupling during periods of instability. The study provides insights into reactor behavior during these events, including the impact of assumed operator actions on the oscillatory behavior due to reactor instabilities and on the eventual shutdown of the reactor. It shows the effect of gap conductance, turbine bypass fraction and initial flow rate. and #8195;

An experimental investigation on penetration of buoyancy-driven cold liquid into a vertical channel with hot, forced-flow

An experimental investigation on penetration of cold liquid into a modeled subassembly channel under mixed convection, from the hot plenum of a prototypical fast breeder reactor (FBR) design has been conducted. The penetration phenomenon occurs under certain natural circulation conditions during the operation of the direct reactor auxiliary cooling system (DRACS) for decay heat removal and can influence the natural circulation head that determines the core flow rate and consequently the core cooling rate. In the present experiment, a simplified test section simulating the upper plenum and a subassembly channel was constructed wherein both temperature and velocity measurements were made to investigate the penetration of cold water into a vertical channel with upward flowing hot water. Upon comparing temperature and velocity data we found an overall similarity in their respective spatio-temporal distributions as expected. The data suggested correlations describing the onset of flow penetration and penetration depth into the channel, expressed as a combination of Reynolds and Grashof numbers for a given Prandtl number. Flow penetration was identified based on the temperature and velocity data. This work suggests a dimensionless grouping based on an order of magnitude analysis that satisfactorily casts the experimental data while noting the subtle differences with other investigations.

Flow-Induced Accident and Transient Analyses of a Direct-Cycle, Light-Water-Cooled, Fast Breeder Reactor Operating at Supercritical Pressure.

This paper deals with the safety design of a light-water-cooled fast breeder reactor operating at supercritical pressure (SCFBR) hich has the advantages of high thermal efficiency and simple reactor system. A new computer code was developed and five important flow-induced events were analyzed for the 1,245 M We SCFBR with two coolant loops. A core flow rate must be maintained for the safety of once-through type plant. Thus this reactor has two high-pressure auxiliary feedwater, four low pressure coolant injection and automatic depressurization systems. They are activated by the emergency signals of the decrease in coolant flow rate. The feedwater pumps should have fly-wheels for prolonging the coast-down time. It has been found that the peak cladding surface temperature of fuel rods does not exceed the allowable temperature under the accidents of the total loss of feedwater flow and feedwater pump shaft seizure. The minimum deterioration heat flux ratio satisfies the criterion under the transients of the ...

Development of a MELCOR Self-Initialization Algorithm for Boiling Water Reactors

The MELCOR code, developed by Sandia National Laboratories, is suitable for calculating source terms and simulating severe accident phenomena of nuclear power plants. Prior to simulating a severe accident transient with MELCOR, the initial steady-state conditions must be generated in advance. The current MELCOR users` manuals do not provide a self-initialization procedure; this is the reason users have to adjust the initial conditions by themselves through a trial-and-error approach. A MELCOR self-initialization algorithm for boiling water reactor plants has been developed, which eliminates the tedious trial-and-error procedures and improves the simulation accuracy. This algorithm adjusts the important plant variable such as the dome pressure, downcomer level, and core flow rate to the desired conditions automatically. It is implemented through input with control functions provided in MELCOR. The reactor power and feedwater temperature are fed as input data. The initialization work of full-power conditions of the Kuosheng nuclear power station is cited as an example. These initial conditions are generated successfully with the developed algorithm. The generated initial conditions can be stored in a restart file and used for transient analysis. The methodology in this study improves the accuracy and consistency of transient calculations. Meanwhile, the algorithm provides all MELCOR usersmore » an easy and correct method for establishing the initial conditions.« less

ＢＷＲ再起動計画自動作成システムの開発

system to automatically generate re-startup plans for BWR plants has been developed. By including consideration of high speed re-startup after reactor shutdown, where xenon By including consideration of high speed re-startup after reactor shutdown, where xenon By including consideration of high speed re-startup after reactor shutdown, where xenon current reactor conditions. For this purpose, not only BWR core managing expertise in a knowledge base, but also core simulation programs are used. To realize both higher accuracy and shorter computing time of core characteristics predictions by the simulation programs, the error tendency between the simulation results and actual plant data is learned using an artificial neural network. To make the plan for re-startup operation, first, allowable xenon concentrations and corresponding control rod patterns that satisfy thermal constraints at the rated power are estimated. Secondly, based on the estimation results, control rod withdrawal sequence, reactor power, control rod pattern and core flow rate for each time are determined. A prototype system was developed to evaluate its performance. The system was capable of generating a plan automatically within 30min.

Core‐annular flow of liquid membrane emulsion

AbstractA modeling and experimental study of the core‐annular flow (CAF) was carried out with a viscous liquid membrane emulsion as the core and an aqueous solution as the annulus. A hydrodynamic model has been developed for stable CAF with a centered, nondeformable core surrounded by a turbulent aqueous annulus with or without a drag reducer. This model has been verified with data from flow tests with viscous emulsions in a horizontal flow loop of 3/4‐in. (19‐mm) tubing and a fullscale, down‐hole well tubing of 2‐7/8 in. (73 mm). By the use of this model along with the data, the effects of annular size on total and core flow rates and pressure drop are investigated for annuli with and without a drag reducer.

Evaluation of core thermal and hydraulic characteristics of HTTR

Japan Atomic Energy Research Institute has started the development of the high temperature engineering test reactor (HTTR), a graphite-moderated, helium gas-cooled reactor with 30 MW thermal power and maximum outlet coolant temperature of 950 °C. This paper describes the core thermal and hydraulic (T/H) design procedure, including the validation of the computer code system, design criteria pertaining to the fuel design limit and the evaluated core T/H charateristics. The core T/H design of the HTTR has been carried out considering the specific characteristics of the core structure and the fuel based on R&D results. The coolant flow rate and temperature distribution are evaluated by the flow network analysis code flownet. The fuel temperature distribution is evaluated by the fuel temperature analysis code temdim with multi-cylindrical model using hot spot factors. Fuel design limit for anticipated operational occurrences and fuel temperature limit for normal operation are specified at 1600°C and 1495°C, respectively based on experimental results. Several design considerations are also adopted to realize a high reactor outlet coolant temperature of 950°C. As a result of core T/H design, the effective core flow rate and maximum fuel temperature during the high temperature test operation are 88% and 1492°C, respectively.

Concept and Design of a Supercritical-Pressure, Direct-Cycle Light Water Reactor

The concept of a supercritical-pressure, direct-cycle light water reactor is presented. Its feasibility is assessed by a study of its neutronic and thermal-hydraulic design. The system pressure is 250 bars. The coolant density decreases continuously in the core, and the coolant is fed directly to the turbines. This eliminates the recirculation system, steam separators, and dryers. The diameter of the reactor pressure vessel is smaller than that of a pressurized water reactor (PWR), and the vessel wall is not very thick despite the high pressure. The required core flow rate is about one-eighth that of a PWR. There are only two coolant loops in a 1,145-MW (electric) reactor, and the turbines are smaller than those of a light water reactor. These features greatly simplify the reactor plant. The thermal efficiency is improved 19% over that of a PWR.

Measuring Method for ABWR Core Flow Rate

Abstract A measuring method of core flow rate in the Advanced Boiling Water Reactor (ABWR), which equips Reactor Internal Pumps (RIPs) for reactor coolant recirculation, were evaluated by 1/5 scaled model tests and turbulent analyses. The conclusions obtained are as follows: (1) An interrelationship among RIPs was so weak that ABWR core flow was derived accurately as the summation of the individual RIP flow rate, which was estimated from each pump O-H characteristics (Q: pump flow rate, H: differential pressure head along the RIP's upstream and downstream). Measurement accuracy of core flow was confirmed to meet the requirements for expected ABWR operating conditions. Furthermore, it appeared likely that the requirements should be satisfied even in hypothetical operating states, namely operation with one pump out of service. (2) It was also confirmed that the turbulent flow analysis was useful to estimate the flow pattern near the RIPs.

Measuring Method for ABWR Core Flow Rate.

Abstract A measuring method of core flow rate in the Advanced Boiling Water Reactor (ABWR), which equips Reactor Internal Pumps (RIPs) for reactor coolant recirculation, were evaluated by 1/5 scaled model tests and turbulent analyses. The conclusions obtained are as follows: (1) An interrelationship among RIPs was so weak that ABWR core flow was derived accurately as the summation of the individual RIP flow rate, which was estimated from each pump O-H characteristics (Q: pump flow rate, H: differential pressure head along the RIP's upstream and downstream). Measurement accuracy of core flow was confirmed to meet the requirements for expected ABWR operating conditions. Furthermore, it appeared likely that the requirements should be satisfied even in hypothetical operating states, namely operation with one pump out of service. (2) It was also confirmed that the turbulent flow analysis was useful to estimate the flow pattern near the RIPs.

Thermal-hydraulic behavior of a marine reactor during oscillations

The effect of ship motion, such as heaving and rolling, on the thermal-hydraulic behavior of marine reactors was investigated. The COBRA-IV-I CODE was modified to analyse the thermal-hydraulic performance on the critical heat flux under oscillating acceleration conditions. The critical heat flux in the code was verified experimentally using freon as a comparison. The Critical Heat Flux Ratio (CHFR) at the hottest channel of the PWR subchannel was analysed using the same code. A system code RETRAN-02/MOD2-GRAV was developed by improving RETRAN-02/MOD2 to simulate the thermal hydraulic transient under ship motion. It was verified by comparison using the experimental results of both two-phase natural circulation flow under heaving motion and single-phase natural circulation flow at an inclined attitude. The code was used to analyse reactor plant behavior in the nuclear ship Mutsu. Natural circulation flow during rolling motion was investigated experimentally. The characteristics of loop flow and core flow rates were clarified. The core flow rate correlated well with the Reynolds number of rolling motion.

Natural circulation characteristics of a marine reactor in rolling motion

A series of single-phase natural circulation tests in a model reactor with rolling motion was performed, in order to investigate the natural circulation characteristics of a marine reactor in a stormy weather. It was found that the loop flow rate in each leg changes cyclically with the rolling angle due to the inertial force of the rolling motion. As the rolling period becomes shorter, both the amplitude of the loop flow rate oscillations becomes larger and the phase lag between the rolling angle and loop flow rate oscillations becomes larger. On the contrary, the core flow rate does not oscillate, though its value changes with the rolling period. Its change correlates well with both the Reynolds number for rolling motion and the Rayleigh number, and it is considered to be caused by the change of the thermal driving head and pressure loss through the loop. In order to simulate core flow rate change with the rolling period, a simple one-dimensional analytical model was proposed. The calculated results agree well with the experimental results, and its verification is demonstrated.

Development of an on-line reactor stability monitoring system in a boiling water reactor

Reactor stability is an important topic in BWR plant design and operation. The objective of the present study is to establish a method for estimating the reactor stability in an operating BWR by measuring process noise signals, such as neutron flux or core flow rate. In the present study, several methods, based on autoregressive modeling, are presented and their reliability is quantitatively evaluated, using both simulated and real plant noise data. On the basis of these examinations, an on-line stability monitoring system was developed. Through field tests on the system over a year in an operating BWR, the validity of the system has been confirmed. Also, valuable data about reactor stability has been obtained.