Boiling Water Reactor Power Plant Research Articles

Thermal analysis of closed feedwater heaters in nuclear power plants

A model is introduced for the thermal analysis of closed feedwater heaters in which wet steam is extracted from the steam turbine (and piped into the heater). Where the fluid flowing through the steam turbine is wet steam, application of this model is of relevance to nuclear power plant diagnostics. At present, all feedwater heater calculations are performed by setting the stage moisture removal effectiveness, on the assumption that the mass flows and/or the enthalpy rates of the steam bled from the turbine are known values. Our model enables detection of multiple malfunctions, even where inaccurate readings may have been registered in the control room at an earlier point in time. Furthermore, there is no immediate need for on-site inspection, which is an important advantage in Boiling Water Reactor Power Plants where feedwater heaters are placed in high radiation dose areas.

Thermo-hydraulic reasons for the Local Power Range Monitor (LPRM) spiking

Local Power Range Monitor, LPRM, spiking has always been treated as a defect in the detector or in the electronic components that adapt the core’s power signal. Until now, the only studies have been based on the installation of condensers, which is a purely reactive approach, or in other cases on seeking the root cause in the existence of tin whiskers, which is very difficult to analyse from the forensic perspective. Detailed investigations at a boiling water reactor power plant (BWR6) have studied the occurrence of spiking, as well as the power plant’s conditions upon the occurrence of power spikes. This analysis leads to the conclusion that spiking takes place under certain conditions affecting the core and recirculation loop flow. The equations that control spiking will be expanded based on this assertion. Finally, it has been checked that spiking has no connection with fuel failures, since the locations of the LPRMs with spiking do not match the locations of the failed elements.

A study on the minimum steam cooling pressure of EOP for BWR power plants

The minimum steam cooling pressure (MSCP) is an important parameter for safe operation of boiling water reactor (BWR)-type nuclear power plant for the anticipated transient without scram (ATWS) scenario with reactor pressure vessel (RPV) water level unknown. Under such situation, the operator is requested to open the safety/relief valves (SRVs) and control the RPV pressure slightly above the MSCP so that adequate core cooling can be maintained. It is derived based on steam cooling strategy. The MSCP, defined to be the lowest RPV pressure at which the covered portion of the core, is capable of generating sufficient steam to preclude peak cladding temperature (PCT) in the uncovered portion of the core from exceeding 1088 K (1500 °F). It is calculated by two parameters – (1) the minimum bundle steam flow (Wg-1500) to maintain PCT < 1088 K (1500 °F) and (2) the number of SRVs available for opening. For current emergency operating procedure (EOP), only one set of MSCP derived based on one value of Wg-1500 for the ATWS condition. Furthermore, it is derived based on decay power of 2.2% rated power. Thus, the current MSCP used for the ATWS accident scenarios was deemed inadequate. The purpose of this paper (work) is to study the MSCP used in the ATWS conditions. For case of ATWS of 13% full power, controlling RPV pressure at MSCP of current approach ends up with core melt. The Wg-1500 is suggested to be replaced by the steam generation rate at minimum steam cooling RPV water level (MSCRWL), which is a function of power level. Simulation result indicates controlling RPV pressure at MSCP is equivalent to controlling the RPV water level at MSCRWL. The revised MSCP is dependent on the ATWS power level.

Shell side heat transfer characteristics with an eggcrate support plate for water in parallel flow

The shell side heat transfer and pressure drop for water in parallel flow with an eggcrate support plate were experimentally investigated in order to obtain higher performance for the heat exchanger in a boiling water reactor power plant. The following three conclusions were reached. (1) The shell side heat transfer characteristics with the eggcrate support plate were twice as large as those of the shell side parallel flow. An equation using the Reynolds number of the eggcrate support plate could predict the heat transfer coefficient. (2) The shell side pressure drop characteristics with the eggcrate support plate were about five to six times as large as those of shell side parallel flow. (3) The enhancement constant of heat transfer with the eggcrate support plate, using Colburn's j-factor and friction factor f, was the same as that of the ROD-baffle type, and was about two times as large as that of the segmental baffle type. © 2000 Scripta Technica, Heat Trans Asian Res, 29(2): 91–112, 2000

Shell side heat transfer characteristics with an eggcrate support plate for water in parallel flow

The shell side heat transfer and pressure drop for water in parallel flow with an eggcrate support plate were experimentally investigated in order to obtain higher performance for the heat exchanger in a boiling water reactor power plant. The following three conclusions were reached. (1) The shell side heat transfer characteristics with the eggcrate support plate were twice as large as those of the shell side parallel flow. An equation using the Reynolds number of the eggcrate support plate could predict the heat transfer coefficient. (2) The shell side pressure drop characteristics with the eggcrate support plate were about five to six times as large as those of shell side parallel flow. (3) The enhancement constant of heat transfer with the eggcrate support plate, using Colburn's j-factor and friction factor f, was the same as that of the ROD-baffle type, and was about two times as large as that of the segmental baffle type. © 2000 Scripta Technica, Heat Trans Asian Res, 29(2): 91–112, 2000

Shell Side Heat Transfer Characteristics to Water Flowing Parallel with an Eggcrate Support Plate.

The shell side heat transfer and pressure drop to water flowing parallel with the Eggcrate Support Plate was experimentally investigated in order to get higher performance for the cooler in a Boiling Water Reactor power plant. The following three conclusions were reached in the study. (1) The shell side heat transfer characteristics with the Eggcrate Support Plate were twice as large as those of the shell side parallel flow. An equation using the Reynolds number of the Eggcrate Support Plate could predict the heat transfer coefficient. (2) The shell side pressure drop characteristics with the Eggcrate Support Plate were about five to six times as large as those of the shell side parallel flow. (3) The enhancement constant of heat transfer with the Eggcrate Support Plate, using the Colburn's j factor and friction factor f, was the same as that of the ROD-baffle type, and was about two times as large as that of the segmental baffle type.

The study of CRUD removal resin XT-1033

In order to reduce the occupational radiation load at BWR (Boiling Water Reactor) and ATR (Advanced Thermal Reactor) type Nuclear Power Plants, it is important to improve the iron removal efficiency at the CD (condensate demineralizer) of feed water system.The CRUD removal efficiency of CD depends on the cation exchange resin's ability which is known to be improved with time.The newly developed cation exchange resin “XT-1033” shows a excellent CRUD removal efficiency in the study of applocation at actual plant from fresh age.

Experimental evaluation of an operator decision aid system for BWR power plants

An experiment was carried out to evaluate the effectiveness of an integrated operator decision aid (IODA) system for boiling water reactor (BWR) power plants. The IODA was developed to enhance the operating safety, reliability and quality of plants. It assists an operator's monitoring and diagnostic capabilities under adverse plant situations using computers and color CRT display devices. Three functions—a standby systems management system (SSMS), a disturbance analysis system (DAS), and a post-trip operational guidance (PTOG) system—have been developed as aids corresponding to the operator's various roles in the safe operation of a nuclear power plant. These functions were designed with full consideration of the operator's problem-solving process, and they have been integrated into a comprehensive IODA system for abnormal BWR power plant conditions. To ensure the applicability of the test results to real-life situations, emphasis was placed on establishing realistic test conditions—including a fully equipped experimental control room, a full-scope plant simulator, participation of experienced operating crews, and well-designed test transient scenarios. Experimental data were collected by means of computers and audio-visual devices, and were processed and summarized to facilitate analysis. Qualitative analysis of the test results was performed with emphasis on the decision-making process of an operator. The analysis points to the utility of the IODA as an operator aid, especially in diagnosing adverse plant situations and in formulating strategies for countermeasures.

Operational Margin Monitoring System for Boiling Water Reactor Power Plants

This paper reports on an on-line operational margin monitoring system which has been developed for boiling water reactor power plants to improve safety, reliability, and quality of reactor operation. The system consists of a steady-state core status prediction module, a transient analysis module, a stability analysis module, and an evaluation and guidance module. This system quantitatively evaluates the thermal margin during abnormal transients as well as the stability margin, which cannot be evaluated by direct monitoring of the plant parameters, either for the current operational state or for a predicted operating state that may be brought about by the intended operation. This system also gives operator guidance as to appropriate or alternate operations when the operating state has or will become marginless.

Microprocessor-based fault-tolerant reactor control and information system

The reactor manual control system (RMCS) and the rod position information system (RPIS) applied to boiling water reactor power plants are among the most important systems in controlling the output of nuclear reactor power. A highly reliable RMCS and RPIS employing microprocessors is presented. The developed equipment has been made fault tolerant by adopting the redundancy of each system. The amount of cabling normally required has been reduced by multiplexing transmission via fiber-optic cable. The size of the control panel has been reduced and maintainability improved. The system hardware and software have been standardized into units and modules, which provides the equipment with sufficient flexibility to cope with plants with a varying number of control rods. This system can be applied to the replacement of conventional systems in operating plants as well as initial installation in new plants. >

Practical expression for exception and its utilization to leakage identification of power plant

The aphorism “Exceptio probat regulam”, which means that the exception can serve as an examiner of the rule, motivated the authors to introduce the exceptional proposition in simply-structured diagnosis. The cancellation law in binary logic was fuzzified to define a practical technique for calculating the linguistic truth values appearing in the exception. Research involved a trial to ascertain whether the fuzzy reasoning with participation of the exception was effective or not in identifying leaking locations in the cooling systems of a Boiling Water Reactor power plant. Several examples were examined to demonstrate the potential for their mixed utilization. This result led us to the finding that the present fuzzy diagnosis is capable of dealing with a large number of fragmentary items of information in order to reinforce the diagnostic quality.

非干渉I-PD制御系による沸騰水型原子力発電プラントの給水ポンプ切り替え

非干渉I-PD制御系による沸騰水型原子力発電プラントの給水ポンプ切り替え

A Real-Time Expert System for Nuclear Power Plant Failure Diagnosis and Operational Guide

A real-time expert system (DIAREX) has been developed to diagnose plant failure and to offer a corrective operational guide for boiling water reactor (BWR) power plants. The failure diagnosis model used in DIAREX was systematically developed, based mainly on deep knowledge, to cover heuristics. Complex paradigms for knowledge representation were adopted, i.e., the process representation language and the failure propagation tree. The system is composed of a knowledge base, knowledge base editor, preprocessor, diagnosis processor, and display processor. The DIAREX simulation test has been carried out for many transient scenarios, including multiple failures, using a real-time full-scope simulator modeled after the 1100-MW(electric) BWR power plant. Test results showed that DIAREX was capable of diagnosing a plant failure quickly and of providing a corrective operational guide with a response time fast enough to offer valuable information to plant operators.

Design Principles of a Simple and Safe 200-MW(thermal) Nuclear District Heating Plant

Kraftwerk Union AG has almost completed the development of a dedicated 200-MW(thermal) nuclear district heating plant to provide environmentally clean energy at a predictably low cost. The concept can easily be adapted to meet power requirements within the 100- to 500-MW(thermal) range. This technology is the product of the experience gained with large pressurized water reactor and boiling water reactor power plants, with respect to both plant and fuel performance. The major development task is that of achieving sufficiently low capital cost by tailoring components and systems designed for large plants to the specific requirements of district heating. These requirements are small absolute power, low temperatures and pressures, and modest load following, all of which result in the characteristics that are summarized. A fully integrated primary system with natural circulation permits a very compact reactor building containing all safety-related systems and components. Plant safety is essentially guaranteed by inherent features. The reactor containment is tightly fitted around the reactor pressure vessel in such a way that, in the event of any postulated coolant leak, the core cannot become uncovered, even temporarily. Shutdown is assured by gravity drop of the control rods mounted above the core. Decay heat is removed from the core by means of natural circulation via dedicated intermediate circuits of external aircoolers. The design work for systems, components, and structures has almost been completed. Detailed accident analyses have demonstrated that the plant’s safety characteristics are excellent.

Optimization of Fuel Exchange Machine Operation for Boiling Water Reactors Using an Artificial Intelligence Technique

Optimization of fuel assembly exchange machine movements during periodic refueling outage is discussed. The fuel assembly movements during a fuel shuffling were examined, and it was found that the fuel assembly movements consist of two different movement sequences; one is the “PATH,” which begins at a discharged fuel assembly and terminates at a fresh fuel assembly, and the other is the “LOOP,” where fuel assemblies circulate in the core. It is also shown that fuel-loading patterns during the fuel shuffling can be expressed by the state of each PATH, which is the number of elements already accomplished in the PATH actions. Based on this fact, a scheme to determine a fuel assembly movement sequence within the constraint was formulated using the artificial intelligence language PROLOG. An additional merit to the scheme is that it can simultaneously evaluate fuel assembly movement, due to the control rods and local power range monitor exchange, in addition to normal fuel shuffling. Fuel assembly movements, for fuel shuffling in a 540-MW(electric) boiling water reactor power plant, were calculated by this scheme. It is also shown that the true optimization to minimize the fuel exchange machine movements would be costly to obtain due to the number of alternatives that would need to be evaluated. However, a method to obtain a quasi-optimum solution is suggested.

Time Behavior of the Neutron Flux at the Vessel Boundary of a Boiling Water Reactor Power Plant

The results of a recent measurement of the neutron fluence at the vessel outer boundary of the Caorso Nuclear Power Plant show that the neutron flux per unit power was not constant during the reactor operation. This effect, usually not mentioned, may induce serious errors in the results of neutron dosimetry by the multiple foil activation technique.

Variable Structure Controller for Direct Cycle Boiling Water Reactor Power Plant

This paper proposes a new control strategy based on the concept of variable structure systems for the design of controller for a direct cycle boiling water reactor power plant with recirculation control so that the power plant exhibits load following characteristics. A special feature of the variable structure controller is that it can effectively improve the transient performance of the system while at the same time ensuring zero steady state error. Further, when the variable structure system is operated in the sliding mode, the system exhibits low sensitivity to plant parameter variations. Digital simulation results indicate that the proposed controller yields better system dynamic performance than that obtained by using a linear optimal controller.

A Symptom-Oriented Method for Posttrip Operation Guidance in Boiling Water Reactors

A symptom-oriented method for providing operational guidelines under postrip conditions is developed to support operators of boiling water reactor power plants. The method is dedicated to providing the operator with two levels of operational guidelines for a plant cold shutdown. Level I involves a selection of basic operating tactics, and level II involves a determination of operation priority and a confirmation of operation results. Technical evaluation of the guidance algorithm is made using off-line simulation tests postulding the events in loss-of-feedwater and loss-ofcoolant accidents. Results show that the developed method is adequate.

Validation of a Computerized Operator Support System

This paper presents the experimental validation of a computerized operator support system (COSS) for BWR (Boiling Water Reactor) power plants. To assess the effects of the COSS on operator performance, a realistic experimental control room and a full scope BWR plant simulator were developed. The control room includes a main and two auxiliary control panels equipped with color cathode-ray tubes (CRT's), control switches, annunciators and paging devices. A supervisory console with CRT's is also installed in the control room. Test transients were selected by taking account of their severity, complexity, generality, duration time and relevance to the functions of the COSS. Experimental data of the operating crew's performance and the plant status during simulated transients were collected on-line by using computers and audio/video tape recorders. The data were processed and transformed into sunmarized forms to facilitate analysis. Questionnaires and interviews were performed to collect subjective data for analyzing and decision making process of the operators. The results of the experiment show the usefulness of the COSS as an operator aid, especially in diagnosing the causes of disturbances and in establishing counteractive strategies

A new digital control strategy for a nuclear boiling water reactor

A method is proposed for the design of constant gain optimal feedback control laws for discrete-time systems which reduce trajectory sensitivity to small system parameter variations and ensure the closed-loop eigenvalues at the prescribed locations. The selected performance index is of standard linear regulator type, modified to include a quadratic sensitivity term. An efficient computational procedure based on direct cost oprimization using gradient type algorithm is also reported. A case study of a Boiling Water Reactor power plant has been presented to illustrate the application of the design procedure.