Patent Application
20210027901
Generally, the present disclosure relates to the field of data processing. More specifically, the present disclosure relates to methods and systems for facilitating the management of reactor transient conditions associated with reactors
Existing techniques for facilitating the management of reactor transient conditions associated with reactor are deficient with regard to several aspects. For instance, current technologies do not analyze reactor data in real-time following a reactor transient condition (dynamic operating transient). Furthermore, current technologies do not generate evaluations following the reactor transient condition. Moreover, current technologies do not communicate and distribute evaluations to reactor personnel. Further, current technologies do not provide confirmatory information regarding the reactor transient conditions. Further, current technologies do not provide recommended action corresponding to the reactor following the reactor transient condition.
Therefore, there is a need for improved methods, systems, apparatuses and devices for facilitating the management of reactor transient conditions associated with reactors that may overcome one or more of the above-mentioned problems and/or limitations.
This summary is provided to introduce a selection of concepts in a simplified form, that are further described below in the Detailed Description. This summary is not intended to identify key features or essential features of the claimed subject matter. Nor is this summary intended to be used to limit the claimed subject matter's scope.
Disclosed herein is a method of facilitating the management of reactor transient conditions associated with reactors, in accordance with some embodiments. Accordingly, the method may include a step of receiving, using a communication device, at least one reactor data associated with a reactor from a reactor computer. Further, the method may include a step of determining, using a processing device, at least one reactor transient condition associated with the reactor based on the at least one reactor data. Further, the method may include a step of receiving, using the communication device, a plurality of reactor design data and a plurality of reactor measurement data associated with a plurality of reactor components of the reactor from the reactor computer. Further, the method may include a step of analyzing, using the processing device, the plurality of reactor design data and the plurality of reactor measurement data. Further, the method may include a step of generating, using the processing device, at least one notification corresponding to the at least one reactor transient condition based on the analyzing. Further, the method may include a step of transmitting, using the communication device, the at least one notification to at least one user device associated with at least one user.
Further, disclosed herein is a method of facilitating the management of reactor transient conditions associated with reactors, in accordance with some embodiments. Accordingly, the method may include a step of receiving, using a communication device, at least one reactor data associated with a reactor from a reactor computer. Further, the method may include a step of determining, using a processing device, at least one reactor transient condition associated with the reactor based on the at least one reactor data. Further, the method may include a step of analyzing, using the processing device, the at least one transient condition. Further, the method may include a step of identifying, using the processing device, at least one reactor component of the plurality of the reactor components based on the analyzing. Further, the method may include a step of receiving, using the communication device, at least one reactor design data and at least one reactor measurement data corresponding to the at least one reactor component. Further, the method may include a step of evaluating, using the processing device, the at least one reactor design data and the at least one reactor measurement data. Further, the method may include a step of generating, using the processing device, at least one notification corresponding to the at least one reactor transient condition based on the evaluation. Further, the method may include a step of transmitting, using the communication device, at least one notification to at least one user device associated with at least one user.
Further disclosed herein is a system for facilitating the management of reactor transient conditions associated with reactors, in accordance with some embodiments. Accordingly, the system may include a communication device communicatively coupled with a reactor computer associated with a reactor. Further, the communication device may be configured for receiving at least one reactor data associated with the reactor from the reactor computer. Further, the communication device may be configured for receiving a plurality of reactor design data and a plurality of reactor measurement data associated with a plurality of reactor components of the reactor from the reactor computer. Further, the communication device may be configured for transmitting at least one notification to at least one user device associated with at least one user. Further, the system may include a processing device configured for determining at least one reactor transient condition associated with the reactor based on the at least one reactor data. Further, the processing device may be configured for analyzing the plurality of reactor design data and the plurality of reactor measurement data. Further, the processing device may be configured for generating the at least one notification corresponding to the at least one reactor transient condition based on the analyzing.
Both the foregoing summary and the following detailed description provide examples and are explanatory only. Accordingly, the foregoing summary and the following detailed description should not be considered to be restrictive. Further, features or variations may be provided in addition to those set forth herein. For example, embodiments may be directed to various feature combinations and sub-combinations described in the detailed description.
The accompanying drawings, which are incorporated in and constitute a part of this disclosure, illustrate various embodiments of the present disclosure. The drawings contain representations of various trademarks and copyrights owned by the Applicants. In addition, the drawings may contain other marks owned by third parties and are being used for illustrative purposes only. All rights to various trademarks and copyrights represented herein, except those belonging to their respective owners, are vested in and the property of the applicants. The applicants retain and reserve all rights in their trademarks and copyrights included herein, and grant permission to reproduce the material only in connection with reproduction of the granted patent and for no other purpose.
Furthermore, the drawings may contain text or captions that may explain certain embodiments of the present disclosure. This text is included for illustrative, non-limiting, explanatory purposes of certain embodiments detailed in the present disclosure.
As a preliminary matter, it will readily be understood by one having ordinary skill in the relevant art that the present disclosure has broad utility and application. As should be understood, any embodiment may incorporate only one or a plurality of the above-disclosed aspects of the disclosure and may further incorporate only one or a plurality of the above-disclosed features. Furthermore, any embodiment discussed and identified as being “preferred” is considered to be part of a best mode contemplated for carrying out the embodiments of the present disclosure. Other embodiments also may be discussed for additional illustrative purposes in providing a full and enabling disclosure. Moreover, many embodiments, such as adaptations, variations, modifications, and equivalent arrangements, will be implicitly disclosed by the embodiments described herein and fall within the scope of the present disclosure.
Accordingly, while embodiments are described herein in detail in relation to one or more embodiments, it is to be understood that this disclosure is illustrative and exemplary of the present disclosure, and are made merely for the purposes of providing a full and enabling disclosure. The detailed disclosure herein of one or more embodiments is not intended, nor is to be construed, to limit the scope of patent protection afforded in any claim of a patent issuing here from, which scope is to be defined by the claims and the equivalents thereof. It is not intended that the scope of patent protection be defined by reading into any claim limitation found herein and/or issuing here from that does not explicitly appear in the claim itself.
Thus, for example, any sequence(s) and/or temporal order of steps of various processes or methods that are described herein are illustrative and not restrictive. Accordingly, it should be understood that, although steps of various processes or methods may be shown and described as being in a sequence or temporal order, the steps of any such processes or methods are not limited to being carried out in any particular sequence or order, absent an indication otherwise. Indeed, the steps in such processes or methods generally may be carried out in various different sequences and orders while still falling within the scope of the present disclosure. Accordingly, it is intended that the scope of patent protection is to be defined by the issued claim(s) rather than the description set forth herein.
Additionally, it is important to note that each term used herein refers to that which an ordinary artisan would understand such term to mean based on the contextual use of such term herein. To the extent that the meaning of a term used herein—as understood by the ordinary artisan based on the contextual use of such term—differs in any way from any particular dictionary definition of such term, it is intended that the meaning of the term as understood by the ordinary artisan should prevail.
Furthermore, it is important to note that, as used herein, “a” and “an” each generally denotes “at least one,” but does not exclude a plurality unless the contextual use dictates otherwise. When used herein to join a list of items, “or” denotes “at least one of the items,” but does not exclude a plurality of items of the list. Finally, when used herein to join a list of items, “and” denotes “all of the items of the list.”
The following detailed description refers to the accompanying drawings. Wherever possible, the same reference numbers are used in the drawings and the following description to refer to the same or similar elements. While many embodiments of the disclosure may be described, modifications, adaptations, and other implementations are possible. For example, substitutions, additions, or modifications may be made to the elements illustrated in the drawings, and the methods described herein may be modified by substituting, reordering, or adding stages to the disclosed methods. Accordingly, the following detailed description does not limit the disclosure. Instead, the proper scope of the disclosure is defined by the claims found herein and/or issuing here from. The present disclosure contains headers. It should be understood that these headers are used as references and are not to be construed as limiting upon the subjected matter disclosed under the header.
The present disclosure includes many aspects and features. Moreover, while many aspects and features relate to, and are described in the context of methods and systems facilitating the management of reactor transient conditions associated with reactors, embodiments of the present disclosure are not limited to use only in this context.
In general, the method disclosed herein may be performed by one or more computing devices. For example, in some embodiments, the method may be performed by a server computer in communication with one or more client devices over a communication network such as, for example, the Internet. In some other embodiments, the method may be performed by one or more of at least one server computer, at least one client device, at least one network device, at least one sensor and at least one actuator. Examples of the one or more client devices and/or the server computer may include, a desktop computer, a laptop computer, a tablet computer, a personal digital assistant, a portable electronic device, a wearable computer, a smart phone, an Internet of Things (IoT) device, a smart electrical appliance, a video game console, a rack server, a super-computer, a mainframe computer, mini-computer, micro-computer, a storage server, an application server (e.g. a mail server, a web server, a real-time communication server, an FTP server, a virtual server, a proxy server, a DNS server etc.), a quantum computer, and so on. Further, one or more client devices and/or the server computer may be configured for executing a software application such as, for example, but not limited to, an operating system (e.g. Windows, Mac OS, Unix, Linux, Android, etc.) in order to provide a user interface (e.g. GUI, touch-screen based interface, voice based interface, gesture based interface etc.) for use by the one or more users and/or a network interface for communicating with other devices over a communication network. Accordingly, the server computer may include a processing device configured for performing data processing tasks such as, for example, but not limited to, analyzing, identifying, determining, generating, transforming, calculating, comparing, computing, compressing, decompressing, encrypting, decrypting, scrambling, splitting, merging, interpolating, extrapolating, redacting, anonymizing, encoding and decoding. Further, the server computer may include a communication device configured for communicating with one or more external devices. The one or more external devices may include, for example, but are not limited to, a client device, a third party database, public database, a private database and so on. Further, the communication device may be configured for communicating with the one or more external devices over one or more communication channels. Further, the one or more communication channels may include a wireless communication channel and/or a wired communication channel. Accordingly, the communication device may be configured for performing one or more of transmitting and receiving of information in electronic form. Further, the server computer may include a storage device configured for performing data storage and/or data retrieval operations. In general, the storage device may be configured for providing reliable storage of digital information. Accordingly, in some embodiments, the storage device may be based on technologies such as, but not limited to, data compression, data backup, data redundancy, deduplication, error correction, data finger-printing, role based access control, and so on.
Further, one or more steps of the method disclosed herein may be initiated, maintained, controlled and/or terminated based on a control input received from one or more devices operated by one or more users such as, for example, but not limited to, an end user, an admin, a service provider, a service consumer, an agent, a broker and a representative thereof. Further, the user as defined herein may refer to a human, an animal or an artificially intelligent being in any state of existence, unless stated otherwise, elsewhere in the present disclosure. Further, in some embodiments, the one or more users may be required to successfully perform authentication in order for the control input to be effective. In general, a user of the one or more users may perform authentication based on the possession of a secret human readable secret data (e.g. username, password, passphrase, PIN, secret question, secret answer etc.) and/or possession of a machine readable secret data (e.g. encryption key, decryption key, bar codes, etc.) and/or or possession of one or more embodied characteristics unique to the user (e.g. biometric variables such as, but not limited to, fingerprint, palm-print, voice characteristics, behavioral characteristics, facial features, iris pattern, heart rate variability, evoked potentials, brain waves, and so on) and/or possession of a unique device (e.g. a device with a unique physical and/or chemical and/or biological characteristic, a hardware device with a unique serial number, a network device with a unique IP/MAC address, a telephone with a unique phone number, a smartcard with an authentication token stored thereupon, etc.). Accordingly, the one or more steps of the method may include communicating (e.g. transmitting and/or receiving) with one or more sensor devices and/or one or more actuators in order to perform authentication. For example, the one or more steps may include receiving, using the communication device, the secret human readable data from an input device such as, for example, a keyboard, a keypad, a touch-screen, a microphone, a camera and so on. Likewise, the one or more steps may include receiving, using the communication device, the one or more embodied characteristics from one or more biometric sensors.
Further, one or more steps of the method may be automatically initiated, maintained and/or terminated based on one or more predefined conditions. In an instance, the one or more predefined conditions may be based on one or more contextual variables. In general, the one or more contextual variables may represent a condition relevant to the performance of the one or more steps of the method. The one or more contextual variables may include, for example, but are not limited to, location, time, identity of a user associated with a device (e.g. the server computer, a client device etc.) corresponding to the performance of the one or more steps, environmental variables (e.g. temperature, humidity, pressure, wind speed, lighting, sound, etc.) associated with a device corresponding to the performance of the one or more steps, physical state and/or physiological state and/or psychological state of the user, physical state (e.g. motion, direction of motion, orientation, speed, velocity, acceleration, trajectory, etc.) of the device corresponding to the performance of the one or more steps and/or semantic content of data associated with the one or more users. Accordingly, the one or more steps may include communicating with one or more sensors and/or one or more actuators associated with the one or more contextual variables. For example, the one or more sensors may include, but are not limited to, a timing device (e.g. a real-time clock), a location sensor (e.g. a GPS receiver, a GLONASS receiver, an indoor location sensor etc.), a biometric sensor (e.g. a fingerprint sensor), an environmental variable sensor (e.g. temperature sensor, humidity sensor, pressure sensor, etc.) and a device state sensor (e.g. a power sensor, a voltage/current sensor, a switch-state sensor, a usage sensor, etc. associated with the device corresponding to performance of the one or more steps).
Further, the one or more steps of the method may be performed one or more number of times. Additionally, the one or more steps may be performed in any order other than as exemplarily disclosed herein, unless explicitly stated otherwise, elsewhere in the present disclosure. Further, two or more steps of the one or more steps may, in some embodiments, be simultaneously performed, at least in part. Further, in some embodiments, there may be one or more time gaps between performance of any two steps of the one or more steps.
Further, in some embodiments, the one or more predefined conditions may be specified by the one or more users. Accordingly, the one or more steps may include receiving, using the communication device, the one or more predefined conditions from one or more and devices operated by the one or more users. Further, the one or more predefined conditions may be stored in the storage device. Alternatively, and/or additionally, in some embodiments, the one or more predefined conditions may be automatically determined, using the processing device, based on historical data corresponding to performance of the one or more steps. For example, the historical data may be collected, using the storage device, from a plurality of instances of performance of the method. Such historical data may include performance actions (e.g. initiating, maintaining, interrupting, terminating, etc.) of the one or more steps and/or the one or more contextual variables associated therewith. Further, machine learning may be performed on the historical data in order to determine the one or more predefined conditions. For instance, machine learning on the historical data may determine a correlation between one or more contextual variables and performance of the one or more steps of the method. Accordingly, the one or more predefined conditions may be generated, using the processing device, based on the correlation.
Further, one or more steps of the method may be performed at one or more spatial locations. For instance, the method may be performed by a plurality of devices interconnected through a communication network. Accordingly, in an example, one or more steps of the method may be performed by a server computer. Similarly, one or more steps of the method may be performed by a client computer. Likewise, one or more steps of the method may be performed by an intermediate entity such as, for example, a proxy server. For instance, one or more steps of the method may be performed in a distributed fashion across the plurality of devices in order to meet one or more objectives. For example, one objective may be to provide load balancing between two or more devices. Another objective may be to restrict a location of one or more of an input data, an output data and any intermediate data there between corresponding to one or more steps of the method. For example, in a client-server environment, sensitive data corresponding to a user may not be allowed to be transmitted to the server computer. Accordingly, one or more steps of the method operating on the sensitive data and/or a derivative thereof may be performed at the client device.
Overview:
1. The Purpose of RT-EVALS
The present disclosure may describe methods, systems, devices and apparatuses for facilitating the management of events associated with the power plants. Further, the system may analyze the key data as it is recorded by the plant computer, in real-time, for a commercial nuclear power plant following any dynamic operating transient, such as a reactor trip from full power. The goal of the analysis is to analyze the plant data as it is recorded to identify and confirm, on a real-time basis, if any condition develops that may challenge the reactor in any way and provide suggestions for remedial actions, again on a real-time basis. Further, the system may also be used to analyze transient plant conditions that could arise when the plant is in a shutdown condition for refueling or maintenance.
The present disclosure may describe RT-EVALS (Real-Time Evaluations). Further, the RT-EVALS may analyze key plant data (from the plant computer), in real-time, for a commercial nuclear power plant following any dynamic operating transient, such as a reactor trip from full power. Further, the evaluations of the plant response may be communicated to designated plant personnel outside of the main control room thereby providing a common, informed understanding supported by multiple levels of confirmatory information. Rapid distribution of this information to designated personnel simultaneously minimizes confusion and maximizes the understanding of the ongoing plant response. In addition, these evaluations are combined with recommendations of actions to be taken if needed. In a more general sense, RT-EVALS can also be used to monitor the system performance when the reactor has been shut down, and perhaps depressurized, for maintenance and/or refueling purposes where the available plant computer monitoring instrumentation may be reduced. The RT-EVALS assessments are accomplished through continuing analysis of the developing plant information/data with the essential feature being the development of confirmatory information that can be assembled through assessments of other independent plant measurements. Equally important, the RT-EVALS results may be displayed/observed on cell phones or computer tablets that are authorized for plant management and operating personnel.
Further, the RT-EVALS assessments of the plant responses, combined with the depth of confirmation from independent system data, directly assist the plant management and operating personnel in several ways during any plant upset conditions. Further, a common understanding of the ongoing plant responses which minimizes (or eliminates) confusion associated with the understanding/interpretation of individual system behaviors thereby reducing the need for extensive, repeated inter-personnel communications regarding the status of individual systems and/or components. Further, the dynamic interpretation of the developing individual measurements and the application of the resulting insights to central concerns during the ongoing event is an essential feature. RT-EVALS is applicable to Pressurized Water Reactor (PWR) designs. Further, the system may provide a confirmation for an indication that a Reactor Coolant System (RCS) pressure boundary failure may have caused the plant transient or has been compromised as a result of the plant transient. Further, the system may provide a confirmation for an indication that a steam void may be formed within the RCS. Further, the system may provide confirmation of the development of an RCS steam void to a size that may challenge sustained cooling for the reactor core, such as an uncovering of part of the reactor core. Further, the system may provide a confirmation for an indication that the containment boundary has been, or may be compromised as a result of the evolving plant transient with the possible consequence that radioactive fission products may be released from the RCS and containment. Further, the RT-EVALS may continually access the measurements of key instrumentation from the plant computer and searches for (1) any challenge to the designed performance and (2) confirmatory indications from independent measurements to support decision-making related to the above listed central questions. Confirmation of indicated behaviors, or lack thereof, is vital information for plant management personnel and those staffing the Technical Support Center (TSC). Further, the confirmation of the principal concerns for the plant response, including sustained core cooling and heat removal, as well as containment integrity are examined by the RT-EVALS through several paths using multiple layers of the plant measurements, some of these use current measurements in innovative ways to assess the status of the core and the RCS. Outputs of these confirmatory investigations are conveyed, along with the relevant data, to the individuals monitoring RT-EVALS to establish a uniform understanding of the ongoing response combined with the status of confirmatory information for each of the above central questions. Confirmation of suspected behavior is essential to addressing possible developing condition(s) that could present challenges to adequate core cooling and/or the integrities of the RCS and containment pressure boundaries.
2. Structure of the RT-EVALS
3. How RT-EVALS is Used
As noted previously, this “real-time” event/accident RT-EVALS is not intended to be used in the main control room, which already has well developed operating and emergency operating procedures. Rather, this real-time approach is a tool that can perform a composite system analysis along with a representation for the depth of confirmation in real-time and transfer this to computer tablets and/or cell phones to uniformly communicate the results of combined reactor system analyses to management and operating personnel who are outside of the main control room. These are personnel that have a need to know and understand the evolving plant behavior and the depth of confirmation obtained from the measurements. If there is something that needs to be communicated to the main control room, these persons are the ones to communicate that information along with the support of the real-time analysis. As noted above, the principal purpose is to provide a common understanding, in real-time with confirmation, of the ongoing plant response, through comparisons of the key plant measurements, in terms of the four central questions listed above. Specifically, are there any changes, or updates to the plant status or any measurements/indications related to the status, that call attention to possible long term challenges to the reactor core, RCS or containment integrities? Confirmation, or lack thereof, of such indications would be communicated by those monitoring the plant response through RT-EVALS. As noted above, any need to communicate this information to the main control room would be handled by the TSC personnel.
When structuring this real-time approach, it is essential that the lessons learned from plant operational transients, as well as those learned from nuclear power plant accidents like the Three Mile Island Unit 2 (TMI-2) and the Fukushima core damage accidents are specifically included and evaluated, where relevant. The TMI-2 reactor was a PWR with a large dry containment and the evolving event/accident provides a convenient example to highlight the importance of developing and supplying confirmatory information to plant personnel to eliminate/minimize confusion when somewhat surprising conditions are encountered. This event/accident was initiated by a Loss of Feedwater (LOF) event which caused the turbine to be isolated and the reactor scrammed. These necessary automatic actions caused a short term pressurization of the Reactor Coolant System (RCS) such that Power Operated Relief Valve (PORV) on the pressurizer opened, as designed, to enable venting of high-pressure steam for a short interval. However, contrary to the design, the PORV remained stuck in the open position when the RCS pressure decreased below the valve setpoint and this led to a sustained loss of coolant from the RCS into the containment.
The outer surface of the tailpipe (TP) connecting the PORV to the Reactor Coolant Drain Tank (RCDT) in the containment (for Westinghouse PWR designs this tank is designated the Pressurizer Relief Tank (PRT)) was instrumented with surface thermocouples to indicate/detect if there had been, or was a continuing steam, or steam-water mixture discharge through the tailpipe. However, the control room operators had not been given any specific guidance regarding the expected temperature readings if either a short term, or a continuing discharge were to occur. Considering the critical flow of a compressible fluid like steam or a steam-water mixture through the valve (Henry and Fauske, 1971) and the downstream expansion into the tailpipe, a continuing steam discharge through the PORV and into the tailpipe would result in a range of temperatures (approximately 212° F./100 C). Conversely, a sustained discharge of a flashing, steam-water mixture could produce higher temperatures (about 300° F./149 C). Had these temperature ranges been provided to the control room operators or the plant staff, they would have understood the meaning of the measured valve of 285 F. Unfortunately, since the measured water temperatures circulating in the RCS coolant loops was about 550° F., the operators assumed that the highest temperature reading of 285° F. was representative of the tailpipe experiencing a cool down following opening and closing of the valve. Because of this misinterpretation/confusion, the steam-water discharge continued for two hours and twenty-two minutes, and it led to the uncovering and overheating of the reactor core and eventually destruction of the nuclear fuel pins.
Pursuing this further, the RCDT pressure was increasing rapidly due the sustained discharge from the PORV, but the readout for this measurement was in an instrument cabinet in the back of the control room and not easily accessible to the operators, nor were they told to check this readout. However, if they had been instructed to look at these RCDT measurements, or if the pressurization had been conveyed to them in some manner, it would have confirmed that the relief valve was stuck open. If their colleagues outside the control room would have had an analytical tool that continually analyzed and compared the essential measurements, this condition would have been discovered early in the accident. With this understanding, the control room operators could have closed the block valve that was in series with the PORV and stopped the leak early in the event with no damage to the reactor core. (Since the operators were attempting to figure this out on their own, the stuck open valve was not discovered until much later into the accident).
Digging deeper into the core measurements, a relevant example of a misinterpreted measurement from the TMI-2 accident (
Each of these illustrates the most insidious consequence of accident situations: confusion generated by (a) uncertainties in the interpretation of individual instruments, (b) the inability to rapidly check and confirm the available information through the response of other independent instruments and (c) the inability to quickly have realistic forward-looking projections based on the current plant status. Further, the system focuses on the spectrum of responses that could be generated by the evolving event/accident. Further, the system may focus on the status of the confirmatory information that is available. Furthermore, the system may provide the summarized confirmatory information in a timely manner to support the necessary decision-making process. Further, the system may facilitate the formulation of realistic, short turn around, near term projections of the event/accident based on the confirmed plant status if certain actions may be or may be not implemented.
Consequently, RT-EVALS is a real-time analytical tool that drives at the heart of possible areas where confusion could develop and prevents the confusion from occurring by communicating the extent of confirmation. This approach maximizes the use of available time and resources by always searching for confirmatory information and distributing this information promptly to those involved with plant management and the Technical Support Center (TSC). If there is information that needs to be communicated to the main control room, it should be by these individuals after they have reviewed the confirmatory analyses. Consequently, this tool provides a real-time evaluation along with a firm basis of using the currently available plant information to determine the event/accident behavior and then recommend corrective actions when needed.
4. The Information to be Supplied by the Plant Computer
Table 1 identifies some of the most important plant measurements to be examined by the engineering modules in assessing the individual PWR plant responses to possible transients, such as a reactor scram. Using the measured plant data, the RT-EVALS determines whether any of the measured information is trending outside the boundaries for the response to the initiating transient. When the identified condition(s) is(are) within the expected boundaries, the confirmatory measurements are evaluated with trends being noted but actions are not conveyed. However, if one or more measurements are/is trending toward, or is outside of the expected boundaries, the condition is identified and examined to see if other independent measurements confirm this observation. (An example is the SRM signal, shown in
Example of the Data Needed from a PWR Plant Computer
5. The Plant Information and Plant Computer Instrumentation Used by the Engineering Modules
Each engineering module has specific inputs that are needed from the plant design information file and selected information from the plant computer. The major inputs influencing the individual calculations within each module are listed below.
5.1 Reactor Core Module
Plant Design Information
Plant Computer Measurements
5.2 Reactor Coolant System (RCS) Module
Plant Design Information
Plant Computer Measurements
5.3 Steam Generator (SG) Module
Plant Design Information
Plant Computer Measurements
5.4 Pressurizer (PZR) Module
Plant Design Information
Plant Computer Measurements
5.5 Containment Module
Plant Design Information
Plant Computer Measurements
6. The Role of Each Engineering Module
6.1 Reactor Core Module
This engineering module has several roles depending on the event and the possible event progression toward any severe (core damage) condition. For all event evaluations, this module continually assesses three major plant features.
If the core fission reaction has been shut down and if the SRM measurements demonstrate a monotonic decrease consistent with the decay curve from full power, the core is behaving as expected following a reactor trip and the core is covered and cooled by water. When this is the case, the only further analyses that would be performed by this module would be a survey of the measurements reported by the RVLIS and the core exit thermocouple(s). If these were to indicate values different than a core region full of water and core exit temperatures that are consistent with, or less than, the saturation temperature corresponding to the RCS pressure, then the evaluation module would examine the results from the other engineering modules to see if the RCS shows any indication of a steam void forming and whether the containment has observed steam addition to the building atmosphere.
However, there are two additional situations that may require further analyses should they occur. One of these involves an event response in which the plant instrumentation shows that some, or all control rods are not fully inserted. Moreover, if the SRM does not show the normal decay of the neutron flux within the first few minutes, then there is a possibility that there is an Anticipated Transient Without Scram (ATWS). This would trigger a specific set of Emergency Operating Procedures (EOPs) by the control room operators. These procedures dictate the control room operator actions, and the RT-EVALS should not be used until the core has demonstrated a shutdown behavior.
A second situation is like that which occurred during the TMI-2 accident where the reactor core nuclear fission was shown to be shut down by the fully inserted control rods and the SRM signal demonstrated the rapid decrease associated with a control rod insertion (reactor scram). About 30 minutes after the reactor scram, the SRM signal changed and began to increase. However, this was not caused by a return to fission power in the core! It was the result of a loss of water in the core region (formation of a steam void) which caused less absorption of the strong gamma rays and the delayed neutrons produced in the core. Using this lesson learned, if an event should result in a SRM measurement that eventually experiences a reversal in the monotonic decreasing signal, the Core Module evaluates this as a loss of water inventory from the core region and calculates the extent of a steam void that is forming in the core.
If there is confirmation of a steam void forming within the core and the RCS, the core exit temperature measurement(s) is/are examined to determine if there is an indication of inadequate core cooling. Should an event progress to the point where the steam void would be sufficiently large to challenge core cooling, the Core Module estimates the time before the maximum core temperature could increase to the point where the Zircaloy cladding for the uranium fuel pellets could begin rapid, exothermic oxidation that could cause major core damage. For example, the rapidly rising SRM signal in
Should an event evolve into inadequate core cooling and core overheating that achieves either measured or calculated core exit temperatures that exceed 1300 F (˜700 C), RT-EVALS will begin to assess the possible formation of hydrogen as a result of cladding oxidation in the steam environment that becomes significant at these temperatures.
Fundamental calculations (EPRI, 1992) show that essentially the upper half of the core must be uncovered before the core exit temperatures reach the above levels. If this would occur, the exothermic oxidation reaction between the high temperature Zircaloy cladding and steam would generate a thermal excursion and eventually the local chemical energy release would eventually exceed the decay heat generated by the fuel. This rapid chemical reaction would continue increasing at greater rates until the fuel pin cladding would melt, liquefy the fuel pellets, drain into the lower core region and freeze. At this point, the relocated frozen core material would essentially block the steam flow through the core and begin to freeze in the lower, cooler core regions. As this frozen material collects it would greatly reduce the chemical reaction. This order of magnitude change in the core geometry (see
Should an accident progress to this point, Henry (2019) has shown that the order of magnitude reduction in the oxidation for all three in-reactor Phebus experiments is consistent with oxidation of the remaining metals in the core from the core debris upper surface. Natural circulation of steam downward to the debris upper surface in conjunction with the upward flow of the hydrogen generated by the oxidation as illustrated by the Countercurrent Flow Late Stage model (CCFLS). Consulting
6.2 Reactor Coolant System (RCS) Module
There are several RCS measurements that can be used to evaluate the event transient as it relates to the formation of a steam void in the RCS as well as long term cooling of the reactor core. These include the RCS system pressure, the cold leg and hot temperatures for the individual coolant loops, the mass flow rate measurements for each loop with a Reactor Coolant Pump (RCP) (also called Main Coolant Pumps (MCPs) for some PWR designs) operating along with a mass balance constructed from the water sources of addition (injection) and removal (letdown) from the RCS. If there is a loss of coolant from the RCS, there would be a decrease in the system pressure and in addition, the hot leg temperatures would be essentially the saturation temperature corresponding to the measured RCS pressure. (This relationship is evaluated within the RCS Engineering Module using a set of equations entitled STEAM-WATER PROPS which is a fast execution routine that agrees well with Keenan et al, 1978). This is one of the ways that the RCS Module analyzes other measurements to determine whether these also indicate that a steam void could be forming in the RCS. Through these analyses that are submitted to the Evaluation Module, the RT-EVALS determines the depth of confirmation that is provided by the reactor and containment system measurements.
While the RCS pressure and coolant temperatures can indicate whether a steam void could form, these cannot be used to calculate the magnitude of the steam void. However, if the RCPs/MCPs are active in one, or more coolant loops, the individual mass flow rate measurements in each coolant loop with an operating RCS/MCP provide a means for estimating the average steam void that is undergoing forced circulation through the coolant loop as demonstrated by Henry (2011). Here again, the TMI-2 plant data demonstrates how mass flow rate measurements respond to the formation of two-phase, steam-water mixtures in the coolant loops. These evaluations can be used to either assess the steam void formations in the coolant loops with running (energized) RCPs or act as a confirmatory calculation for another such calculation in another module, such as the SRM evaluation in the Core Engineering Module.
Another assessment provided by the RCS Module is to use the rate of increase in the steam void formed in the coolant loops and the RPV to estimate the possible break size that could be responsible for the coolant loss. Such a calculation would be helpful in terms of detecting what may have been the initial cause of the event and therefore what may be the way in which the event progression could be terminated. From a mass balance, the estimated mass flow rate (WLOCA) through a LOCA can be assessed by:
WLOCA=VRCS(ρf−ρg)[(α2−α1)/(t2−t1)]+Ww,inj+Ww,accum+WwPZR−Ww,LETD−(QD−QSG)/hfg
The variables in this equation are:
Once the discharge mass flow rate is estimated, the break area (ALOCA) is evaluated assuming that the mass flow rate is limited by the two-phase critical discharge for a steam-water mixture with an average void fraction given by αAVG=(α2+α1)/2. For low and moderate void fractions, the Henry-Fauske critical flow model (Henry and Fauske, 1971) can be approximated for using the following equation:
ALOCA=WLOCA/SQRT{Cd[2(1−αAVG)ρfPRCS(1−η)]}
In the above equation, the term i is the critical pressure ratio that is defined as the ratio of the pressure in the minimum flow area or throat (Pt) of the break or open valve divided by PRCS. As indicated by the experimental data referenced in the Henry and Fauske paper, the value of η for low void fraction, moderate void fraction mixtures have a value of about 0.8, with the discharge coefficient through a valve or an orifice like break also being about 0.8.
Recalling the confusion over the progression of the TMI-2 accident, such calculations would have shown early in the accident (within the first 30 minutes) that a breach in the RCS/PZR pressure boundary, that was approximately of the size of a stuck open PORV, was responsible for the RCS coolant loss. Given the event, there was only one valve that would have been opened by the initiating event (the pressurizer PORV) and this could have been eliminated by closing the block valve in series with the PORV. If the block valve would have been closed at 30 minutes into the accident, the event would have been terminated and the reactor core would not have been damaged.
6.3 Steam Generator (SG) Module
Each steam generator has thousands of tubes through which the high-temperature RCS coolant from the core is circulated and energy is transferred to the secondary water outside the tubes. For at-power conditions, this energy transfer results in a large steam generation rate that is the steam being ducted to the turbine—generator set where electricity is produced. Since these SGs are the principal means of extracting heat from the RCS and the thousands of SG tubes are part of the RCS pressure boundary, this engineering module is of key importance to the evaluation of the plant response.
One possible event/accident sequence to be considered by the Evaluation Module is a High Energy Line Break (HELB) since this could potentially occur either inside or outside of the containment. Should a break occur outside of the containment building as occurred in the feedwater piping rupture at the Surry nuclear plant (1989), this could present an immediate hazard to plant personnel. If such an event occurred, an important signal is the containment pressure since there would be pressurization of the building if a break were to occur inside of the leak-tight containment building, but no pressure increase would occur if the break was external to the containment. As a result, the Evaluation Module checks the Containment Engineering Module to determine whether there is any containment building pressurization and also the SG Engineering Module to see if there is any depressurization detected in only one SG. If so, the RT-EVALS would begin with an indication of a HELB.
Industry experience has also shown that the steam generator tubes can be subjected to long term erosion and corrosion such that individual tube walls have experienced thinning and single tubes have ruptured (Steam Generator Tube Rupture—SGTR) when a plant has been in operation or is in the process of being brought to operating conditions as was the case for the Doel-2 plant on Jun. 25, 1979 (Stubbe et al, 1984 and NRC, 1979). Because the steam generator tubes have a diameter of the order of 1 cm, the SG depressurization rate would be very slow, so there is an extended time to address the event/accident condition.
Consequently, a SGTR is a break location where the RCS pressure boundary would fail with the RCS coolant being discharged through the breach into the steam generator secondary side. As a result, the discharge would cause the secondary side water level, pressure, and radiation level to increase in that SG, with the other SGs remaining unaffected. The increasing water, pressure and radiation levels would trip the reactor, and a transition to long term cooling of the reactor core would begin. With the possibility of a Loss of Coolant Accident (LOCA) combined with the role that the SGs could have in establishing the RCS depressurization and long term cooling, the SG Engineering Module is naturally an essential module to assess the transient performance of all of the SGs in the design, including the unit with the SGTR and transmit the results of these evaluations to the Evaluation Module.
Depending on the plant design, there could be two, three or four SGs, and if the Russian VVER PWR designs are included, there could be six SGs monitored by the SG Engineering Module. The important information to be supplied to the SG Module are the individual feed water flow rates, water levels, pressures, radioactivity levels of the RCS coolant and the status of the Main Steam Isolation Valves (MSIVs). If a reactor trip transient is initiated, the auxiliary feedwater flow rates to the individual generators must also be monitored. These measurements encapsulate the operation of each of the SGs.
6.4 Pressurizer (PZR) Module
In the range of interest, water expands (becomes less dense) with increasing temperature and is nearly incompressible. Consequently, any volume that is completely water-filled would be subjected to a rapid pressurization if energy were added to the volume. To cushion such pressure transients, PWR designs have a separate, vertically oriented tank (the pressurizer) connected directly to one of the hot leg pipes of the water-filled (subcooled water) RCS. To effectively cushion both pressurizations and depressurizations, this tank is approximately half-filled with water and half-filled with steam. With steam being far more compressible than water, the steam volume cushions the RCS and promotes well-controlled, steady-state operation. Further protection against the possibility of high RCS pressures during rapid transients, such as a reactor trip, is provided by a cold water spray into the top of the PZR (see
Consequently, this sizable storage of cold water combined with the potential for valve leakage as well as the instrumentation associated with the tailpipe and the PRT/RCDT, the Pressurizer (PZR) Engineering module provides another essential component to be included in the RT-EVALS. The measurements of interest for this component are the water level in the pressurizer tank, the valve stem positions for the relied valves where available, the tailpipe temperatures as well as the pressure and temperatures for the PRT/RCDT containment collection tank. (With the direct connection between the RCS and the pressurizer, these can be considered to be at the same pressure with the only difference being the static head of water in the PZR during the transient response.) With their individual responses being well characterized, the tailpipe and collection tank responses are considered in submodules that are called by the PZR Engineering Module.
6.4.1 Pressurizer (PZR) Responses
By design, the PZR responses for anticipated operational transients involve either short term expansions or compressions of the steam volume with the pressurizer water level eventually returning to some relatively constant, measurable water level. Should the PZR lose the entire water inventory within a minute or less, the first cause to consider is a Loss of Coolant Accident (LOCA) in the RCS may be occurring as was observed in the Loss of Fluid Tests (LOFT), see for example Guntay (1990). Conversely, as observed in the TMI-2 accident, if the pressurizer approaches a level that almost fills the pressurizer vessel, it becomes a clear indication that one, or more of the pressurizer valves has/have opened and has/have not reclosed. This was the initiator for the small LOCA in the TMI-2 accident. (This behavior was also observed in LOFT Test L3-0 (Modro, et al, 1987).) If the pressurizer water level remains within the central zone of the measured height, then the event is likely not LOCA related, for example the Mannshan Station Blackout observed the PZR water level to decrease from 60% to 20% of the measured height in one hour (see Che-Hao, 2015).
As an example, consider the pressurizer water level response (see
Especially note the large fluctuations in the pressurizer water level in the early stages of the transient. These and other behaviors must be part of the system evaluation that investigates the crucial measurements to determine the nature of the developing transient. The character of the transient is suggested by analyzing the ensemble of the measurements with the assessment providing the answers to the questions: (1) what behavior characterizes the ongoing behavior, (2) what are the confirmatory observations, (3) what actions should be considered, (4) what actions are recommended and (5) how much time is available to accomplish these actions? As defined below, the PZR behavior is an essential component of this ensemble.
6.4.2 PZR Valve Discharge Rates
For most reactor trip transients, such as a loss of load, loss of off-site power, etc. the transient occurs sufficiently rapidly that the RCS pressurization causes the opening of the PORV and perhaps the safety valves. The TMI-2 Electromatic Relief Valve (ERV), which was the PORV for the TMI-2 plant, had an opening set point of 2255 psig (2269.7 psia/15.65 MPa) compared to the nominal operating pressure of 2200 psig/15.17 MPa (NSAC, 1980).
Eventually, the PORV and perhaps one or more safety valves would likely open. With the magnitude of the pressure difference and the RCDT (which is about the containment pressure), the flow through these valves would be limited by the maximum compressible flow of the fluid being discharged. For the discharge of single-phase steam flow through one, or more of these valves, the mass flow rate discharged (Wst) can be determined by:
Wst=CdAvSQRT[{2γ/(γ−1)P0 ρ0(ηpow(2/γ))[1−pow((γ−1)/γ)]}]
where η={2/(γ+1)}pow[γ/(γ−1)]=Pt/P0
(In these equations, the use of “pow” indicates that the following bracketed term is the power for the term immediately before the “pow” and SQRT is the square root of the term in brackets.)
In this expression, Av is minimum flow area through the valve(s), Cd is the empirical compressible flow discharge coefficient for the valve, P0 is the PZR pressure, ρ0 is the steam density in the pressurizer, γ is the isentropic coefficient for steam at P0 and η is the ratio of the throat pressure (Pt) divided by P0. At a pressure of 15.56 MPa saturated steam has a density of 104.2 kg/m3 and γ=1.25. Initially assuming a value of Cd=1, the values given above result in a steam mass flow rate of 26,576 kg/m2/s or 142,495 lbm/hr. The characterization for the TMI-2 ERV in the NSAC-80-1 report gives a mass flow rate of “approximately 100,000 lbm/hr of saturated steam”, which corresponds to a mass flow rate of 12.6 kg/s and a value of Cd=0.7. This value is close to that which would be expected for such systems (Henry and Fauske, 1971). With this mass flow rate, the volumetric flow rate leaving the PZR would be 0.121 m3/s.
Once the valve opens to vent steam, the water level would increase as water flows into the pressurizer from the hot leg to replace the steam volume vented. When the RCS approaches being full of water, which is the applicable configuration with the water level continuing to increase. Two facets of this process could influence the transient PZR water level measurement. The first is that the reduction in pressure may cause some flashing of water to steam within the bulk of the PZR water inventory. This has a minor influence on the static head of the water inventory, but the formation of steam bubbles within the liquid water would cause an expansion of the water volume where this occurs with the consequence of some level swell of the upper water surface.
As the water level approaches the top of the pressurizer vessel, this level swell would cause a two-phase, steam-water mixture to enter the open valve instead of single-phase saturated steam. A discharge of a two-phase, steam-water mixture from the pressurizer has an important influence in terms of (i) the mass flow rate increases, and (ii) the volumetric flow rate decreases. Experimental data (Ginsberg, et al, 1977, Grolmes et al, 1985, Fletcher and Denham, 1993 and Henry et al, 2015) have shown that the two-phase water level increases (typically described as level swell) with the rate of steam formation beneath the collapsed water level. Knowledge of the steam discharge mass flow rate and the dimensions of the PZR vessel provide the necessary information to calculate the level swell. This is a simple calculation that can be evaluated in real-time by RT-EVALS during the evolving event.
For the TMI-2 event/accident, this would generate an average void fraction of about 7% over the 400 inches of measured height. Consequently, multiplying the measurement height and 0.07 gives a level swell of about 28 inches. (It is important to note that the level swell does not change the collapsed water level that is measured by the static pressure devices such as differential pressure transducers.)
6.4.3 PZR Tailpipe Temperature Measurements
With the continued steam and/or steam-water discharges indicated by
If there is a critical discharge (flow) of steam or a steam-water mixture through a stuck open valve, the discharge into the much larger tailpipe can be assessed as a freely expanding jet that eventually occupies the entire cross-sectional area of the tailpipe. The discharge mass flow rate of steam (Wst) or a two-phase mixture is the product of the fluid density at the valve throat (pt), the effective cross-sectional area of the valve (Cd At) and the throat velocity (Ut) which is the sonic velocity of the steam or a two-phase mixture:
Wd=ρt(CdAt)Ut
The free expansion of the critical flow jet downstream of the throat as the lower pressure in the expansion zone (Pe) accelerates the critical flow jet to the velocity Ue as defined by the momentum equation
Pt−Pe=Cd(Wd/At)(Ue−Ut)
Assuming an isentropic expansion for the steam or two-phase mixture, then the above equations can be solved by iteration to determine the consistent values of Pe and Ue that are needed for the discharge mass flow rate and the piping cross-sectional area (Ae). The pressure Pe determines the temperature of the expanding jet.
Of the valves that vent from the top of the PZR, the PORV has the lowest setpoint, so it would be the first, and possibly the only valve to open. Therefore, it is the most likely to become stuck in the open position. Considering this to be the case, the temperature of the expanding steam, or steam-water discharge can be quantified as being in equilibrium at the calculated pressure for the fully expanded jet. The results of the TMI-2 measured temperatures and those calculated for the stuck open PORV are listed in
Both of the values recorded at 24 minutes and 58 seconds, as well as that one hour, 20 minutes and 31 seconds, are when the PZR water level is near the top of the vessel, so a steam-water mixture was being discharged, which results in a much higher pressure for a fully expanded jet and the temperature is consequently more than 50 F higher. As listed in the table, the measured maximum values are also increased by a similar increment. Hence, the measured tailpipe temperature measurements also indicate that there is a two-phase mixture flow through the pipe and into the RCDT.
Lastly, at 2 hours, 17 minutes and 53 seconds, the measured PZR water level has decreased such that only steam could be vented from the PZR and both the measured and calculated tailpipe temperatures have also decreased nearly to the atmospheric saturation temperature. From these comparisons with different PZR water level measurements, the tailpipe temperature measurements provide confirmation that a continuous discharge was ongoing for over two hours following the initiating transient. Moreover, the measured values indicated when the discharge was steam and when it was a two-phase mixture. Therefore, this measurement provided an ongoing powerful confirmation of the event behavior, and in the RT-EVALS methodology, this would be transmitted to the Evaluation Module for its understanding of what measurements have independently confirmed behaviors. If only the control room operators had been given the temperature measurement levels to expect during such a discharge, they would have known immediately what was happening in the plant.
6.4.4 RCDT/PRT Pressure Measurements
In addition to the use of the tailpipe information,
6.5 Containment Module
All PWR designs in the western hemisphere have leak-tight, high-pressure containment buildings that encapsulate the RCS and the SGs. By law, these containment buildings are periodically pressure tested to their design basis pressure. With its role of containing any RCS coolant that could be discharged as a result of an event/accident condition, the Containment Engineering Module is also a fundamental component of the RT-EVALS evaluations.
With its role as a containment of the RCS water and steam that could be lost from the RCS, the measurements of interest for this module are any features that could cause elevated temperatures, pressures and/or radiation levels in the building and any increases in the sump water level(s) that could be detected. All containments are designed with cooling systems to remove decay heat from the building by water sprays that condense steam discharged into the building and some containments also have safety-related air coolers that are designed to remove decay heat through steam condensation. There may also be other air coolers that are designed for the smaller heat loads that are related to steady-state plant operation. All containments have decay heat removal systems that remove high temperature water from the RCS and/or containment, pass the water through heat exchangers to cool the water and then return it to the RCS and/or containment. Consequently, this RT-EVALS uses the containment pressure and temperature measurements as well as the information on the water and steam-air flow rates through the heat exchangers and coolers and the temperature differences across these heat exchangers and coolers (large and small) were available depending on the plant design. Returning to the TMI-2 accident data for example calculations that would be performed by the Containment Engineering Module, the first indication of a developing condition in the containment (aka as the reactor building in the TMI-2 documentation) is building pressurization that began about 15 minutes into the event. Steam discharge into the building started with the RCDT rupture disk failure and this caused the pressure to increase by 2 psi (0.14 bars) over an interval of about 5 minutes and then this overpressure remained nearly constant over the next 2 hours.
Given the measured building pressurization, the Containment Engineering Module uses this as one means to estimate the steam flow rate into the containment. Specifically, the estimate assumes no steam condensation during the initial pressurization as represented by the differential form of the perfect gas equation:
Wst˜[(VgconMw/(RTg)]{dP/dt−(P/Tg)dTg/dt}
In this equation, R is the Universal Gas Constant (8314 J/kg-mols/K), Tg and P are the measured gas pressure (Pa) at the beginning of the pressurization, Vgcon is the gas volume (m3) in the containment, Mw is the molecular weight of steam (18) and the rates of change for the pressure and temperature are dP/dt and dTg/dt respectively. (It is noted that the first term (dP/dt) is about four times larger than the second term so if the temperature measurements are slower in responding than the pressure measurements, the estimate will be larger and therefore more conservative.) Since the condensation rate would likely increase directly with the steam partial pressure, it is preferable that the estimate for the pressurization rate be taken as a current value minus the initial pressure once the containment pressurization begins.
Examining the data from the TMI-2 accident, the initial pressurization rate at 15 minutes into the event is approximately (0.0069 psi/s or 47.6 Pa/s) with the measured air temperature experiencing about an 11° C. increase during the five-minute pressure increase (NSAC 1980). With a containment gas volume of 56,600 m3 and a gas temperature of ˜300 K, the estimated steam discharge rate into the containment is about 14.5 kg/s. With a heat of vaporization for water of 2.366×10E6 J/kg, this steam discharge corresponds to an energy addition rate of 34 MW. This is comparable to the decay heat generated in the core and is a clear indication that there is a developing situation with considerable steam discharge into the containment that needs to be continually evaluated by the RT-EVALS. This information would be passed to the Evaluation Module to compare with the results of other engineering calculations provided by the other engineering modules.
Once the containment pressure increased by 2.5 psi (0.17 bars) (NSAC, 1980) and the air temperature increased to about 135 F (57 C), this condition remained nearly constant for two hours. This was because the five containment air coolers were all operating at their highest design flow rate of 42,590 scfm (20.1 m3/s) at atmospheric pressure (NSAC, 1980). With the design volumetric flow rate and an air density of about 1.19 kg/m3 at moderate pressures, this would be a mass flow rate of 23.9 kg/s for each cooler.
The energy removal rate via steam condensation in the containment air coolers (aka fan coolers) can be estimated by assuming that the air entered each air/fan cooler with a humidity of 100% and that the steam was completely condensed as it passed through the cooler. This estimate of the energy removal rate for a single air/fan cooler can be quantified by:
dE/dt=QFANFLOW1*ρst*hfg
where ρst and hfg are the saturated steam density and latent heat of vaporization at the steam partial pressure respectively. Assuming that the extent of pressure increase is steam partial pressure (0.17 bars), the steam density would be approximately 0.114 kg/m3 and the latent heat of vaporization would be the value given above (Keenan et al, 1978). The estimate for total energy removal rate for the containment is the product of the above equation and the number of fan coolers operating (NFANS).
dEtot/dt=NFANS*dE/dt
Using the above expressions, the energy removal rate for each air cooler is estimated at 5.42 MW and the total removed by five fan coolers is calculated to be 27.1 MW, which is also comparable to the decay heat generated in the core (˜30 MW). As noted with the other estimate of the steam discharge to the containment, a heat load approaching, or exceeding the decay heat is a clear indication of a LOCA in either the RCS or PZR or possibly a HELB within the containment. If the steam discharge continues for tens of minutes, it is clearly a LOCA.
With these estimates of the pressurization rate and the energy removal rate from the air coolers, the Containment Engineering Module would have detected and concluded that the ongoing event has the characteristic of a LOCA. Both of these would be reported to the Evaluation Module where these could serve as an independent confirmatory observation for the analyses previously performed by the other engineering modules.
In this regard, it is important to note that the TMI-2 containment pressurization did not start until the Reactor Coolant Drain Tank (RCDT) rupture disk burst at about 15 minutes into the event. Therefore, the containment contribution to the Evaluation Module may occur somewhat later than the analyses provided by the Core, RCS and PZR Engineering modules. Nevertheless, the internal engineering evaluations of the module either confirm or add to the depth of confirmation of diagnosis from other engineering modules.
Both containment estimates are comparable to the decay heat and this is more than sufficient to detect whether there is a developing situation that could challenge the containment integrity. Moreover, it is also clear that the starting of either the safety grade air coolers and/or the design basis containment sprays would be sufficient to reduce or limit the containment pressure. It is important to note that these two estimates occur at two different time intervals, the pressurization rate is only meaningful for a few hundred seconds and then containment pressure became nearly constant. Nevertheless, the Containment Engineering Module stores both histories so the Evaluation Module can look backward in the event history to potentially understand the detailed history of the event development if necessary.
6.6 The Evaluation Module
This module gathers the results from the engineering calculations and evaluations provided by each of the five Engineering Modules. With these, the Evaluation Module then searches for a possible fit for the event progression as well as confirmation from independent data sources that a type of accident condition is developing.
There are four types of plant initial conditions that need to be considered: (i) at-power (ii) scrammed from an at-power-state, (iii) shutdown with Reactor Pressure Vessel (RPV) head in place and bolted down consistent with the design basis and (iv) a shutdown state for refueling/maintenance with the RPV head either removed or not bolted down. Fundamentally, there are only two event/accident types that relate to possible challenges to the reactor core, (1) a Loss of Coolant Accident (LOCA) and (2) a loss of adequate heat removal. The RT-EVALS methodology will use the plant measurements, in real-time, as described above to decipher whether a LOCA is in the RCS, the PZR, one of the SGs or an Interfacing System LOCA (ISLOCA) and search for possible corrective actions. Moreover, if the accident type is a loss of adequate heat removal, RT-EVALS will rapidly search in real-time for the alternate ways to achieve the needed heat removal capability and continually provide forward-looking projections regarding the time available before core damage could be anticipated, including flexible strategies (FLEX) where needed capabilities could be transported to the plant site. There are other event types that influence the balance of plant, such as a High Energy Line Break (HELB) and these are also considered by the Evaluation Module.
Also, there are accident event types that could involve an Anticipated Transient Without Scram (ATWS) that are addressed by the control room EOPs. These are part of the operator training on plant-specific simulators and are not specifically examined in the RT-EVALS. For the remaining accident types, the reactor would be scrammed and a possible LOCA would have the potential to be the most rapidly developing. Therefore, the Evaluation Module begins by examining the SG, RCS, Core PZR and Containment Engineering modules for any signs of a HELB, one of the possible LOCAs mentioned above or a loss of adequate heat removal. Consequently, the first quire is whether there is evidence of a HELB since this could possibly be an immediate hazard to plant personnel.
Information Transmitted by the Steam Generator Module
Information Transmitted by the Core Module
Information Transmitted by the Pressurizer Module
Information Transmitted by the RCS Module
Information Transmitted by the Containment Module
Projection of Near-Term Behavior
As noted above, the principal objective of the RT-EVALS is to analyze the plant information from all of the major components to develop a common understanding of the developing transient on a real-time basis along with the depth of confirmation provided by independent measurements. In addition, the methodology can use the much faster than real-time internal models to provide near term projections for addressing “what if” questions related to actions that could be taken and also to provide projections related to “how long before . . . ” assessments. This could have an important function as part of the implementation of the FLEX capabilities where additional power, pumping and heat removal systems can be supplied from a remote site. For example, if those in the Technical Support Center (TSC) want to know what would be the influence of hooking up a fire water pump to the secondary side of a SG in 15 minutes and injecting a flow rate of 10 kg/s, the near term projections will take the current status, assume that the event progression will remain as it is currently progressing for 15 minutes and begin to inject this flow rate into one of the SGs and graphically illustrate on a cell phone or a computer tablet how the event progression will be changed. Most importantly, this will be calculated and displayed much faster than real-time so those in the TSC can determine if the action will accomplish the desired behavior. As noted, where appropriate these assessments use the same physical models, such as the size of a LOCA, the steam-water discharge mass flow, the temperature increase caused by decay heat for uncovered portions of the reactor core that have been used to evaluate the evolving transient, so these will produce conservative projections for the near term behavior. Moreover, if needed, the models embedded in the Engineering Modules can make projections into the core damage states of early and late phase hydrogen generation that could result from extensive overheating of the reactor fuel pins, particularly the Zircaloy cladding in a steam environment.
7. Recommended Actions and Near-Term Projections Block
This segment of the RT-EVALS methodology utilizes the comparative information and depth of confirmation provided by the Evaluation Module to recommend actions to be taken to recover from the accident conditions and/or mitigate the consequences of the developing transient. The objective of the actions is always to eventually terminate the event progression, but some actions may need to be taken in the interim, such as controlled venting of the RCS, the SGs and/or the containment, to maintain the capabilities of the reactor/containment designs. In general, these recommendations focus on making sure there is adequate water injection to the RCS and the SGs for core cooling and for decay heat removal. Furthermore, there are considerations related to depressurizing the RCS and potentially the SGs, but these are generally achieved through the EOPs. Other actions could include using the FLEX capabilities provided to the reactor site, as well as short term containment venting to keep the pressure well below the design basis level.
According to some aspects, a new, real-time methodology is disclosed. The methodology can be used to interpret the transient plant data as it is recorded to diagnose, confirm and communicate to the plant management and designated personnel whether there are developing conditions that could eventually challenge cooling of the reactor core, integrity of the RCS and/or containment boundaries.
According to further aspects, the methodology is based upon Engineering Modules that perform engineering evaluations of the transient plant data, as it is recorded, in innovative ways that, while somewhat different is completely consistent with the ongoing physical processes and what has been observed in well-documented plant transient behavior and accidents.
According to further aspects, the engineering modules have subordinate modules that analyze the response and performance of dedicated systems such as temperature measurements on the pipe connecting the PZR relief valves to the RCDT/PRT, the pressure history measured in the RCDT/PRT, water injection to the RCS, water addition to the containment sprays, localized heat removal for the containment and others.
According to further aspects, the engineering modules make use the readings of standard existing instrumentation, such as the SRM, RCS loop flow rates, containment pressure history in innovative ways that are consistent with the behavior of these instruments in the well documented Three Mile Island Unit 2 accident and other plant transients.
According to further aspects, the engineering modules can accept manual entries of observations in the plant such as “steam is being discharged into the turbine building”, “steam is being discharged into the auxiliary building”, “one of the service water pumps is disabled for maintenance”, or others.
According to further aspects, the information from the engineering modules can be supplied to the evaluation module to compare the results from each engineer module to determine the nature of the developing transient as well as the depth of confirmatory analyses regarding the transient behavior. Examples of challenging transients that could develop are: a High Energy Line Break (HELB) inside of containment, a HELB outside of containment, a Loss of Coolant Accident (LOCA) in the Reactor Coolant System, a LOCA in the pressurizer (PZR), a Loss of Feedwater (LOFW) event, an Interfacing System LOCA (V Sequence), a Steam Generator Tube Rupture (SGTR), the loss of one or more Emergency Core Cooling Systems (ECCS), Loss of Off-Site Power (LOSP), total loss of off-site and on-site AC power designated as a Station BlackOut (SBO), loss of adequate core cooling during mid-loop maintenance operations and others.
According to further aspects, the results of the Evaluation Module can be supplied to the Recommended Actions block to supply to the plant management for their decisions of whether or not to: (i) provide additional information to the main control room, (ii) provide specific help to specific parts of the plant, (iii) request the need for help from a FLEX facility, (iv) inform state and local regulatory agencies and others.
According to further aspects, the results of the Evaluation Module can be used to provide near term projections, with each recording of the plant data, regarding the progression of the event with the objective being to establish an event timeline quantifying intervals of when specific plant capabilities would be required and whether these capabilities were available on-site or would need to transport to the site.
According to further aspects, the RT-EVALS methodology is a real-time tool designed for the management and operating personnel outside of the main control room that analyzes the evolving plant information for a commercial PWR nuclear power plant following an operating transient that can be observed in real-time on cell phones or computer tablets that are authorized for plant management and operating personnel.
8. Early Phase H2 Generation
1. Background
Under accident conditions where continuous water addition to the reactor core could have been disrupted, the water inventory in the Reactor Coolant System (RCS) could maintain core cooling through water vaporization (boiling). However, without sufficient water injection, boiling would deplete the water inventory in the Reactor Pressure Vessel (RPV) and in the reactor core. Steam generated beneath the water level would increase the level somewhat (level swell) and this would act to extend the core cooling interval. Most importantly, when the water level is above the top of the core, boiling (vaporization) of the coolant will maintain the core temperatures close to the saturation temperature (boiling point) corresponding to the RCS pressure. However, without water addition to offset the vaporization, the water level would eventually decrease below the top of the reactor core. Once uncovered, the top of the reactor core would begin to overheat.
2. Uncovering of the Reactor Core
Water vaporization (boiling) due to the core decay heat generation without water addition would cause a decrease in the water level (Lw). When the core is submerged in water, it will not overheat. Should the water level decrease below the Top of Active Fuel (TAF) (Lw<the core height Lc), the water level will begin to decrease at a rate determined by the extent of decay heat generated beneath the water level. This can be estimated with the following one-dimensional continuity expression in terms of the dimensionless core height z=Lw/Lc:
−ρwAwLcdz/dt+Wadd=zQD/hfg
This approximate formulation assumes that the heat flux is uniform over the core height and that there is no significant oxidation of the Zircaloy cladding. Other terms are defined as: Aw is the cross-sectional area for water in the RPV with access to the core (core+bypass+downcomer), hfg is the latent heat of vaporization of water at the RCS pressure, Wadd is the water addition rate to the core and ρw is the water density. QD is the total decay generated in the core at a given time. El Wakil (1971) has shown that the decay heat can be represented over long time intervals following shut down of the fission reaction by:
QD//Qc0=0.095tpow(−0.26)
In this expression, Qc0 is the long-term core power level (2780 MWt for TMI-2) that characterized operation before the shutdown and t is the time since reactor scram in seconds.
To estimate the potential for core overheating, it is conservative to assume that the water entering the core is saturated at the local pressure. Additionally, the following solution assumes that the RCS pressure remains constant as the core water level changes, so the water properties remain constant. Integrating the above expression from 1 to z results in:
−ln(z)=C1Δt
C1=QD[ρwAwhfgLc]−1
Future projections for accident conditions that could lead to uncovering of the reactor core need to provide realistic representations of the transient fuel pin cladding history. To develop a perspective on the controlling processes for the rapidity of the increase in the cladding temperature, it is helpful to examine the TMI-2 core temperature increase as water vaporization depleted the core water inventory according to the above one-dimensional model. From the core design information, the active core height (Lc) is 3.66 m and the term Aw is about 14.9 m2. Furthermore, at the time that the core water level began to decrease below Top of Active Fuel (TAF) (approximately 100 minutes (6000 seconds) that corresponds to when the last Main Coolant Pump was tripped), the RCS pressure was approximately 7 MPa with a saturation temperature of 286 C (559 K), hfg is about 1.5×10E6 J/kg and the water density is 741 kg/m3 (Keenan et al, 1978). Substituting these values into the above equation and solving for the time interval for the water level to reach the lower levels of the core gives the results shown in
As shown in
dT/dt)AB=QD/[mccc]
where mc is the core mass with cc being the average specific heat of the core materials. The total mass of materials in the TMI-2 core is 129,700 kg (Henry, 2011) and for the estimates of the adiabatic rate of rise, the average specific heat can be taken to be 500 J/kg/K. Adiabatic heating of the TAF region by decay heat alone is shown in the third column of
Since the Zircaloy cladding is in a steam environment, increasing cladding temperatures increases the rate of oxidation, which is described by the following balance:
Zr+2H2O→ZrO2+2H2+ΔHR
Once the cladding wall temperature (Tw,TAF) has been estimated, the oxidation behavior for this highest temperature locale can be also be determined using the Arrhenius equation proposed by Cathcart et al (1977) that was derived from zirconium oxidation experiments taken in the temperature range up to 1580° C. (1853K). This correlation for the Zircaloy mass reacted per square meter (w in units of kg/m2) within a specified time interval (t) in seconds
is given as:
w2=294t exp{−1.654×10E8/R/Tw,TAF}
where “exp” is the natural logarithm raised to the power of the bracketed term that follows and R is the universal constant of 8314 J/k/kg-mol. The fourth column in
Once the core was approximately half uncovered, the core exit region would have reached temperatures where the rate of chemical energy release is comparable to the decay heat. As the depletion of the core water inventory continued, the energy release rate due to oxidation became the dominant energy source and by the time that the water level decreased to 40% of the active core height, the core exit temperature reached the melting temperature of the stainless steel, and/or Inconel core components which are usually the in-core instrumentation. This was the onset of rapid oxidation as well as configurational changes in the reactor core, and for a short interval, the oxidation occurred as fast as steam was generated in the lower part of the core; a condition characterized as “steam starvation”. With the exponential character of the oxidation reaction, most of it occurs during the relatively short “steam starvation” interval defined by the fuel pin temperature approaching the stainless steel/Inconel melting temperatures and the melting/liquefication of the Zircaloy cladding and the oxidic reactor fuel that results in the downward flow of molten debris.
Note that the number of the hydrogen kg moles produced equals the kg moles of water vaporized. (ΔHR is the energy (heat) released by the oxidation reaction, which is 6.84×10E6 J/kg of zirconium reacted.) Consequently, the resulting calculated H2 mass-generated is 448 kg. This is in close agreement with the hydrogen generation estimate of 460 kg by Hendrie (1989) and it also agrees well with the observations in the Phebus in-reactor experiments (Bourdon et al, 2002, Di Giuli et al, 2015 and Sangiorgi et al, 2015) for early phase hydrogen generation.
With the bottom half of the core being much cooler, the molten material froze and blocked the coolant flow passages in that region. This ended the steam flow through the core along with the “steam starvation” behavior. Post-accident observations of the TMI-2 lower core region confirmed the frozen debris, blocked flow path configuration. Equally important, this provides a fast, realistic method for estimating when major core damage could occur as well as reliable estimates of the possible extent of hydrogen generation during the early phases of the core degradation. This methodology is a straightforward, and fast execution time calculation used to characterize the early phase hydrogen generation in the RT-EVALS evaluations. The hydrogen generation rate evaluation transitions to the late phase hydrogen generation model discussed in LATE PHASE H2 GENERATION.
9. Late Phase H2 Generation
1.0 Background
With the intact fuel assembly configuration and the steam supply from the decay heat generated in the submerged portion of the fuel, the early phase hydrogen generation for a severe core damage event would have the conditions that could lead to a runaway chemical reaction. As discussed in Early Phase H2 Generation, this oxidation reaction is realistically represented as being limited by the extent of steam generated in the water covered part of the core when the upper region is at a sufficiently high temperature for a “steam starvation” condition to exist. Nevertheless, the early phase has a somewhat self-limited nature since the chemical energy released would melt a significant fraction of the core. The downward relocation combined with the subsequent freezing of this molten mass plugs and then destroys the fuel pin configuration and the capability for steam to flow through this region (see
2.0 Possible Sustained Oxidation During the Late Phase
The oxidation of the unreacted core materials could continue as steam could be circulated around the blocked region(s) to the upper surface of the debris bed where the lighter metallic constituents would likely tend to be concentrated (see
Q/SQRT[Dpow(5)g(Δρ/ρavg)]
or
Q=C0SQRT[Dpow(5)g(Δρ/ρavg)]
where
It is to be noted that the Phebus experiments were conducted by injecting steam at a fixed rate into the bottom of the in-reactor test assemblies regardless of the core condition. For a commercial power reactor accident, the compacted core configuration would be the result of molten debris relocating downward into the lower, cooler segments of the reactor core and freezing on these structures. Therefore, the only steam flow that could be produced in this configuration would be a limited amount generated by water vaporization in the lower plenum resulting from radiant heat transfer from the bottom of an overheated core region or some facet of the accident sequence that continues to produce steam that could be circulated through the Reactor Cooling System (RCS).
Considering the broad spectrum of accident conditions, it is possible that the accident sequence could influence the steam partial pressure above the core materials. For example, a Small Break Loss of Coolant Accident (SBLOCA) and the slow RCS depressurization could generate steam that could propagate through the RCS, or the accident could be initiated at an elevated pressure and the early stage of the core damage could be responsible for generating a leakage from the RCS by melting in-core instrumentation for some designs. In this accident progression phase, the steam partial pressure in the region above the degraded could be difficult to determine.
Whatever the accident sequence, a conservative estimate for the late phase oxidation behavior can be developed by assuming a sufficient supply of steam to support countercurrent flow natural circulation with a driving force determined by the density difference between steam and hydrogen at the same temperature as was observed in the three Phebus experiments (see
It is also important to consider that the countercurrent flows developed in a reactor system would have a much larger core upper surface area and the L/D should be considered as the order of unity. The coefficient characterizing the countercurrent flow could be somewhat larger than 0.05, perhaps as large as 0.1. However, for a reactor system, the “gas with the greater density” would likely be a mixture of steam and hydrogen and not pure steam as it was for the Phebus experiments.
For the accident evaluation, the major unknown is whether there is a significant flow of steam to the region above the fully degraded core geometry. As was noted in “Early Phase H2 Generation”, for the TMI-2 accident, the core was rapidly covered by water shortly after the early phase of core degradation. While this certainly provided steam to the upper surface, it also rapidly cooled the debris upper surface to form a crust that subsequently impeded the continued oxidation of the unreacted metals in the core. Nevertheless, this did not prevent the molten debris core from finding a relocation path out of the core region and into the water baffle, lower core support, and lower plenum regions. From this it can be deduced that initial cooling of the upper regions of the core debris and the relocation were central to cooling the core debris within the RPV.
10. Non-Nuclear Applications of RT-EVALS
While the concept of the RT-EVALS methodology began in considering how to efficiently use the available plant information in a manner that maximizes the use of the plant resources to counter any event trends that could damage the reactor core, there are two additional applications of this methodology that could be used in some parts of the petro-chemical industry. The first of these is a straightforward application of the methodology to fixed location chemical reactor facilities such as one, or more chemical reactors located within a chemical plant. A second application relates to the monitoring of chemical recipe shipments that are sensitive to changes in the environment (for example the ambient temperature history) and/or the duration of the transportation process.
With the fixed location application, the RT-EVALS methodology could be applied in essentially the same manner that it is for commercial nuclear power plants with the “Core Engineering Module” containing the representation of commercial chemical reaction kinetics that are associated with the specific chemical recipe undergoing a controlled exothermic chemical reaction. Examples of transients of interest would be a loss of cooling to the reaction vessel, an external fire near or around the reaction vessel and other event sequences. For the first transient, the exothermic reaction could enable a continuing increase in the recipe temperature which would accelerate the chemical reaction. The monitoring of the recipe temperature history would provide real time assessments of the reaction behavior, including uncertainty bands on the thermocouple measurements, as well as long term projections of the transient behavior. This would include the possibility that the pressure in the chemical reaction vessel may exceed the lifting pressure of the safety relief valve and discharge some of the recipe into a holding tank or scrubbing system downstream as well as the possibility that vapors and/or non-condensable gases could be released to the environment. For the second event, a fire external to the reaction vessel could potentially cause the recipe reaction rate to increase sufficiently to cause an overpressure in the vessel that would lift the safety relief valve. These processes would be characterized in the “Reactor Coolant System Engineering Module” and the “Containment Engineering Module” for the chemical reaction vessel and the downstream components respectively. For many of these applications, the “Pressurizer Engineering Module” and the “Steam Generator Engineering Module” would not be needed, so these would be bypassed. However, there could be other designs where these would be appropriate for the design. Of course, the “Evaluation Module” and the “Recommendations Actions and Projections of Near Term Behavior” components of the methodology would serve the same purpose that they would be have for commercial nuclear power plants even though the physical processes evaluated in the engineering modules would be specific to the system being considered.
For assessing the safety of a chemical recipe during transportation, RT-EVALS needs to be applied in a manner where the transient behavior of the recipe can be measured and interpreted during the transportation interval. Consequently, the measurements and the evaluation methodology must travel with the chemical recipe. Therefore, the implementation needs to be an on-board computerized hardware instrumentation package that includes a RT-EVALS evaluation methodology for the specific chemical recipe. In transit, the hardware package would continuously monitor the average temperature of the chemical recipe and use the methodology at each measurement interval to provide long term projections of the recipe behavior during the remaining transportation interval, given the status of the recipe (average temperature, the rate of temperature increase and the pressure) at any time. If this assessment results in an average recipe temperature being above the stated allowable temperature for the end of the transport, the evaluation package would transmit an alert along with real time estimates of the time available until the reaction could potential reach a “runaway condition” assuming the current environmental conditions remained unchanged. Along with this, the RT-EVALS evaluations within the hardware would recommend possible actions that could be taken. Each recommendation would be accompanied by an assessment of the long term behavior. Possible candidate actions could be: (i) spraying the external surface of the reaction vessel with water, (ii) quenching the chemical recipe by injection water into the reaction vessel, (iii) quenching the chemical reaction by injecting a chemical retardant, (iv) submerging the reactor vessel in water, (v) depressurizing the reaction vessel before the recipe temperature would reach “runaway conditions” and (vi) emptying the reaction vessel contents into a holding tank if the design permits. Some of these may not be applicable to a given design, but it is likely that more than one could be implemented.
As a result, the RT-EVALS would provide real time projections of the behavior of a chemical recipe during transportation when the reaction kinetics of the recipe (response to temperature and pressure) have been characterized in a laboratory experiment. If the real time projection should result in the recipe average temperature would be greater than the allowable limit before the end of the journey, the RT-EVALS will transmit an “alert” signal and develop a list of possible actions to be taken. RT-EVALS will develop real time projections for each recommended action and part of this assessment needs to be an estimate of the time required to implement a given action. The RT-EVALS enables the operating personnel to manually enter an estimate of the duration from the present time that would be needed to implement a recommended action. Hence, this would be part of the long term evaluations. As long as the average temperature of the chemical recipe remained greater than the safe limit for the end of the transportation, the hardware unit would continue to transmit an “alert” status along with a projection of the time available before a “runaway condition” could be possible if no actions would be taken. Once an action would be taken, the alert would continue to be broadcast with the recognition that an action has been taken and this would project the long term response considering the action taken. In summary, the on-board hardware instrument package, that travels with the chemical recipe, would continuously monitor the recipe average temperature and transmit the status along with real time long term projections of the recipe behavior during the transportation.
Referring now to figures,
A user 112, such as the one or more relevant parties, may access online platform 100 through a web based software application or browser. The web based software application may be embodied as, for example, but not be limited to, a website, a web application, a desktop application, and a mobile application compatible with a computing device 2800.
Further, the communication device 202 may be communicatively coupled with a reactor computer associated with a reactor. Further, the communication device 202 may be configured for receiving at least one reactor data associated with the reactor from the reactor computer. Further, the communication device 202 may be configured for receiving a plurality of reactor design data and a plurality of reactor measurement data associated with a plurality of reactor components of the reactor from the reactor computer. Further, the communication device 202 may be configured for transmitting at least one notification to at least one user device associated with at least one user.
Further, the processing device 204 may be configured for determining at least one reactor transient condition associated with the reactor based on the at least one reactor data. Further, the processing device 204 may be configured for analyzing the plurality of reactor design data and the plurality of reactor measurement data. Further, the processing device 204 may be configured for generating the at least one notification corresponding to the at least one reactor transient condition based on the analyzing. Further, the processing device 204 may be configured for developing the at least confirmation of the at least one notification corresponding to the at least one reactor transient condition based on the analyzing.
In further embodiments, the processing device 204 may be configured for analyzing the at least one reactor transient condition. Further, the processing device 204 may be configured for identifying at least one reactor component of the plurality of reactor components based on the analyzing. Further, the communication device 202 may be configured for receiving at least one reactor design data and at least one reactor measurement data corresponding to the at least one reactor component.
In further embodiments, the communication device 202 may be configured for receiving at least one independent reactor measurement data from at least one independent reactor measuring device associated with the plurality of reactor components of the reactor. Further, the communication device 202 may be configured for transmitting at least one confirmatory data to the at least one user device. Further, the processing device 204 may be configured for analyzing the at least one independent reactor measurement data and the at least one reactor transient condition. Further, the processing device 204 may be configured for analyzing the at least one independent reactor measurement data and the at least one reactor transient condition. Further, the processing device 204 may be configured for generating the at least one confirmatory data corresponding to the at least one reactor transient condition based on the analyzing.
In further embodiments, the processing device 204 may be configured for analyzing the at least one reactor transient condition. Further, the processing device 204 may be configured for generating at least one remedial action data corresponding to the at least one reactor transient condition based on the analyzing. Further, the communication device 202 may be configured for transmitting the at least one remedial action data to the at least one user device.
In further embodiments, the communication device 202 may be configured for receiving at least one manual entry associated with at least one reactor component of the plurality of reactor components from the at least one user device. Further, the processing device 204 may be configured for analyzing the plurality of reactor design data, the plurality of reactor measurement data, and the at least one manual entry.
In further embodiments, the communication device 202 may be configured for receiving at least one user control variable associated with the at least one reactor transient condition from the at least one user device. Further, the communication device 202 may be configured for transmitting at least one variable projection to the at least one user device. Further, the processing device 204 may be configured for analyzing the at least one user control variable and the at least one reactor transient condition. Further, the processing device 204 may be configured for generating the at least one variable projection corresponding to the at least one reactor transient condition based on the analyzing.
In further embodiments, the processing device 204 may be configured for determining a plurality of options corresponding to the at least one reactor transient condition. Further, the processing device 204 may be configured for generating at least one alert corresponding to the at least one option. Further, the communication device 202 may be configured for transmitting the plurality of options to the at least one user device. Further, the communication device 202 may be configured for receiving at least one option indication associated with at least one option of the plurality of options from at least one user device. Further, the processing device 204 may be configured for transmitting the at least one alert to at least one external user device associated with at least one external user.
In further embodiments, the communication device 202 may be configured for receiving at least one independent reactor measurement data from at least one independent reactor measuring device associated with the plurality of reactor components of the reactor. Further, the communication device 202 may be configured for transmitting at least one projection to the at least one user device. Further, the processing device 204 may be configured for analyzing the at least one independent reactor measurement data and the at least one reactor transient condition. Further, the processing device 204 may be configured for generating the at least one projection corresponding to the at least one reactor transient condition based on the analyzing.
Further, in some embodiments, the processing device 204 may include at least one engineering module, an evaluation module, and a decision module. Further, the engineering module may be configured for performing at least one engineering evaluation on the plurality of reactor design data and the plurality of reactor measurement data to generate at least one engineering analysis data corresponding to at least one engineering module. Further, the evaluation module may be configured for comparing the at least one engineering analysis data and identifying the at least one reactor transient condition. Further, the decision module may be configured for generating a plurality of options based on the at least one reactor transient condition.
Further, at 304, the method 300 may include a step of determining, using a processing device (such as the processing device 204), at least one reactor transient condition associated with the reactor based on the at least one reactor data. Further, the at least one reactor transient condition may refer to loss of load, loss off-site power, etc.
Further, at 306, the method 300 may include a step of receiving, using the communication device, a plurality of reactor design data and a plurality of reactor measurement data associated with a plurality of reactor components of the reactor from the reactor computer. Further, the plurality of reactor design data may include dimensions, designed pump flow rates, maximum power generation, etc. associated with the reactor. Further, the plurality of reactor design data may include maximum designed core operating power, reactor nuclear fuel assembly dimensions, number of assemblies, etc. Further, the plurality of rector measurement data may include core power, central rod position, system pressure, tailpipe temperature, etc.
Further, at 308, the method 300 may include a step of analyzing, using the processing device, the plurality of reactor design data and the plurality of reactor measurement data.
Further, at 310, the method 300 may include a step of generating, using the processing device, at least one notification corresponding to the at least one reactor transient condition based on the analyzing. Further, the at least one notification may facilitate the identification of the at least one reactor transient condition. Further, the at least one notification may include a textual content associated with the at least one reactor transient condition.
In some embodiments, the method 300 may include a step of generating confirmation of a notification corresponding to the reactor transient condition based on the analyzing.
Further, at 312, the method 300 may include a step of transmitting, using the communication device, the at least one notification to at least one user device associated with at least one user. Further, the at least one user may include an individual, an institution, and an organization that may want to receive the at least one notification corresponding to the at least one reactor transient condition. Further, at 312, the method 300 may include a step of transmitting, using the communication device, the confirmation to at least one user device associated with at least one user.
Further, at 404, the method 400 may include a step of identifying, using the processing device, at least one reactor component of the plurality of the reactor components based on the analyzing.
Further, at 406, the method 400 may include a step of receiving, using the communication device, at least one reactor design data and at least one reactor measurement data corresponding to the at least one reactor component.
Further, at 504, the method 500 may include a step of analyzing, using the processing device, the at least one independent reactor measurement data and the at least one reactor transient condition.
Further, at 506, the method 500 may include a step of generating, using the processing device, at least one confirmatory data corresponding to the at least one reactor transient condition based on the analyzing.
Further, at 508, the method 500 may include a step of transmitting, using the communication device, the at least one confirmatory data to the at least one user device.
Further, at 604, the method 600 may include a step of generating, using the processing device, at least one remedial action data corresponding to the at least one reactor transient condition based on the analyzing.
Further at 606, the method 600 may include a step of transmitting, using the communication device, the at least one remedial action data to the at least one user device.
Further, at 704, the method 700 may include a step of analyzing, using the processing device, the plurality of reactor design data, the plurality of reactor measurement data, and the at least one manual entry.
Further, at 804, the method 800 may include a step of analyzing, using the processing device, the at least one user control variable and the at least one reactor transient condition.
Further, at 806, the method 800 may include a step of generating, using the processing device, at least one variable projection corresponding to the at least one reactor transient condition based on the analyzing.
Further, at 808, the method 800 may include a step of transmitting, using the communication device, the at least one variable projection to the at least one user device.
Further, at 904, the method 900 may include a step of transmitting, using the communication device, the plurality of options to the at least one user device.
Further, at 906, the method 900 may include a step of receiving, using the communication device, at least one option indication associated with at least one option of the plurality of options from at least one user device.
Further, at 908, the method 900 may include a step of generating, using the processing device, at least one alert corresponding to the at least one option.
Further, at 910, the method 900 may include a step of transmitting, using the communication device, the at least one alert to at least one external user device associated with at least one external user.
Further, at 1004, the method 1000 may include a step of analyzing, using the processing device, the at least one independent reactor measurement data and the at least one reactor transient condition.
Further, at 1006, the method 1000 may include a step of generating, using the processing device, at least one projection corresponding to the at least one reactor transient condition based on the analyzing.
Further, at 1008, the method 1000 may include a step of transmitting, using the communication device, the at least one projection to the at least one user device.
Further, at 1104, the method 1100 may include a step of determining, using a processing device, at least one reactor transient condition associated with the reactor based on the at least one reactor data.
Further, at 1106, the method 1100 may include a step of analyzing, using the processing device, the at least one reactor transient condition.
Further, at 1108, the method 1100 may include a step of identifying, using the processing device, at least one reactor component of the plurality of the reactor components based on the analyzing.
Further, at 1110, the method 1100 may include a step of receiving, using the communication device, at least one reactor design data and at least one reactor measurement data corresponding to the at least one reactor component.
Further, at 1112, the method 1100 may include a step of evaluating, using the processing device, the at least one reactor design data and the at least one reactor measurement data.
Further, at 1114, the method 1100 may include a step of generating, using the processing device, at least one notification corresponding to the at least one reactor transient condition based on the evaluation
Further, at 1116, the method 1100 may include a step of transmitting, using the communication device, at least one notification to at least one user device associated with at least one user.
Further, at 1204, the method 1200 may include a step of analyzing, using the processing device, the at least one independent reactor measurement data and the at least one reactor transient condition.
Further, at 1206, the method 1200 may include a step of generating, using the processing device, at least one confirmatory data corresponding to the at least one reactor transient condition based on the analyzing.
Further, at 1208, the method 1200 may include a step of transmitting, using the communication device, the at least one confirmatory data to the at least one user device.
Further, at 1304, the method 1300 may include a step of generating, using the processing device, at least one remedial action data corresponding to the at least one reactor transient condition based on the analyzing.
Further, at 1306, the method 1300 may include a step of transmitting, using the communication device, the at least one remedial action data to the at least one user device.
NF=Q/SQRT[Dpow(5)g(Δρ/ρavg)]
or
Q=C0SQRT[Dpow(5)g(Δρ/ρavg)]
Where, Δρ=difference between the densities of the steam and hydrogen
ρavg=average of the two gas densities
g=gravitational acceleration and
C0=an empirical coefficient that replaces the Froude number since experiments show this to be a function of the length-to-diameter (L/D) ratio for the natural circulation flow.
Further, the graphical representation 2300 may show the results (gray dots) of the calculated pressurization assuming the PZR PORV is stuck open and with a steam-water mixture discharging into the drain tank that is half full of water. This simple calculation, which may be performed much faster than real time, is in good agreement with the measured behavior. Consequently, the discharge flow rate may be estimated from the measured pressurization rate if it was needed. In summary, what this RT-EVALS methodology accomplishes is the immediate usage of all the relevant information to detect an evolving challenge and determine the depth of confirmation of the conclusion. This may be accomplished through straightforward calculations that may be executed essentially as rapidly as the data may be available from the plant computer.
With reference to
Computing device 2800 may have additional features or functionality. For example, computing device 2800 may also include additional data storage devices (removable and/or non-removable) such as, for example, magnetic disks, optical disks, or tape. Such additional storage is illustrated in
Computing device 2800 may also contain a communication connection 2816 that may allow device 2800 to communicate with other computing devices 2818, such as over a network in a distributed computing environment, for example, an intranet or the Internet. Communication connection 2816 is one example of communication media. Communication media may typically be embodied by computer readable instructions, data structures, program modules, or other data in a modulated data signal, such as a carrier wave or other transport mechanism, and includes any information delivery media. The term “modulated data signal” may describe a signal that has one or more characteristics set or changed in such a manner as to encode information in the signal. By way of example, and not limitation, communication media may include wired media such as a wired network or direct-wired connection, and wireless media such as acoustic, radio frequency (RF), infrared, and other wireless media. The term computer readable media as used herein may include both storage media and communication media.
As stated above, a number of program modules and data files may be stored in system memory 2804, including operating system 2805. While executing on processing unit 2802, programming modules 2806 (e.g., application 2820 such as a media player) may perform processes including, for example, one or more stages of methods, algorithms, systems, applications, servers, databases as described above. The aforementioned process is an example, and processing unit 2802 may perform other processes. Other programming modules that may be used in accordance with embodiments of the present disclosure may include machine learning applications.
Generally, consistent with embodiments of the disclosure, program modules may include routines, programs, components, data structures, and other types of structures that may perform particular tasks or that may implement particular abstract data types. Moreover, embodiments of the disclosure may be practiced with other computer system configurations, including hand-held devices, general purpose graphics processor-based systems, multiprocessor systems, microprocessor-based or programmable consumer electronics, application specific integrated circuit-based electronics, minicomputers, mainframe computers, and the like. Embodiments of the disclosure may also be practiced in distributed computing environments where tasks are performed by remote processing devices that are linked through a communications network. In a distributed computing environment, program modules may be located in both local and remote memory storage devices.
Furthermore, embodiments of the disclosure may be practiced in an electrical circuit comprising discrete electronic elements, packaged or integrated electronic chips containing logic gates, a circuit utilizing a microprocessor, or on a single chip containing electronic elements or microprocessors. Embodiments of the disclosure may also be practiced using other technologies capable of performing logical operations such as, for example, AND, OR, and NOT, including but not limited to mechanical, optical, fluidic, and quantum technologies.
In addition, embodiments of the disclosure may be practiced within a general-purpose computer or in any other circuits or systems.
Embodiments of the disclosure, for example, may be implemented as a computer process (method), a computing system, or as an article of manufacture, such as a computer program product or computer readable media. The computer program product may be a computer storage media readable by a computer system and encoding a computer program of instructions for executing a computer process. The computer program product may also be a propagated signal on a carrier readable by a computing system and encoding a computer program of instructions for executing a computer process. Accordingly, the present disclosure may be embodied in hardware and/or in software (including firmware, resident software, micro-code, etc.). In other words, embodiments of the present disclosure may take the form of a computer program product on a computer-usable or computer-readable storage medium having computer-usable or computer-readable program code embodied in the medium for use by or in connection with an instruction execution system. A computer-usable or computer-readable medium may be any medium that can contain, store, communicate, propagate, or transport the program for use by or in connection with the instruction execution system, apparatus, or device.
The computer-usable or computer-readable medium may be, for example but not limited to, an electronic, magnetic, optical, electromagnetic, infrared, or semiconductor system, apparatus, device, or propagation medium. More specific computer-readable medium examples (a non-exhaustive list), the computer-readable medium may include the following: an electrical connection having one or more wires, a portable computer diskette, a random-access memory (RAM), a read-only memory (ROM), an erasable programmable read-only memory (EPROM or Flash memory), an optical fiber, and a portable compact disc read-only memory (CD-ROM). Note that the computer-usable or computer-readable medium could even be paper or another suitable medium upon which the program is printed, as the program can be electronically captured, via, for instance, optical scanning of the paper or other medium, then compiled, interpreted, or otherwise processed in a suitable manner, if necessary, and then stored in a computer memory.
Embodiments of the present disclosure, for example, are described above with reference to block diagrams and/or operational illustrations of methods, systems, and computer program products according to embodiments of the disclosure. The functions/acts noted in the blocks may occur out of the order as shown in any flowchart. For example, two blocks shown in succession may in fact be executed substantially concurrently or the blocks may sometimes be executed in the reverse order, depending upon the functionality/acts involved.
While certain embodiments of the disclosure have been described, other embodiments may exist. Furthermore, although embodiments of the present disclosure have been described as being associated with data stored in memory and other storage mediums, data can also be stored on or read from other types of computer-readable media, such as secondary storage devices, like hard disks, solid state storage (e.g., USB drive), or a CD-ROM, a carrier wave from the Internet, or other forms of RAM or ROM. Further, the disclosed methods' stages may be modified in any manner, including by reordering stages and/or inserting or deleting stages, without departing from the disclosure.
Although the present disclosure has been explained in relation to its preferred embodiment, it is to be understood that many other possible modifications and variations can be made without departing from the spirit and scope of the disclosure.
| Number | Date | Country | |
|---|---|---|---|
| 62879299 | Jul 2019 | US |
