Hoppe_M7_LabReport

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Excelsior University *

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NUC 323

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Feb 20, 2024

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Course: NUC350 Section: Module 7 Instructor Name: Professor Mathus Name(s): Zachary Hoppe __________________________________________________________________________ Title : Turbine Load Reduction and Loss of Coolant Accident in a Pressurized Water Reactor. _______________________________________________________________________ Abstract: This lab allowed the observation and analysis of two large transients in a Pressurized Water Reactor simulation. This is a culmination of the prior labs in this class, allowing for an unidentified scenario to be analyzed, discovering the cause of the plant transient and identifying the key points of plant response to the two different scenarios. This lab allows for a significant amount of data to be interpreted and understood, showing an advanced understanding of integrated plant operations for nuclear power. __________________________________________________________________________ Introduction: The primary objective of this lab is to complete two review scenarios on the PWR simulator. These scenarios will provide an opportunity to integrate the practices and knowledge obtained in prior labs, as well as to demonstrate a sufficient and practical understanding of integrated plant operations and response. PWRs have safety ingrained throughout their design, as well as a high margin of stability across different plant transients and casualties. Even a worst-case scenario is designed to be contained and mitigated with an automatic shutdown of key systems, and automatic initiation of protective functions. This design is critical to maintain safe and continuous operation of nuclear power plants. It is therefore important for to have the ability to recognize proper plant response to transients, as well as the required response to significant casualties that may bring the plant into departure from design conditions, where damage and danger become possibilities. This lab will serve to identify these scenarios, and anaylze the appropriate plant response to ensure full understanding of critical systems, as well as confirm appropriate response to unintended events. _________________________________________________________________________

Methods: Part B: a) Load the PWR simulator with “Module 7: Review Scenario 1” selected. b) Select “Run” on the ITS Panel to begin the simulation. c) Observe and record data to answer the following questions: i. Identify the plant event that is taking place. ii. Describe the response of the following plant system parameters: 1. Turbine Generator Output 2. Reactor thermal power 3. Reactor Coolant System Parameters (Including temperatures, pressure, Pressurizer Pressure, and Pressurize Level) 4. Main Steam System parameters (Including steam flow and steam pressure. iii. Describe the response of the following plant control systems: 1. Rod Control System 2. Pressurizer Pressure Control System 3. Pressurizer Level Control System iv. Describe the response of the following Engineered Safety Feature signals and systems, if applicable: 1. Reactor Protection System (Reactor Trip Signals) 2. Engineered Safety Feature Signal – Reactor Coolant Injection Signal 3. Engineered Safety Feature Signal – Containment Isolation Signal 4. Engineered Safety Feature Signal – Main Steam Isolation Signal 5. Engineered Safety Feature System – Reactor Coolant Injection System Part C:

a) Load the PWR simulator with “Module 7: Review Scenario 2” selected. b) Select “Run” on the ITS Panel to begin the simulation. c) Observe and record data to answer the following questions: i. Identify the plant event that is taking place. ii. Describe the response of the following plant system parameters: 1. Turbine Generator Output 2. Reactor thermal power 3. Reactor Coolant System Parameters (Including temperatures, pressure, Pressurizer Pressure, and Pressurize Level) 4. Main Steam System parameters (Including steam flow and steam pressure. iii. Describe the response of the following plant control systems: 1. Rod Control System 2. Pressurizer Pressure Control System 3. Pressurizer Level Control System iv. Describe the response of the following Engineered Safety Feature signals and systems, if applicable: 1. Reactor Protection System (Reactor Trip Signals) 2. Engineered Safety Feature Signal – Reactor Coolant Injection Signal 3. Engineered Safety Feature Signal – Containment Isolation Signal 4. Engineered Safety Feature Signal – Main Steam Isolation Signal 5. Engineered Safety Feature System – Reactor Coolant Injection System ________________________________________________________________________ Results:

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Part B: 7) Utilizing data collected, answer the following questions: i. Identify the plant event that is taking place. This evolution consisted of plant response to an electrical load reduction. ii. Describe the response of the following plant system parameters: 1. Turbine Generator Output TG output lowered from 1400MW to 1000MW steadily as the transient occurred. The following graphic illustrates this trend. 2. Reactor thermal power: Reactor thermal power lowered as TG loading did, showing a normal response of following steam demand. The following graphic illustrates this trend.

3. Reactor Coolant System Parameters (Including temperatures, pressure, Pressurizer Pressure, and Pressurize Level) Tavg lowered slightly as the programmed Tref lowered in response to a reduction in plant power. Plant pressure and Pressurizer Pressure followed a matching trend, as the pressurizer is the driving control of coolant system pressure. During the initial power mismatch, there was a brief insurge into the pressurizer, causing pressure and level to rise as the steam bubble was compressed. As power lowered and coolant temperature lowered, the pressurizer went through an outsurge, causing level and pressure to lower, until the heaters activated and maintained level and pressure without significant change on the parameters. The following graphics illustrate these trends.

4. Main Steam System parameters (Including steam flow and steam pressure. Main Steam flow lowered throughout the transient, as the turbine throttles were shut to limit supply as load decreased. Steam Generator pressure initially rose up due to the backpressure from reduced flow, and slowly lowered back to normal values as the plant reached a steady state equilibrium. There was a brief peak in steam flow when the steam dumps opened to help normalize plant conditions. The following graphics illustrate these trends.

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iii. Describe the response of the following plant control systems: 1. Rod Control System All rods were initially at the top of the core. The rod control system was operating with CEA Mode Select in “Auto Sequential,” and responded to the power mismatch and Tavg-Tref mismatch by driving in the control group of rods, Group 5. At the 56 position of Group 5, Group 4 also began an insertion. As Group 4 reached the 56 position, Group 6 rods began a small insertion, securing at the 144 position. Inward motion secured for Group 5 at the 5 position bottom stop, and Group 4 at the 50 position. The following graphic shows final rod configuration at the end of the transient. 2. Pressurizer Pressure Control System As pressurizer pressure initially rose, the Pressurizer Master Controller signalled to supply spray flow into the pressurizing, slightly collapsing the steam bubble. When the outsurge of coolant caused a decreasing pressure, the BLANK pressurizer heaters energized to maintain pressurizer pressure. As pressure was restored, the heaters deenergized. 3. Pressurizer Level Control System

The Pressurizer Level Master Controller made continuous changes to the charging flow into the pressurizer to maintain programmed level during the insurge and outsurge. Post-transient level was slightly lower than the initial level at full power. The following graph shows pressurizer level in comparison to program level. iv. Describe the response of the following Engineered Safety Feature signals and systems, if applicable: 1. Reactor Protection System (Reactor Trip Signals) There were no trip signals from the reactor protection system during this transient. 2. Engineered Safety Feature Signal – Reactor Coolant Injection Signal There were no Engineered Safety Features actuated during this transient. 3. Engineered Safety Feature Signal – Containment Isolation Signal There were no Engineered Safety Features actuated during this transient. 4. Engineered Safety Feature Signal – Main Steam Isolation Signal There were no Engineered Safety Features actuated during this transient. 5. Engineered Safety Feature System – Reactor Coolant Injection System There were no Engineered Safety Features actuated during this transient. Part C:

7) Utilizing data collected, answer the following questions: i. Identify the plant event that is taking place. Rapidly lowering pressurizer level combined with a rapidly increasing containment pressure indicate a large LOCA. ii. Describe the response of the following plant system parameters: 6. Turbine Generator Output Turbine generator output dropped rapidly to 0MW when the turbine tripped. The following graphic illustrates this trend. 7. Reactor thermal power Reactor power went from 100% to approximately 4.5% as soon as the reactor scrammed. The 4.5% residual power is due to decay heat. The following graphic illustrates this trend. 8. Reactor Coolant System Parameters (Including temperatures, pressure, Pressurizer Pressure, and Pressurize Level) Reactor coolant pressure dropped substantially as the reactor tripped. Tavg quickly reached the “minimum” indication of 530,

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with Thot lowering down to approximately 270F, and Tcold averaging approximately 303F. As the emergency injection and cooling systems continue to operate, temperatures will continue to drive downwards. Reactor coolant pressure and pressurizer pressure followed identical trends, rapidly decreasing several hundred pounds on the initiation of the leak, and then steadily lowering down to a depressurized condition due to the size of the rupture. Pressurizer level dropped to zero almost instantly, and was the immediate major indication of the LOCA. The following graphics illustrate these trends. 9. Main Steam System parameters (Including steam flow and steam pressure.

Steam flow dropped to zero as the steam isolation valves tripped shut. Steam pressure initially spiked due to the loss of loading, but then steadily drifted downward as the heat input from the reactor was lost. The following graphics illustrate these trends. Describe the response of the following plant control systems: 10. Rod Control System Initially at full power, all rods were fully withdraw. In response to the reactor trip signal, all rod control devices deenergized, allowing all rods to become fully inserted. The following graphic illustrates final rod positioning. 11. Pressurizer Pressure Control System Upon initiation of the casualty, the pressurizer pressure master controller defaulted to a signal of 0 to prevent energization of heaters while uncovered.

12. Pressurizer Level Control System The pressurizer level master controller attempted to make up for leakage with a high indication, driving more charging flow, but was unable to compete with the severity of the casualty and had no effect. The following graphic illustrates the level mismatch against program band. iii. Describe the response of the following Engineered Safety Feature signals and systems, if applicable: 13. Reactor Protection System (Reactor Trip Signals) A large number of reactor trip signals were received throughout the scenario due to the severity of the LOCA and its impact on various plant parameters. The initiating trip that caused the reactor to trip was PRESS PZR LOW, actuated when Pressurizer Pressure dropped below 1830psig. The following graphic shows the signaled trips shortly after the initiation of the casualty.

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14. Engineered Safety Feature Signal – Reactor Coolant Injection Signal The Safety Injection Signal activated at the same time as the reactor trip, due to low pressurizer pressure. 15. Engineered Safety Feature Signal – Containment Isolation Signal The containment isolation signal was activated when containment pressure reached 3.5psig, approximately two minutes after the initiation of the event. 16. Engineered Safety Feature Signal – Main Steam Isolation Signal The main steam isolation signal activated when containment pressure raised above 17psig. 17. Engineered Safety Fature System – Reactor Coolant Injection System All stages of reactor coolant injection were activated. The HPCI system was continuously injecting make up water, but was unable to compensate for the extreme leak rate in the primary system. At approximately 600psig, the accumulators were able to discharge into the core, rapidly lowering their level to zero.

At approximately 530psig, the RHR pumps, which had been active already, able to inject water into the primary due to pressure lowering beneath their maximum discharge head. Pressurizer level did not restore, and pressure maintained approximately 0psig long into the simulation. The following graphics illustrate the ECF active signals, as well as operating injection systems.

__________________________________________________________________________ Discussion : Part B: The first scenario illustrated a large steam demand transient in a PWR. Turbine loading decreased approximately 400MW within a few minutes, causing large changes across a variety of plant parameters. It was an excellent display of the automatic rod control system, as rod positions and speeds were adjusted in order to correct the temperature mismatch at the new power level once reactor power followed steam demand, proportionally lowering. The effects of the turbine throttles repositioning were observable in the changes in steam pressure and flow as the plant took time to reach a new equilibrium.

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The thermal effects also translated into pressurizer level, showing a notable transient due to the insurge and outsurge caused by large temperature changes in the reactor coolant system. The pressurizer level and pressure master controllers worked flawlessly to signal heaters, spray, and charging flowrate to compensate for these changes. Part C: In the second simulation, a severe LOCA was initiated, causing an almost immediate trip of the reactor plant. Pressurizer level dropped out of sight almost instantly, and plant pressure rapidly decreased to a depressurized condition shortly after the initiation of the event. Although ESF systems were quick to respond and actuated correctly, they were unable to restore level or pressure within the timeframe of the simulation. Continuously lowering primary temperatures and a stable level of decay heat showed that emergency systems were successfully in providing constant decay heat removal and shutdown cooling, preventing a serious thermal event in the fuel. The containment building was successful in preventing the release of any contaminants to the surrounding environment, and the containment spray system was able to maintain pressures and temperatures within limits. The Reactor Protection Complex activated without hesitation, detecting a large series of limit violations and unsafe conditions simultaneously as well as throughout the duration of the event. This provided for immediate automatic protective features, such as the reactor trip and setting containment, which prevented this serious LOCA from escalating into a further disaster. __________________________________________________________________________ Conclusion: This lab allowed for two important observations of reactor plant transients. The first showed a major, though well within design limits, load reduction, causing a transient that had widespread effects throughout plant systems. Rod control and pressurizer controllers played a critical role in maintaining appropriate plant parameters. The analysis of transients in pressure, temperature, and steam pressures allowed for a detailed understanding of the response of the integrated plant systems in response to a large change in power. The second situation was a severe LOCA, resulting in a depressurized reactor coolant system being maintained by emergency injection. This casualty showed the results of a major loss of reactor coolant inventory, and how it impacts pressure and temperature within the core. The response of the Reactor Protection complex and Engineered Safety Systems was invaluable in understanding core response to a worst-case scenario, allowing for insight and knowledge on how to adequately prepare for and respond to such events.

This lab also allowed for prior knowledge and experiences in the course to be leveraged in order to fully understand and identify the events taking place. __________________________________________________________________________ References: 3KEYSTUDENT Generic PWR Simulator . (n.d.). WS-Corp. https://www.3keystudent.com/ M7.5 Lab: PWR Simulations. Excelsior University. https://excelsior.instructure.com/courses/36956/assignments/1036158?module_item_id=3316253