# Nuclear power plant - Control system

• Luchekv
In summary: This is done by a feedback mechanism which senses the core temperature.So far, this is the best I can do. I'm not an engineer, so any help or criticism would be much appreciated.In summary, Luchek is researching a way to control the power output of a nuclear plant using control rods. He is new to the field and needs help understanding the process. He believes that the control rods indirectly control the heat by controlling the neutron flux. He is looking to create a mathematical model to understand how the power is controlled. His initial question was how active the control rods are, and Gmax
Luchekv
Hi guys,

I do apologize if this thread is in the incorrect place..

Long story short, I'm doing a university project based on developing an AI to control the power plant energy output. From my research so far, I've concluded that the control rods control the fission rate. The problem is, I for the life of me cannot find whether the control rods are only used to completely stop the reaction or if they're lowered/raised accordingly to control the heat and keep it at a specified value.

So far my plan is to have the control system engage the control rods in a percentage manner e.g. If the heat generated is exceeding the optimum value...the rods are raised to 75% to bring it back down after which the system lowers them back down. That's how I visualize the process at the moment.

If someone could confirm that for me and if that is the case...where could I find the mathematics behind how much heat is generated from uranium-235 and at what rate is the reaction slowed if the control rods are at 100%

I must admit, I am very new to nuclear engineering, I come from an electrical/system control background.

Luchek

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Here's a discussion on nuclear control rods:

https://en.wikipedia.org/wiki/Control_rod

They don't control the heat but rather the neutron flux which controls the rate of fission and the end result is the heat generated that in turn creates steam and then theelectricity generated.

I've already had a read of that article, and you are correct in your statement of they indirectly control heat through controlling flux.
But my problem is with what it says in the last sentence:

"Typical shutdown time for modern reactors such as the European Pressurized Reactor or Advanced CANDU reactor is 2 seconds for 90% reduction, limited by decay heat."

Which makes me think that control rod activity isn't very often seeing as it only needs 2 secs to create a 90% reduction..meaning there is another way of maintaining the desired temps...

EDIT: Not 100% sure that the sentence means shutdown can be achieved solely by control rods being engaged for 2 seconds..

The core power is controlled by the negative temperature reactivity feedback (as coolant temperatures go down, reactivity and power go up). So, the core power is really controlled by the turbine-generator output. If the turbine throttle valves are opened wider, more steam is pulled, this drops the core temperature, which increases reactivity, and the core power goes up to match the new turbine demand. But, now the core temperature is low. So, the control rods (and soluble boron, in PWRs) are used to restore the core inlet temperature to the desired value. For power reactors running at essentially constant power, the control rods normally don't move at all (other than small amounts just to exercise the mechanisms). Some load-following designs move the rods around to maintain desired temperatures as a function of the changing load.

But the main take away is, core power follows turbine power. Not the other way around.

That 2 seconds to decrease power to 90% is ~ 2 seconds for the rods to fall in. The control mechanism let's the operators move the rods slowly, or just release them so they fall right in (that typically takes 2 or 3 seconds).

dlgoff and Luchekv
Great post Gmax, I'm looking at buying an 'intro to nuclear engineering' book to fully understand the process. However I probably should have defined the scope of my project better. The primary objective of it is to argue the benefits of an AI control system as opposed to a conventional one, benefits being precision, no need to calibrate, trend and pattern recognition, adaptive learning and future prevention of certain scenarios etc etc..not necessarily the design of the plant itself

In saying that I'm looking at a very basic structure of the plant. I've watered it down to a simple Rankine cycle with a boiler, turbine, condenser and a pump. Before I can apply my control theory I need a mathematical model which is why I asked my initial question of how active the control rods are(to see how accurate I can create the simulation). Your answer is a bit of a set back (in terms of how its actually controlled), but that is something I will address in the report and will have to continue on with my current plan of attack as this is a solo project and I don't think I'd have the time to accurately portray the full process..

My plan of attack for the mathematical model was to figure out how much heat is generated by fission per minute so I was thinking of something along the lines of:
A very crude equation...Heat generated = Neutron flux - (Neutron flux absorbed by rods*(percentage of rod lowered))

Power to the core is assumed constant and the water supplied to the tank is also at a constant level. Which just leaves me to work out at what rate the water is turned to steam vapor and the pressure build up...and then again transpose those values to the turbine to get a kW output. So my control system will just focus on inserting/taking out the rods as needed to stabilize the cycle.

I have also found this equation for power released in a reactor: (I'm assuming its thermal energy)
https://physicsforums-bernhardtmediall.netdna-ssl.com/data/attachments/79/79118-ab7c405aa0d5a08d46da0274ea2c6e3a.jpg
What are your thoughts on my approach?

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The older version of Glasstone and Sessonske has a good chapter on reactor controls

this edition was my textbook in 1968
http://www.ebay.co.uk/itm/1963-NUCLEAR-REACTOR-ENGINEERING-Glasstone-Sesonske-HC-DJ-/181909726789

newer ones omitted the chapters on instrumentation

this link may be of help, too

Luchekv
Luchekv said:
In saying that I'm looking at a very basic structure of the plant. I've watered it down to a simple Rankine cycle with a boiler, turbine, condenser and a pump. Before I can apply my control theory I need a mathematical model which is why I asked my initial question of how active the control rods are(to see how accurate I can create the simulation). Your answer is a bit of a set back (in terms of how its actually controlled), but that is something I will address in the report and will have to continue on with my current plan of attack as this is a solo project and I don't think I'd have the time to accurately portray the full process..
One should be aware of the difference between PWRs and BWRs.

In most PWRs, control rods are withdrawn during operation, and reactivity control is maintained by soluble boron in the coolant, in addition to the moderator density and fuel temperature (Doppler feedback). Some PWRs use grey rods. B&W units were designed with axial power shaping rods (APSR) which used mainly Inconel absorbers, and EdF PWRs use particular grey rods. The Westinghouse AP1000 uses a special kind of grey rod to enable load following.

Some reactivity control (power distribution) is realized by burnable absorbers in the fuel (ZrB2 and Gd, or Er), or burnable poison assemblies which reside in guide-tubes of assemblies not under control rods.

Some PWRs have a fair amount of nucleate boiling.

BWRs use control rods in the core during operation, and they do not use soluble boron in the coolant. Like PWR fuel, the BWR fuel will contain a burnable absorber, usually Gd in the form of gadolinia. Groups (4 or 8) of control rods are used in deep and shallow combinations, and as the cycle progresses, the patterns are changed, and the amount of control rods used decreases, until ARO (all rods out).

Core monitoring and periodic calibration will not be going away with an AI system.

Michal Kovac and Luchekv
A typical pwr will have control rods in automatic temperature control. The rods function in this mode and are used this way primarily during startup/shutdown at low power conditions.

When the turbine is online, the steam draw from the steam generators causes changes in core power and core deltaT. As you raise steam load, your core average temperature tends to go down, so you have to either remove control rods or dilute boron. At a certain point you end up removing all rods and using boration to affect reactor coolant average temperature, while using turbine throttle rate to affect deltaT and reactor power.

Control rods affect average temp, steaming affects deltaT. And what's challenging is if your turbine load set is at 1000 MW, and you insert control rods, core average temperature drops, but the turbine draws more steam to maintain 1000 MW, causing larger deltaT, colder inlet Temps, reactor power going up, and less margin to thermal limits.

BWRs are operated differently and is easier to model. The turbine operates in pressure control mode. It will automatically modulate how much steam it draws from the core to maintain reactor pressure at a near constant point. So if reactor power goes down, pressure will drop, and the turbine throttles will go shut to maintain pressure.

For a BWR, rods are used for both local flux control, and global power control. Typically, we set a rod pattern around 45-50% and then we use core recirculation flow to raise power from 50% to 100%. Control rods are rarely used to maintain power above 50% in a BWR, and are only notched out for compensating against fuel depletion.

My BWR originally had automatic flux control. The control system would vary the recirculation flow through the core automatically to maintain a desired power level. The rods are all manual. We ended up removing automatic flux control due to calibration issues and some weird responses. So now we operate entirely in manual flow control and manual rods.

The scram system is completely separate from the rod control system. For a pwr, the scram system will energize grippers which can modulate to slowly insert or remove control rods. When a scram is required, the grippers are deenergized causing the rods to drop into the core under gravity. For a BWR, the rods have hydraulic accumulators which are held shut by power and air supply, and the scram system will open the accumulators up and they will rapidly insert the rods. The point in trying to make, is normal rod drive systems are very slow and are meant for controlling the reactor. Inserting 4 rods into my core takes over a minute. And we have 145 of them. A scram is a protection system action which trips all control rods simultaneously to shut the plant down. That does happen in 2 seconds. But the normal rod drive systems have limitations on insertion and withdrawal rates to prevent excessive lower rates.

Luchekv
Just a small addition, you never withdraw control rods completely in a PWR. One bank will be inserted a few steps (called "bite" position) to retain a minimum differential control rod worth.

Luchekv
I really do admire your knowledge guys, I wish my university was offering such a course...I'm still trying to get my head around the processes and the relationships of core power, core temp, reactor power etc...so bear with me in that regard.

Astronuc said:
One should be aware of the difference between PWRs and BWRs.

In most PWRs, control rods are withdrawn during operation, and reactivity control is maintained by soluble boron in the coolant, in addition to the moderator density and fuel temperature (Doppler feedback). Some PWRs use grey rods. B&W units were designed with axial power shaping rods (APSR) which used mainly Inconel absorbers, and EdF PWRs use particular grey rods. The Westinghouse AP1000 uses a special kind of grey rod to enable load following.

Some reactivity control (power distribution) is realized by burnable absorbers in the fuel (ZrB2 and Gd, or Er), or burnable poison assemblies which reside in guide-tubes of assemblies not under control rods.

Some PWRs have a fair amount of nucleate boiling.
(Whats nucleate boiling?)
From what I've read, I think simulating a PWR control system would be easier, as already mentioned..the control rods won't be shooting in an out to control reaction, its an unrealistic approach. So from what I can envision, I would have a separate tank feeding in soluble boron into the coolant loop?

Few questions on the soluble boron though, for example you put in 1Kg of it in...will that boron eventually be 'used up' what I mean is, will it get to a point where it can't absorb anymore neutron flux? or does it just stay in the coolant loop until you have to dilute it?

I think this is a better way to go in terms of modelling I can keep everything in mass flow rate terms instead of worrying about rate of steam generation as well.

It didn't allow me to edit my previous post, so I apologize for the double post.

Hiddencamper said:
A typical pwr will have control rods in automatic temperature control. The rods function in this mode and are used this way primarily during startup/shutdown at low power conditions.

BWRs are operated differently and is easier to model. The turbine operates in pressure control mode. It will automatically modulate how much steam it draws from the core to maintain reactor pressure at a near constant point. So if reactor power goes down, pressure will drop, and the turbine throttles will go shut to maintain pressure.

For a BWR, rods are used for both local flux control, and global power control. Typically, we set a rod pattern around 45-50% and then we use core recirculation flow to raise power from 50% to 100%. Control rods are rarely used to maintain power above 50% in a BWR, and are only notched out for compensating against fuel depletion.

My BWR originally had automatic flux control. The control system would vary the recirculation flow through the core automatically to maintain a desired power level. The rods are all manual. We ended up removing automatic flux control due to calibration issues and some weird responses. So now we operate entirely in manual flow control and manual rods.

Even though Hiddencamper mentioned it would be easier to model a BWR, my problem is again on trying to simplify the system, he mentions that rods AND recirculation flow is used. I know its stupid of me try and simplify such a complex system and you can only go so far before its inaccurate.. but again time isn't on my side, that would be 2 control loops I would have to design...where as with the PWR I could just focus on the levels of boron in the coolant and leave the rods as an emergency stop. Would you guys say that's a decent..I guess, estimation of the process?

For a PWR, boron is typically controlled manually.

Talking about a PWR, and the complications modelling it. Before the turbine is online and you are actively steaming, your control rods determine your reactor power. Once the turbine is online, reactor power is going to be based on turbine steam load. Once you are on the turbine, your control rods (and boron) only control average reactor coolant system temperature, not reactor power.

To understand this: If you are at 50% reactor power, and you pull control rods out (or dilute boron), reactor power initially increases, causing the temperature of water exiting the reactor to go up. Hotter water goes to the steam generator. The amount of steam drawn from the steam generators is set by the turbine, and remains constant, so you now have hotter water going into the SGs, but very little increase in heat removal from the reactor. The temperature of water going into the cold side of the reactor goes up, causing power to drop back down to around 50% again. Average reactor coolant system temperature will go up as a result, but power stays around the same amount.

Another scenario: I want to raise power up, I'm at 50%. I open the turbine throttle valves to draw more steam. The steam leaving the steam generator removes more heat from the reactor, additionally more feedwater is admitted to the steam generator which adds more cold water. This causes the temperature of water going into the cold leg of the reactor to drop. Colder water causes reactor power to increase, and the temperature of water on the hot side of the reactor goes up. Average reactor temperature drops slightly, as a result of this power increase, additionally the "Reference temperature", which is a measure of what the average reactor temperature should be for a given power level, goes up. So now I need to raise average temperature to maintain it equal to reference temperature. The operators will withdraw control rods, or dilute boron, the restore T_ave to T_ref.

A naval example: Reactor is at 15-20% power, and the bell sounds for Ahead flank cavitate, basically, raise reactor power to maximum as fast as possible to move the sub as fast as possible for an evasive maneuver. The throttleman will open the turbine throttles as fast as possible, drawing as much steam as he can to raise propeller RPM. Reactor power is going to rapidly raise as steam flow increases. Due to the fact that the reactor power takes about 5-7 seconds to change after a steam flow change, if the throttleman overshoots, he can bring reactor power above 100%. The reactor operator may see this is about to happen and push rods in at 99.5% to ensure they don't exceed 100% and violate their operational limits, however all this does is buy 15-20 seconds of time for the throttleman to lower steam flow. By pushing rods in, power does go down, but average reactor coolant temperature drops too, and the steam draw from the steam generators stays the same, so colder water would go in the reactor pushing power back up above 100%. What a good throttleman and RO would do, is spin up to 95% reactor power quickly, then a little more slowly raise to 99.5-100%. But the moral of this story is that rods do not control reactor power while steaming, only average temperature.

As for boron, typically boron controls is all in manual, and rods are in automatic to arrest a quickly increasing trend due to a transient such as a feed pump trip and turbine runback. If a runback occurred and steam load rapidly dropped to 50%, all the sudden your average RCS temp will be way above T_ref, and rods would step in using automatic control to maintain T_ave near T_ref. Boron would then be added to drop T_ave to T_ref, which would cause control rods to step out again.

As for nucleate boiling! We typically say there is no boiling at all in a PWR, however that's not completely true. Directly around the fuel, there are very small steam bubbles that form on the fuel rods and break off into the coolant stream. Once they break off of the fuel, they rapidly collapse back into a liquid. This is called nucleate boiling, because the steam bubbles are formed at small nucleation sites on the fuel (imperfections). Nucleate boiling greatly improves heat transfer efficiency, up until the fuel starts to reach critical heat flux.

Luchekv
An excellent post, thank you HiddenCamper

I see what you mean by BWR's being easier to model...there's a lot of tail chasing in the PWR.

Earlier you stated that:
"[BWR] The turbine operates in pressure control mode. It will automatically modulate how much steam it draws from the core to maintain reactor pressure at a near constant point. So if reactor power goes down, pressure will drop, and the turbine throttles will go shut to maintain pressure."

I wanted to ask, the factors you mentioned in the examples in your last post such as: [PWR]
- Colder inlet temps = increases in reactor power (Why is this?)
- T_ave/T_ref control
- "Average reactor temperature drops slightly, as a result of this power increase"
- "Now I need to raise average temperature to maintain it equal to reference temperature"

Do these occur in a BWR or are they strictly characteristics of a PWR? or are they automatically dealt with due to the automatic pressure control of the turbine?

I ask because of this post:
"For a BWR, rods are used for both local flux control, and global power control. Typically, we set a rod pattern around 45-50% and then we use core recirculation flow to raise power from 50% to 100%. Control rods are rarely used to maintain power above 50% in a BWR, and are only notched out for compensating against fuel depletion.

My BWR originally had automatic flux control. The control system would vary the re-circulation flow through the core automatically to maintain a desired power level. The rods are all manual. We ended up removing automatic flux control due to calibration issues and some weird responses. So now we operate entirely in manual flow control and manual rods."

What do you mean by "rod pattern" and "core re-circulation flow"
So to recap BWR operation, you would use the control rods until 50% after which the re-circulation flow would take over? and the re-circulation flow would be getting pressure feedback from the turbine so it can adjust itself to increase reactivity I assume thus lowering/increasing reactor pressure??
- Is the pressure that the turbine requires a set number or does it fluctuate?
- Why is reactor pressure more important than delivering pressure to the turbine?
- What happens if the reactor is completely depressurized?

I just want to clarify a few other things, when you say global power control. I assume you mean the power output at the generators?
So what would core power be in relation to then?

A few other questions in regards to BWR's if constant pressure is maintained why would reactor power ever go down? what would be examples of disturbances to the system?

Hope I'm not being a burden.. and hopefully my questions aren't giving you university exam flashbacks haha

Thank you again

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"I wanted to ask, the factors you mentioned in the examples in your last post such as: [PWR]
- Colder inlet temps = increases in reactor power (Why is this?)
- T_ave/T_ref control
- "Average reactor temperature drops slightly, as a result of this power increase"
- "Now I need to raise average temperature to maintain it equal to reference temperature"

Do these occur in a BWR or are they strictly characteristics of a PWR? or are they automatically dealt with due to the automatic pressure control of the turbine?"

These are PWR characteristics. In a BWR at power, the unit is always held at saturation. This means BWR temperature is locked in by pressure in the reactor. Heatup and cooldown are controlled by changing reactor pressure, which isn't normally done during operation. For my BWR, we do not exceed ~10% power until we are at rated pressure. Once we get there, we don't look at temperature again until we come offline, and only monitor pressure to make sure the turbine is responding correctly.

As for colder water making power go up. Colder water removes more heat, and is also more dense, this improves both neutron moderation as well as neutron absorption characteristics in the core, allowing more neutrons to get moderated and more neutrons to return to the fuel to cause fission.

What do you mean by "rod pattern" and "core re-circulation flow"
So to recap BWR operation, you would use the control rods until 50% after which the re-circulation flow would take over? and the re-circulation flow would be getting pressure feedback from the turbine so it can adjust itself to increase reactivity I assume thus lowering/increasing reactor pressure??

So control rods are used to raise power up till around 45-50%. At this point, we start raising up the cooling water flow rate in the reactor. Raising the flow rate causes the steam to leave the reactor more rapidly. The steam bubbles are bad moderators, they don't slow neutrons down. If you push the bubbles out faster, the bubbles spend less time in the core, meaning more liquid water is available for moderation. This causes power to go up.

Normally the turbine is in pressure control mode. In this mode, if pressure starts to go up, the turbine throttles open to draw more steam, and pressure stops going up. This makes more electricity. If pressure starts to go down, the turbine throttles close to draw less steam, and pressure stops going down. So for the reactor, this means if I raise power, the turbine will automatically raise output, and pressure remains nearly constant. Normally recirculation doesn't get any feedback and is in manual control.

In the BWRs with automatic flux control enabled (many/most BWRs in the US don't use this anymore), recirculation will get feedback from the pressure regulator and from the flux monitors to try and hold pressure and power steady. In these plants you can run the turbine in "load set" mode, where you demand X megawatts, and the reactor tries to automatically follow. This is disabled in most plants because advanced core designs put the plant in a mode where you have very tight fuel thermal limits, and manual power control is preferred.

- Is the pressure that the turbine requires a set number or does it fluctuate?

How pressure control mode works for the turbine: You set the "zero power" pressure in the regulator. For example, my plant is 917 PSIG at zero percent power output. Then, as main steam pressure goes up from there, the turbine opens it's throttle valves for pressure control. For my plant, at 100% steam output, the turbine controls pressure 30 PSIG higher than the setpoint and all valves are about full open.

- Why is reactor pressure more important than delivering pressure to the turbine?

For a BWR, if you have the turbine simply draw more steam than is being produced, reactor pressure will drop. Lowering reactor pressure causes water to flash to steam, increasing the steam voids in the core, which further reduces power, and causes pressure to drop even more. The opposite is also true. As a result, unless you have an automatic flux control mode for your recirculation pumps, you can't operate the turbine in Load Set mode.

- What happens if the reactor is completely depressurized?

When you are completely depressurized, shut down cooling will subcool the reactor. This is the only time you really are controlling reactor temperature directly. You line up a heat exchanger directly on the reactor and just keep it cold and subcooled.

Between cold shutdown at about 15% power, reactor power is controlled by rods, and reactor pressure is controlled by using the steam dumps.

I just want to clarify a few other things, when you say global power control. I assume you mean the power output at the generators?
So what would core power be in relation to then?

What I mean with this, is that depending on where the control rod is in the core, it can sometimes change only the neutron flux profile directly around it, or across the entire core. When the rods are in the bottom 1/3rd of the core, they only change the "shape" of the neutron flux, and don't affect total core power much at all. When the rods are 2/3rds or more inserted, moving them will affect total core power.

A few other questions in regards to BWR's if constant pressure is maintained why would reactor power ever go down? what would be examples of disturbances to the system?

Power goes down if you lose recirculation flow, or insert control rods. So if I insert a control rod at power, core power decreases, and the turbine will sense this and throttle the turbine valves shut to hold pressure steady. The reactor now sits at steady state, at a lower power level.

Some disturbances, recirculation pump trip, control rod spurious insertion, loss of feedwater heaters (causes colder water to go into the core, this causes core power to go UP fast). Pressure regulator malfunctions can cause power changes, but typically these happen so fast that you'll scram before you can respond to them. Also, if you have one of your main feedwater pumps trip, the recirculation pumps will rapidly ramp down core cooling flow, voiding the core out and causing a rapid power drop. After the voiding is complete, power is low enough that the one remaining feedwater pump has enough capacity to keep the water level in its normal band.

I like answering these questions, hope this helps!

Luchekv
You have helped immensely!

I have attached a diagram of the reactor setup...don't know exactly how accurate it is, but I have a few more questions in regards to this paragraph:
"So control rods are used to raise power up till around 45-50%. At this point, we start raising up the cooling water flow rate in the reactor. Raising the flow rate causes the steam to leave the reactor more rapidly. The steam bubbles are bad moderators, they don't slow neutrons down. If you push the bubbles out faster, the bubbles spend less time in the core, meaning more liquid water is available for moderation. This causes power to go up."

By "raising the cooling water flow rate" do you mean the re-circulation loop shown in pink/purple in the diagram? or the actual loop leading to the turbine? Obviously there are 2 loops for liquid and they don't interact so how are these compartments set up? How is the water to the turbine boiled if at the same time liquid is cooling the core itself."Normally the turbine is in pressure control mode. In this mode, if pressure starts to go up, the turbine throttles open to draw more steam, and pressure stops going up. This makes more electricity. If pressure starts to go down, the turbine throttles close to draw less steam, and pressure stops going down. So for the reactor, this means if I raise power, the turbine will automatically raise output, and pressure remains nearly constant. Normally recirculation doesn't get any feedback and is in manual control."

You say if you raise power, output increases and pressure stays the same. That makes sense...but how are you increasing that power initially..by what method?

Also how do you calculate that power percentage? and what are typical figures in terms of output Watts/temps/pressures for a nuclear power plant - just so I have something to work with..

Off topic question. The information you have given me is priceless and I'm truly grateful you have given me quite an understanding considering I have come into this knowing nothing...my question is will "Intro to nuclear engineering" books contain this kind of information I ask because I would like to cite it for my report but can't exactly cite a forum.

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By "raising the cooling water flow rate" do you mean the re-circulation loop shown in pink/purple in the diagram? or the actual loop leading to the turbine? Obviously there are 2 loops for liquid and they don't interact so how are these compartments set up? How is the water to the turbine boiled if at the same time liquid is cooling the core itself.

There's only 1 cooling loop. The feedwater and the recirculation loops are all 1 system. The water going to the turbine is water that boiled directly in the core. When feedwater is added to the core, it sprays in the Downcomer region outside of the core shroud. The recirculation pumps draw suction on the downcomer, and provide flow to the jet pumps, which in turn pump water into the bottom head of the core, and up through the core itself. Water boils to steam in the core, and then passes up through the steam separators and dryers. The moisture in the steam is removed, and travels through a path back to the downcomer where it mixes in with the new incoming feedwater. The steam that makes it out is 99.95% or better quality, and goes out to the turbine.

As you boil water in the core, you get a build up of steam bubbles. The steam bubbles are bad at neutron moderation, and are a source of negative reactivity. To control power, I can either push the steam bubbles out faster or slower. If I push the bubbles out faster, power will go up, and if I push the bubbles out slower, power will go down. That's what the recirculation pumps do. If you raise the flow rate through the recirculation pumps, you have a higher flow of water into the core, which pushes steam out faster and raises power.

The core cools itself through boiling. It takes a lot of energy to boil water, so as water goes from liquid to steam, it is absorbing energy from the fuel. There is a point where you are have too much steam and not enough liquid, where the steam starts to act as an insulator and can cause the fuel to overheat. This is called "Transition boiling" and needs to be avoided at all costs. The core monitoring computer will calculate the Critical Power where this transition boiling occurs. Operators have a safety limit, called "Minimum Critical Power Ratio", which ensures you do not ever go into this transition boiling region.

You say if you raise power, output increases and pressure stays the same. That makes sense...but how are you increasing that power initially..by what method?

Raise cooling flow with the recirculation pumps, or pull control rods in the upper 1/3rd of the core. Either method will raise reactor power, which generates more steam and causes pressure to increase. The turbine senses the pressure increase and opens up its throttle valves to accommodate the extra steam being generated.

Also how do you calculate that power percentage? and what are typical figures in terms of output Watts/temps/pressures for a nuclear power plant - just so I have something to work with..

For a BWR, full power is based on the license, but let's just say a generic large BWR/5 plant is around 3300 MW thermal energy (~1000 MW electric output). A reactor of this size will have around 740 fuel bundles, operate around 1025 PSIG pressure in the core at 100% power, and temperature in the core will be saturation temperature using a steam table (about 545 degF).

Off topic question. The information you have given me is priceless and I'm truly grateful you have given me quite an understanding considering I have come into this knowing nothing...my question is will "Intro to nuclear engineering" books contain this kind of information I ask because I would like to cite it for my report but can't exactly cite a forum.

Take a look at the BWR/6 plant description document http://www4.ncsu.edu/~doster/NE405/Manuals/BWR6GeneralDescription.pdf

This describes how the plant is controlled, all the safety systems, fuel design, basic plant operation, etc.
Page 6-3 describes how rods and flow are used to control power. Figure 6-2 shows how the automatic flux controller is designed to work (although I don't think any US plants use it anymore).

Hope this helps!

Luchekv
I was definitely visualizing that incorrectly, I imagined the re-circulation pumps simply sucking water out and then dumping it back in, like the loop in the diagram I uploaded..couldn't see the point of that.. so if I understood that correctly...by the time the feedwater reaches the core it effectively needs that extra 'push' to keep things moving along?

What is the typical flow rate for these re-circ pumps? or in other words what level does the water need to maintain in the core (how many Liters) ? I'm guessing that comes down to the Minimum Critical Power Ratio?

Are re-circ pumps in use during the period of power rising to 45-50%? I should say what's the rate during that stage and what's the rate after it passes 50%

Now I just want to try summarize everything so far to make sure I've got it:
BWR:
Heatup/ Cooldown is controlled by changing pressure
Turbine adjusts the throttle to maintain this pressure
Zero Power Pressure is set ~917 PSIG
~10% power is not exceeded until reactor is at rated pressure
Question: If Zero Power Pressure is 917...and pressure at 100% power is 1025...what is rated pressure? and why can't we exceed 10% until at that pressure?
Control rods are used to to get core power to 45-50% after which re-circulation pumps take over to move steam out of the reactor quicker as steam bubbles are bad moderators, a source of negative reactivity and power goes down.

You mentioned earlier: "The steam bubbles are bad moderators, they don't slow neutrons down."
But doesn't slowing down neutrons slow down the reaction? So less thermal power?
What I mean is, keeping that steam bubble would increase heat as the neutrons wouldn't be slowed?

Is it possible to "flood" the reactor?

..I think I'm going to have to move to America and sign up to a nuclear engineering course

Also that BWR/6 plant description document is perfect, thank you very much! Truly appreciate your help and time.

Luchekv said:
..I think I'm going to have to move to America and sign up to a nuclear engineering course

You might get "hooked" as i did. I took a course in reactor operation and got to do startups & irradiation runs on the school reactor.
5 credit hours of Reactor Physics was a prerequisite. Highlight of my college years.
http://nuclear.mst.edu/research/reactor/

Nuclear plant maintenance was a fun career for me. If you like big machinery , look into an introductory course.
The reactor itself is a small part of the plant.
A freshly graduated mechanical or electrical engineer who knew his way around a reactor was in 1969 somewhat unusual .
I found myself doing a lot of interface between nuclear and non nuclear disciplines.
i assume the same is still true.

Good luck !

old jim

Luchekv
Luchekv said:
You mentioned earlier: "The steam bubbles are bad moderators, they don't slow neutrons down."
But doesn't slowing down neutrons slow down the reaction? So less thermal power?
What I mean is, keeping that steam bubble would increase heat as the neutrons wouldn't be slowed?

Is it possible to "flood" the reactor?
Pressurized water (liquid) moderates the neutrons from fission energies ( in MeV range) to thermal energies (0.02 - 0.03 eV range) where they are more likely to be absorbed by U-235 and Pu-239 and result in fission. Some fraction of the core under goes fast fission. In BWRs, the coolant in the core boils. The bottom 1/3 of the core is liquid. Boiling starts in the lower third of the core and bulk boiling happens in the upper two thirds of the core with the steam fraction increasing with elevation. Flow can be reduced, which causes more boiling, or increased, which causes less boiling, and that can be used to control the power in the core. Some BWRs do spectral shift, which means reducing flow to harden the spectrum in the core, which produces more Pu-239 during operation. Later in the cycle, the flow is increased to push the moderation up the core to utilized the Pu-239/-241 for fission.

Slowing down fast neutrons (moderation) is inherent in LWR systems. It takes on the order of fractions of a milli-second to thermalize a neutron from fast to thermal energies. Of course, there are so-called 'delayed neutrons', which are emitted from certain fission products. The delay allows for control of power changes, particularly power increases.

The bulk of thermal energy in a nuclear reactor comes from the fission (~170 MeV) of the 193-205 MeV released from a fission reaction. The other portion of the energy comes from gamma emission and beta decay from fission products. The neutrons slowing down contribute about 5 MeV.

Luchekv said:
Are re-circ pumps in use during the period of power rising to 45-50%? I should say what's the rate during that stage and what's the rate after it passes 50%
Recirculation pumps are operating while the reactor is producing power.

Early GE BWR models (BWR 2/3/4) used variable speed pumps, while later models (BWR 5/6) used two-speed pumps and flow control valves. The ABWR uses more than two pumps with in-vessel impellers.

See Figure 7.2-8 Power/Flow Map

Luchekv
by the time the feedwater reaches the core it effectively needs that extra 'push' to keep things moving along?

Exactly right. The recirculation pumps push water through the reactor core.

What is the typical flow rate for these re-circ pumps? or in other words what level does the water need to maintain in the core (how many Liters) ? I'm guessing that comes down to the Minimum Critical Power Ratio?

There is a "Power to flow" map, which shows allowable combinations of core flow for a given power level. For a large (3300 MWth) reactor, the core flow is between 85-100 Million Pounds of water /hr.

Are re-circ pumps in use during the period of power rising to 45-50%? I should say what's the rate during that stage and what's the rate after it passes 50%

They can be. It depends on how the plant starts up and the power to flow profile. My plant has 2 speed recirculation pumps, and we use flow control valves. For us, around 35% power we shift the pumps to fast speed, and we use flow to raise power as high as we can. We don't like running in fast speed at low flow, because it puts more wear on the seals. For the plants with variable speed recirculation pumps, they will usually raise power higher before raising flow. It all depends.

Heatup/ Cooldown is controlled by changing pressure - Correct
Turbine adjusts the throttle to maintain this pressure - Correct
Zero Power Pressure is set ~917 PSIG - This is for my plant. Generally it's in the 900-950 PSIG range
~10% power is not exceeded until reactor is at rated pressure - For a typical BWR, you need to be above about 850 PSIG. The reason for this, is during a startup, the reactor mode switch is in the STARTUP position. In this mode, the low steam pressure trips are disabled. If you are < 850 PSIG and put the reactor mode switch in RUN, it will enable the low steam pressure trips, shut the steam lines, and scram the reactor. The other limit, is with the mode switch in STARTUP, at 12% reactor power the control rod withdrawls are disabled by the rod control system, and at 15% power the reactor scrams. So you go to about 10%, and > 850 PSIG, then you can place the mode switch in RUN, which changes the safety system logic for full power operation.

But doesn't slowing down neutrons slow down the reaction? So less thermal power?
What I mean is, keeping that steam bubble would increase heat as the neutrons wouldn't be slowed?

It does increase heat, but the reactor power drops off faster than heat is generated, so you never have a challenge to core cooling. One of the accidents that is analyzed is if a recirculation pump goes from full speed to 0 RPM instantly (it seizes). This would cause massive voiding of the core, core temperature would increase, but power would drop very rapidly due to the voiding, faster than the temperature increase. Eventually the power drops low enough that the voids go away, and the reactor returns to critical at a lower power level.

Is it possible to "flood" the reactor?

It is! And it's very bad, if you get water down the steamlines, they will rapidly cool and can be damaged. Also if a safety valve lifts with hot water in it, you may damage it. BWRs have a "Level 8" High-High water level trip. For all BWRs, this will trip off all high pressure injection systems, and also trip off all turbine drives in the plant, to prevent water intrusion. The "newer" BWRs will also have a reactor scram on Level 8, to ensure you don't overpower the core with coldwater injection.

Just a side note: We have an emergency procedure for reactor flooding to the steam lines in the event we lose all water level indication. If you don't know where water level is, a backup way of ensuring the core is covered is to depressurize the core, then flood to the steam lines and watch water pass through the steam lines. While this can potentially cause wear and tear or damage, it is better than accidentally uncovering the core.

Luchekv
Hiddencamper said:
900-950 PSIG range
I think that is at the HP turbine inlet.

I believe the core inlet is more like 1045-1055 psia (~72 - 72.7 bar), and there is about a 20-25 psid pressure drop across the core depending on the lattice design, including spacer grids and part length fuel rods.

Astronuc said:
I think that is at the HP turbine inlet.

I believe the core inlet is more like 1045-1055 psia (~72 - 72.7 bar), and there is about a 20-25 psid pressure drop across the core depending on the lattice design, including spacer grids and part length fuel rods.

That's correct. You set the pressure regulator at the main steam equalizing header, and it controls main steam system pressure. The pressure in the reactor is going to by MS header pressure plus the throttling losses from the reactor to the steam lines.

My BWR steam dome is 1025 PSIG.

Hiddencamper said:
That's correct. You set the pressure regulator at the main steam equalizing header, and it controls main steam system pressure. The pressure in the reactor is going to by MS header pressure plus the throttling losses from the reactor to the steam lines.

My BWR steam dome is 1025 PSIG.
That would be consistent with a core inlet pressure of ~1055 psia.

One of my old textbooks has the steam dome at 1040 psia (1025 psig) and the core inlet at 1075 psia (1060 psig) for a BWR/6, but I've used that for BWR/4s and /5s, most of which have been uprated. Core flow is about 105 E6 lbm/hr.

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Thank you Jim for the link, I will definitely look into it! and Astronuc for the documents!

Hiddencamper said:
The other limit, is with the mode switch in STARTUP, at 12% reactor power the control rod withdrawls are disabled by the rod control system, and at 15% power the reactor scrams. So you go to about 10%, and > 850 PSIG, then you can place the mode switch in RUN, which changes the safety system logic for full power operation.

Quick question about this limit, with the reactor being at 12% I'm assuming it's still under pressure? Why would control rod withdrawals be disabled? and then scram at 15%? Also what exactly happens during a scram?

Do any of your books show the mathematical relationship between power and pressure or even heat?
Just trying to figure out how I would program this simulation..
Once zero power pressure is hit...what is the approx power that the turbine is producing at that exact point in time(at 1% power?) and I'm guessing the throttles wouldn't open to 100% straight away. I guess I'd need that mathematical relationship as well..

How long does it take for a reactor to reach rated pressure from start up?

Thank you again, everyone.

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Quick question about this limit, with the reactor being at 12% I'm assuming it's still under pressure? Why would control rod withdrawals be disabled? and then scram at 15%? Also what exactly happens during a scram?

The MCPR safety limit calculations are only applicable when the reactor is pressurized. With the reactor not pressurized, the reactor is required to be < 25% power where MCPR is not a concern. The 12% rod block and 15% scram signals are there to ensure that under no circumstances does power exceed 25% prior to the reactor being pressurized. These interlocks are bypassed when the reactor mode switch is placed in run, and the low steamline pressure trip is then enabled to protect the safety limit.

During a normal scram, all the control rods rapidly insert using hydraulic accumulators within 3 seconds. Reactor power quickly drops to decay heat levels. Reactor pressure drops as the turbine control valves try to follow the change in steam availability, until the generator's reverse power trip shuts down the generator and the turbine. The condenser steam dumps automatically open for pressure control, and hold the unit at the zero load pressure. When the scram happens, the rapid drop in heat causes the steam bubbles in the core to collapse, and water level indication drops drastically, feedwater will rapidly increase to full injection until water is brought back in the normal operating band.

Once zero power pressure is hit...what is the approx power that the turbine is producing at that exact point in time(at 1% power?) and I'm guessing the throttles wouldn't open to 100% straight away. I guess I'd need that mathematical relationship as well..
How long does it take for a reactor to reach rated pressure from start up?

BWRs do a nuclear heatup. That means they bring the reactor critical at 2-5% power and slowly heat up. This is different than PWRs, which cannot pull critical until the reactor is fully hot/pressurized. PWRs use their reactor coolant pumps and decay heat to heat up.

For a BWR, by the time you are at rated pressure, you are 3-5% power, and the condenser steam dumps are throttling to control reactor pressure. You then raise power to about 10%, place the reactor mode switch in RUN, then raise power to about 15% where you can put the turbine in service. At that point, you transfer steam flow from the dumps to the turbine.

The steam dumps open the moment you hit zero power pressure. But they only open as much as necessary to maintain pressure.

As for heatup, there is an ASME code requirement for all reactors that heatup is limited to a 100 degF change in a rolling 1 hour window. Plus on the way up in pressure, you have tests to do and equipment to start up like the reactor feed pumps. It takes about 20 hours from the moment you pull the first control rod, until the generator is online, from a cold shut down.

Luchekv
Thank you again!
I think I'm out of questions for now haha...I shall continue reading the documents provided so far.

You have all been great teachers, especially you Hiddencamper. Truly appreciate your time.

## 1. What is the purpose of a control system in a nuclear power plant?

The control system in a nuclear power plant is responsible for regulating and maintaining the safe operation of the plant. It monitors and controls various aspects including reactor power, temperature, pressure, and water flow to ensure efficient and safe production of electricity.

## 2. How does the control system work in a nuclear power plant?

The control system consists of sensors, controllers, and actuators that work together to monitor and adjust the reactor's conditions. The sensors provide data on various parameters, which is then processed by the controllers to determine the appropriate actions for the actuators to take, such as adjusting the control rods or coolant flow.

## 3. What safety measures are in place for the control system in case of an emergency?

Nuclear power plants have multiple layers of safety measures in place, including redundant control systems, emergency shutdown systems, and backup power supplies. In the event of an emergency, the control system will automatically shut down the reactor and activate emergency cooling systems to prevent a meltdown.

## 4. Can the control system fail in a nuclear power plant?

While highly unlikely, the control system in a nuclear power plant can fail due to various reasons, such as a malfunction or human error. However, nuclear power plants have strict safety protocols and backup systems in place to prevent and mitigate the effects of a control system failure.

## 5. How is the control system in a nuclear power plant regulated?

The control system in a nuclear power plant is regulated by the Nuclear Regulatory Commission (NRC) in the United States and similar regulatory bodies in other countries. These agencies set strict guidelines and conduct regular inspections to ensure the safe operation of nuclear power plants and their control systems.

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