Can you give me some advice of this topic?
If you tell us what your thoughts are, you will be more likely to get some feedback.
Are you referring to software or hardware? You'll have to be more specific.
And the reactor type (BWR, PWR, something else?) has a big influence on the very physical mechanisms used for controlling the reactor power. In a BWR, power control is typically achieved by controlling the recirculation flow, whereas a PWR reactor power may be controlled either by active control rod manoeuvers or through physical feedback mechanisms by controlling the turbine power.
What does one know about control theory?
While this is essentially correct, in PWRs core power is often controlled by varying the soluble boron concentration, particularly in base load plants that do not have grey rods for power shaping or power adjustments for load follow or frequency control. Most PWRs in the US do not have power shaping rods. Reducing feedwater temperature or using the turbine is usually performed at EOC.
That's right, of course. I was thinking more of the load follow -type fast power control, where the dilution/boration control is probably too slow.
Regarding turbine control, BWR:s can be made quite simple: the reactor power is controlled by varying the main circulation pump speed based on feedback from the generator power, and the turbine controller just maintains the steam dome pressure, i.e. "turbine follows reactor". How is it in the referred US PWR plants: are the turbines operated without feedback from the generator power ("following the reactor" similarly to the BWR case), or is the turbine power actively adjusted to meet the desired power output, and the primary inlet temperature will then take care of adjusting the reactor power to match the turbine (="reactor follows turbine")?
Look for a book called "Nuclear Reactor Engineering" by Glasstone & Sessonske.
I find the older version more readable, a sorta pink front cover with a power plant not the later one with red&yellow graffiti motif.
He has an excellent chapter on the subject.
Book is often on Ebay.
I want to know how to design a mode "G" reactor control system for PWR .
Is the PWR roughly 157 assemblies (900 MWe), or 193 or 205 assemblies (1300 to 1450 MWe)?
This should give one some ideas.
for three decades I maintained an analog PWR reactor control and protection system that was designed in late 60's. But i am not familiar with the term "Mode G". I would guess it means something we old guys are used to but not by that name.
Were i in your shoes i'd study what the ancients did in 1960's for starters.
If you set out to re-invent the wheel you'll have to stumble up from bottom of learning curve. Why not start from halfway up?
In early 70's we could load follow and with a little luck survive a somewhat greater than 50% load rejection transient. But over the years increasingly stringent conservatism made us basically an 'all rods out' baseload with chem(Boron) shim.
In our plant the basic automatic control made reactor follow turbine. That way the plant could load follow as directed by system dispatch.
Load on turbine is inferred by measuring steam flow through it which gets shifted(mx+b) into a desired reactor temperature.. Measured temperature is subtracted from desired to produce a temperature error signal. Temperature error becomes a contol rod speed&direction signal that sets the rods into motion. The more temperature error the faster rods move. When measured temperature matches desired there's no error anymore so the rods stop. It really is basically that simple.
For better transient response another difference signal is developed, difference between reactor power and turbine power. Rate of change of this difference signal is added to the temperature error signal. It trims rod speed during transients and can help prevent over/under-shoot.
The reactor would not know whether it is being controlled by an old analog system like mine or by a fancy computer system like i assume you'll build. So your algorithms will probably start from old timey control theory basics.
In my day Bailey Controls had excellent technology, some say the best, and they in turn were owned by Babcock&Wilcox.
You'll probably find some good nuclear plant control system theory in the book "Steam its Generation and Use" published by Babcock & Wilcox.. look for a late 1970's edition ( it's been in print since at least 1920's.)
Description of the Mode G control strategy for a VVER 1000 reactor can be found in this article.
Thank you for your explanation about temperature control and mismatch control of PWR and the other information.
MODE G : load follow.
MODE A : basic load.
who have the description of the Mode G control strategy for a Westinghouse AP1000 reactor?
I need it very much!
Thank you very much
I'm not familiar with that term either.
But it makes me wonder about Modes B, C, D, E, and F. Where are these terms coming from?
You may wish to look at their DCD, which is filed with the NRC. The AP1000 grey RCCA (GRCA) uses 4 AIG rodlets (fingers) and 20 Stainless steel rodlets (as filed in the UK).
http://www.nrc.gov/reactors/new-reactors/design-cert/ap1000/dcd/Tier 2/Chapter 4/4-1_r14.pdf
However, in the US design, W indicates 12 AIG rodlets (fingers) and 12 304SS rodlets
AP1000 Design Control Document
CHAPTER 4 REACTOR
4.1 Summary Description
Of course, one needs a core monitoring system and I&C.
See - http://pbadupws.nrc.gov/docs/ML0832/ML083230868.html - for AP1000 DCD Rev. 17
One would look at Chapter 4 (particularly 4.2, 4.3) and 7.
From section 4.2
I have the impression that "mode G" would be the French control strategy developed for the 900 MWe plants in the late 70's to enable better load following capabilites by utilizing gray control rods, hence the name "G" for "Gris" or "Grey".
There's one source (in French): http://www.sfen.org/IMG/pdf/ST6-15mars2007.pdf
or translated by Google to "English".
Other modes listed are "X" for Axial imbalance power control (for the N4 generation of PWRs) and "T" for what I guess is the core average temperature control used in EPR.
PWR's naturally "load follow" or "follow the turbine". The power of a nuclear power plant with
a PWR is controlled by the turbine throttle valve.
Suppose a factory fires up and there's a big increase in the load on the plant. That would tend to draw more current from the plant, and the additional back-EMF of the generator would tend to slow it down. The generator has to remain phased to or sync'ed to the grid, so a controller on the generator opens the turbine throttle. This increases turbine power and it draws more energy from the primary coolant. The primary coolant goes back to the reactor "cooler", and the moderator temperature feedback forces an increase in reactor power until it balances the additional power demanded by the turbine.
The PWR has a fairly simple controller on the generator that controls the turbine throttle and the reactor just naturally follows.
Yes, that is the "control by physical feedback mechanisms" I mentioned as one alternative in my first post:
I know this "reactor follows turbine" by physical feedbacks is the method used in VVER 440 -type PWR:s operating on base load, whereas the EPR is designed to actively alter the reactor power by control rods in order to keep the reactor average temperature constant and thus minimize thermal transients. Reading Astronuc's answer to my post, correcting that the power in US PWRs would be controlled by chemical shim, I wondered if there would be yet some other control strategy in addition to those two. But apparently they use this "reactor follows turbine by physical feedback" approach, then.
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