Few questions about Chernobyl

[Marko7]

1. In steam separator, is steam mixing with water, or it is completely separated? if yes, where goes condensate? (in deaerators?)
2. In gas circuit, is there loop, or N-O mixture goes from storage, trough core, filters, to chimney?
3. Is diesel generators supposed to generate electricity to power pumps and other systems, or is used separated diesel motor to rotate pumps?
4. What influence xenon to form in the fuel?
5. What increases thermal power and is there some relation with: temperature of output steam, reactivity, xenon percent...
6. What is operating reactivity margin (ORM) and how to calculate it?
7. Is there any data about generator and steam turbine used in Chernobyl, such as: turbine mass flow, isentropic efficiency, generator efficiency?
8. Is this correct: lower boron control rods decrease void in lower side of reactor, and positive void coefficient will less increase radioactivity.

 

[anorlunda]

Wow, that's a lot of specific questions. To be clear, you mean your questions only in the context of the RBMK-type reactor used in Chernobyl, correct?

Even so, it would take a textbook length reply to answer all those questions in detail. Perhaps someone here can point you to a textbook on RBMK design.

 

[Marko7]

Yes, most of questions are RBMK specific, but steam separators are used in thermal PP and are almost same. Maybe ORM is used somewhere else, i found something but very poor explained. Thermal power and xenon is more often in other NPP-s.
Here are some links where i dig data:
RBMK design
ORM
ORM 2 - go to the page 24

 

[Astronuc]

ORM 2 - go to the page 24

24 in report, or in the pdf?

On page 24/25 of pdf or page 14/15 of the report, "There is one other aspect of the safety importance of the ORM that has generally received too little emphasis. The staff of the Chernobyl Unit 4 reactor apparently believed that as long as the lower limit on ORM (15 equivalent rods) was satisfied, no matter what the actual rod configuration was, the demands of safety were met."

 

[Astronuc]

What influence xenon to form in the fuel?

Is one asking why, or how, Xe forms in the fuel?

Isotopes of Xe are direct fission products, as well as decay products, e.g., ¹³⁴Te -> ¹³⁴I > -¹³⁴Xe -> ¹³⁴Cs. ¹³⁵Xe is of interest because it has a large cross-section for absorbing thermal neutrons. It forms as a fission product and from decay of ¹³⁵I, also a fission product. During steady-state operation, there is an equilibrium concentration in the fuel, and it competes with fissile isotope for thermal neutrons. So it is inherent in any fissile system, but is significant because of its effect on the neutronic aspects of the core.

When power (and flux) decreases, ¹³⁵I decays to ¹³⁵Xe/^135mXe, and as power decreases the ¹³⁵Xe concentration increases. Depending on the system, control rods may have to be withdrawn from the core to compensate for the buildup of ¹³⁵Xe.

 

[Marko7]

24 in pdf. But what happens when reactor is offline, power is almost 0 and xenon concentration will be maximum. That is very negative reactivity, how do they start the reactor (on safe way). If they pull rods, reactivity increases, power rise, what happens with xenon concentration.

 

[Astronuc]

One is asking about Xe-override (¹³⁵Xe). The level of Xe depends on various operating conditions, before the reactor is shutdown. For example, if it is scrammed (shutdown abruptly) from full power, the ¹³⁵Xe concentration would be greater than if the reactor was scrammed from a lower power level.

¹³⁵Xe decays with a half-life of 9.14 hours. When the reactor shuts down, the production of ¹³⁵Xe from fission stops, but it is still produced from the decay of ¹³⁵I, which has a half-life of 6.58 hours. When the reactor is shutdown, one would normally wait some period of time before attempting to start up the reactor.

One would wait until the negative reactivity of ¹³⁵Xe can be overcome by the positive reactivity from removing the control rods (control rods have negative reactivity; positive reactivity comes from the fissile content of the fuel).

Reactivity control in RBMK reactors is largely associated with the control rod infrastructure. This is organised into four groups comprising: a set of shortened absorber rods for axial control (× 24, length ~ 3 m), full-length (~ 5 m) absorbing rods for radial control (× 24), absorbing rods for the automatic control of reactor power (× 139) and emergency rods (× 24). There are 211 in total and all are made of boron carbide. RBMK control rods are distinguished in one way in particular by the graphite displacers (or ‘followers’) that form part of their design. These are designed to remain in the core when the absorbers are withdrawn and act to displace the light-water coolant from the space that would otherwise be left behind. Without these, undesirable neutron absorption on the hydrogen in the water that would fill the space left by the rod would occur and undermine neutron economy.

RBMKs have different groups of control rods. If the reactor is shutdown, to restart, one would pull most control rods, the emergency group and the long group, and the reactor would still not increase in power if the reactivity was still negative. Then smaller groups of control rods are gradually withdrawn as the reactor approaches criticality. At some point, the positive reactivity exceeds the magnitude of negative reactivity and the reactor can begin power ascension.

Prior to the accident at Chernobyl 4, the reactor was in an abnormal, unsafe condition.

 

[Marko7]

Thank you, this will help me so much in my project, but there is more 7 questions remaining. I am making a real time simulation of Chernobiyl (just half of one reactor block), not so complicated as real Chernobyl, but very close. Currently it is only one huge diagram of relations in power plant. I am not going to use 4000 data inputs to calculate ORM, jus the basic elements, but i don't know what elements (input water temp and pressure, rods level, fluid flow, output temp and pressure...?).

 

[anorlunda]

but i don't know what elements (input water temp and pressure, rods level, fluid flow, output temp and pressure...?).

I don't understand. What kind of simulation can you make not knowing the equations or the parameters?

 

[Marko7]

well, I am not making real simulation of Chernobyl - it will be almost impossible because I can't even find specifications of turbogenerators, so i will use formulas that i make (y = a^x with lot of additional variables for translation...) and correct them with constants to make simulation look like real NPP. Depending on variable A and time will be calculated variable B. All relevant variables will be displayed on picture. All of this will be made in python.

 

[Astronuc]

Thank you, this will help me so much in my project, but there is more 7 questions remaining. I am making a real time simulation of Chernobiyl (just half of one reactor block), not so complicated as real Chernobyl, but very close. Currently it is only one huge diagram of relations in power plant. I am not going to use 4000 data inputs to calculate ORM, jus the basic elements, but i don't know what elements (input water temp and pressure, rods level, fluid flow, output temp and pressure...?).

Constructing a simulator of a nuclear power plant is not trivial, and especially if one wishes to simulate a fast transient situation as that of the Chernobyl accident. One would need a module for the core, which would include the reactor neutronics and thermal hydraulics with appropriate feedback. The reactor thermal hydraulics would use the pressure, temperatures and flow rates AND local void fraction, which would influence the neutron energy spectrum and the power distribution in the core. The pressure, temperature and flow rate are used to calculate the rate at which heat is transferred to the coolant, and local pressure and temperature determine the state of coolant (liquid, vapor fractions). RBMKs have a positive void fraction, which means that a void (steam) in the coolant channel will increase reactivity (because less absorption of neutrons by coolant due to less water) and thereby increasing local power.

A simple simulator would use a lumped parameter model, but such models are usually constructed on the basis of more detailed models.

From World Nuclear Association - https://www.world-nuclear.org/infor...-power-reactors/appendices/rbmk-reactors.aspx

A minimum "operating reactivity margin" is specified. This margin is referred to by the Soviets as the equivalent of 30 inserted regulating rods. Control rods are required to be partially inserted into the core during operation to enhance the initial negative reactivity insertion rate on scram. In addition, insertion of control rods reduces the positive void coefficient. The rods are also adjusted to correct spatial power instabilities in the core. (It is our understanding that the procedural requirements for 30 equivalent rods minimum reactivity margin is a means of specifying an overall rod configuration that ensures a certain initial negative reactivity rate on scram.

The ORM is calculated from the worth of the 30 control rods, which depends on the position of the control rods and the power distribution in the core. It is further dependent on the enrichment distribution and burnup in the fuel assemblies in the core, and the available boron in the control rods.

I can certainly answer the other questions.

 

[anorlunda]

well, I am not making real simulation of Chernobyl - it will be almost impossible because I can't even find specifications of turbogenerators, so i will use formulas that i make (y = a^x with lot of additional variables for translation...) and correct them with constants to make simulation look like real NPP.

If that is your ambition, I suggest that you begin with a simulation of something you have day-to-day experience with. Then you can judge its realism yourself.

A bicycle could be a difficult thing to model. Perhaps a bouncing tennis ball to start?

I started my career with nuclear plant simulators, but I didn't do it alone. The first simulator model took roughly 60 man-years to develop.

 

[Astronuc]

1. In steam separator, is steam mixing with water, or it is completely separated? if yes, where goes condensate? (in deaerators?)

The moisture in the steam is separated such that steam with low moisture content is sent to the turbines and the liquid is collected in the condensate, similar to how BWRs collect condensate which is sent back to the core as coolant. The condensate from the steam generator is collected in the condensate from the condenser under the turbines.

See - https://en.wikipedia.org/wiki/RBMK#Cooling_and_steam_circuits

Each reactor at Chernobyl had two turbine trains (and turbo-generator sets). Chernobyl 4 turbine sets were designated 7 and 8.

See scenario of the accident here - http://engineeringfailures.org/chernobyl.html

On April 25, 1986, at 0100 hours, the operators started reducing power to perform a test on turbine-generator number 8, prior to a maintenance outage. The test was intended to determine how long the turbine-generator would continue to supply power near rated voltage for essential equipment as the generator coasted down. Similar experiments had been carried out twice before at Chernobyl. At 1305, turbine-generator number 7 was shut down with the reactor at 50 percent power. Much of the electric power for the plant, including four main circulation pumps and two main feed pumps, was now being provided by turbine-generator number 8.

Note in the accident, that the reactor was taken to very low power and as the xenon poisoning of the reactor continued to increase, the operators withdrew additional rods (more than allowed by procedure) to compensate. This was a very abnormal and unsafe condition.

Power was being supplied by generator 8, which was part of the motivation for the test. When a reactor is shutdown, a plant would normally draw power from another reactor on the same site, or from the grid. Diesel generators are only used when the turbine is disconnected and power is unavailable from other sources. During normal operation, a reactor is using electrical power from what its turbine generates. The difference between gross and net electrical power is what the plant uses for operation.

 

[Marko7]

What i was asking is: in case of complete power loss (no 750KV, 350KV, other reactors), there is diesel generators that generate power for computer, rods control etc. But are main circulatory pumps, in that case, being powered electrically or there is emergency pump powered mechanically by diesel motor?

 

[anorlunda]

What i was asking is: in case of complete power loss (no 750KV, 350KV, other reactors), there is diesel generators that generate power for computer, rods control etc. But are main circulatory pumps, in that case, being powered electrically or there is emergency pump powered mechanically by diesel motor?

There are sometimes pumps driven by dedicated steam turbines, but I never heard of a pump directly driven by a diesel engine. Even on railroads and ships, designers prefer the diesel-electric approach rather than direct drive.

 

[Astronuc]

What i was asking is: in case of complete power loss (no 750KV, 350KV, other reactors), there is diesel generators that generate power for computer, rods control etc. But are main circulatory pumps, in that case, being powered electrically or there is emergency pump powered mechanically by diesel motor?

With the reactor shutdown, the cooling would consist of removing decay heat, and that is accomplished with a residual heat removal system, which would be powered by electricity from external sources, including diesel generators if other sources are not available. The assumes an orderly normal shutdown.

In the case of an accident, the emergency core cooling system would be activated, and that would involved electricity from emergency diesel generators if other sources were not involved. However, in the case of Chernobyl 4 accident,"At 1400, the emergency core cooling system was defeated in accordance with the test program."

As the four circulation pumps coasted down, the lower flow allowed more steam to form in the core. Because of the strongly positive void coefficient, the increased steam content in the core initiated a power rise as the test began. The increased power further increased the steam voids. The automatically controlled regulating rods began inserting but the power continued to rise. An operator pushed the reactor scram button at 0123:40 to scram the remaining shutdown rods but these rods were near the top of the core and power continued to rise. A loud noise was heard from the reactor. Noting that the rods had not fully inserted, an operator de-energized the control rod drives hoping this would facilitate the scram. Two to three seconds later, operators heard a second loud noise. The core reactivity exceeded prompt critical, and by Soviet calculation, the power peak exceeded 100 times nominal full power.

The accident (a reactivity insertion accident) happened so fast that the diesel generators would not have time to respond. Furthermore, 2 to 3 seconds is a long time for a reactivity accident, which in the LWRs is expected to happen with 0.03 - 0.1 seconds, which would be the time for a control rod to be ejected (PWR) or drop (BWR), or longer (hours) in a PWR boron dilution event.

See if one can access this article - https://www.sciencedirect.com/science/article/pii/S002954931930175X
Heat removal from RBMK reactor core using non-regular means

 

[gmax137]

What i was asking is: in case of complete power loss (no 750KV, 350KV, other reactors), there is diesel generators that generate power for computer, rods control etc. But are main circulatory pumps, in that case, being powered electrically or there is emergency pump powered mechanically by diesel motor?

I don't know much about RBMK reactors but the Wiki page indicates the diesel-backed electrical busses can feed the reactor circulating pumps. That the diesel generators would need almost a minute to pick up the load from the circulating pumps was the point of the test that led to the plant's destruction.

I will agree with @anorlunda here; building a simulator model for the RBMK plant (or any other design) is very ambitious. The PWR simulator models were built up from smaller models (a point-kinetics core model, a pressurizer model, a steam generator model, a reactor coolant loop model, etc.). These smaller models were developed and used for specific design purposes -- component sizing and so forth. Each of these "little" models incorporates physical design data (dimensions, pipe sizes and lengths etc.) as well as considerable test data (for example, pressure drop vs. flow rate in the various sections of the coolant loops and fuel assemblies).

Combining the models into an integrated plant simulator is no easy task.

 

[artis]

@Marko7 let me try to answer some of this

1)
As the water is flowing through the core pipes from bottom to top with the help of the circulation pumps it gets heated up , this adds to the flow so at normal operation the water starts to form steam in it (steam voids) the further up you go along a pipe , so at the top where the water exits each pipe it has some percentage of steam in it. Then it moves through pipes to the steam separator, as it gets in there the fraction that is liquid falls on the bottom as it's heavier while the lighter steam void fraction tends to rise up because it's lighter , some of that steam condenses in the separator falling back down and joining the water at the bottom , the rest goes out to the turbine hall where it enters the steam turbine.2)
Although I don't have direct links but I would presume that the gas loop is closed and the gas is recirculated from the reactor core cavity through filters and back again, unwanted gasses that have been created due to the operation of the core are filtered out from this loop and stored at a separate chamber for some days until their short lived radioactivity drops to a safe level and then are allowed to exit through the chimney, that is the main purpose of the stack seen above RBMK units.

Now here is another interesting feature and maybe the reason why the RBMK stack is so big and high, from what I read it is also made as an emergency steam dump in case the external reactor pipework bursts, namely the endless pipes going in and out of the steam separators , so that there doesn't form dangerous overpressure inside the building. Maybe @Astronuc can say something about this, at least this is what I read in the Russian language papers regarding the RBMK.
I think that in the PDF you linked in your post #16 they were considering this method as one of the possible emergency passive core cooling ways, because the RBMK has a lot of individual pipes and the steam separators themselves are very large having a large surface area essentially forming a natural radiator/heat exchanger so they though of using water sprays on to the steam separators to cool them down and let the vapor exit the original pathways to the chimney?3)
at least in RBMK diesels are solely to generate electricity then used to power equipment like circulation pumps etc. as far as I know all NPP have the same way and I haven't heard of diesels powering pump directly.5)
This seemingly simple question is actually complicated but in short, output thermal power or heat capacity is increased when the number of nuclear reactions/neutrons in the reactor within given time period is increased.
Insert more rods = more neutrons get absorbed. more neutrons get absorbed = less fission reactions in one second, core cools down.
This is the reactivity part but coolant temp also has a relationship which actually played a crucial role in the 1986 accident. You see water is also a neutron moderator/absorber so as you get less water in the channels you lose some percentage of neutron absorbing capacity of the core.
The higher the power of the core the hotter the water becomes for any given water flowrate, the more steam is in the water it becomes less dense and absorbs less neutrons which then increase the reactivity.

Here is a catch , graphite is a better, more economical neutron moderator as it reflects more of them and absorbs less than water, so as you decrease water percentage in the core pipes more and more moderation is done by the core graphite so the neutron flux increases , this is the positive void coefficient of the RBMK.7)
For this info i think you would have to search the internet in Russian as most of the data with regards to RBMK is in Russian, unlike the VVER Soviet made reactors the RBMK were not exported outside the USSR so they are solely a Soviet/Russian thing and most of their details are only known by Russian specialists.
How good are you with Russian language ?