How Are BWR Fuel Assemblies Managed and Composed?

[girts]

Hi, now I watched a video that I knew before but this time I kind of thought I want to know for sure. The video is about fuel change in a BWR type reactor, now I have question to someone who knows about this or maybe has worked with it, are the rectangular fuel assemblies seen in the video fixed while working with some additional means or do they simply slide in with the help of the crane that lifts them out and back in and simply stay there without any additional support?
What I mean is do they simply drop and push in the rectangular fuel bundle and leave it at that and that's it? Then put the lid back on the vessel and take out the control rods and voula, the reactor is critical again?
here is the link to the BWR refueling video in question



are the rectangular fuel rod assemblies made of Zr or just the individual fuel tubes located inside the rectangular fuel assembly? If so then what material is used for the rectangular fuel rod assembly?

I assume the active core made up of the fuel assemblies is located in the lower part of the vertical vessel in order to limit water cavitation under boiling alongside the fuel assemblies, as the higher up the water becomes less dense while boiling correct?

And after they remove the vessel lid for refueling they flood the reactor compartment with water so that the water rises higher than it normally would in order to have biological shielding as they move the old fuel assemblies out and transfer them next to the reactor vessel where the spent fuel pool is located?http://www.neimagazine.com/uploads/newsarticle/993055/images/199174/large/1-atrium.jpg

is this the fuel rod assembly once taken out of it's rectangular encapsulation that it has inside when inserted inside the core?
thanks

 

[Astronuc]

are the rectangular fuel assemblies seen in the video fixed while working with some additional means or do they simply slide in with the help of the crane that lifts them out and back in and simply stay there without any additional support?
What I mean is do they simply drop and push in the rectangular fuel bundle and leave it at that and that's it? Then put the lid back on the vessel and take out the control rods and voula, the reactor is critical again?

BWR fuel assemblies are inserted into a reactor and retrieved by the fuel handling machine. The telescoping crane lifts an assembly, or lowers it. The assembly is not pushed, but rather set in place. Four assemblies form a cell, and in the middle of most cells sits a control blade. In a refueling outage, the highest burnup assemblies are discharged from the core. Assemblies to continue service may remain in the core, with many or most being shuffled, that is relocated, and fresh fuel is added. Some assemblies may only serve for two 24-month cycles, and some for three cycles.

are the rectangular fuel rod assemblies made of Zr or just the individual fuel tubes located inside the rectangular fuel assembly? If so then what material is used for the rectangular fuel rod assembly?

The rectangular fuel channels, with round corners, have been traditionally Zircaloy-2 or Zircaloy-4, in the annealed state. Channels composed of a Zr-Sn-Nb-Fe alloy (GNF's NSF and Westinghouse ZIRLO) have been introduced relatively recently. Fuel rod cladding is Zircaloy-2, although some modern alloys are derivatives of Zircaloy-2 with additional Fe above ASTM limits for Zircaloy.

I assume the active core made up of the fuel assemblies is located in the lower part of the vertical vessel in order to limit water cavitation under boiling alongside the fuel assemblies, as the higher up the water becomes less dense while boiling correct?

No, the core is about the middle of the reactor pressure vessel (RPV). The core sits on a core support plate, control rod drives sit below the core, and during operation, most control blades are withdrawn below the core. Above the core is the equipment that separates steam from liquid, removing water droplets, and thus 'drying' the steam which is sent through pipes to the high pressure turbine.

And after they remove the vessel lid for refueling they flood the reactor compartment with water so that the water rises higher than it normally would in order to have biological shielding as they move the old fuel assemblies out and transfer them next to the reactor vessel where the spent fuel pool is located?

The reactor cavity is flooded before the vessel head is removed. The level of the water matches the level of the spent fuel pool. The water shields the workers, who move the fuel, and others working on the refueling floor. The spent fuel pool is not in a reactor vessel, but is located in the secondary containment area beside the reactor vessel.
See - http://files.gereports.com/wp-content/uploads/2011/10/NEI-Mark-1-White-Paper.pdf

is this the fuel rod assembly once taken out of it's rectangular encapsulation that it has inside when inserted inside the core?

The image is one of a BWR assembly before the channel is installed, which usually happens at the reactor site, although some modern designs have the channels installed at the manufacturing facility. Traditionally, BWR fuel assemblies are shipped without channels to avoid damage to the spacer grids, which separate the fuel rods, and hold them in a regular array. Fuel assemblies are shipped horizontally in special containers, with special inserts to support the fuel rods during shipment. The inserts are removed during fuel inspection receipt, and the channels are subsequently installed.

 

[anorlunda]

@girts , it is not entirely clear what you are asking about.

The link http://www.neimagazine.com/uploads/newsarticle/993055/images/199174/large/1-atrium.jpg is a picture of a fuel bundle.

Between the bundles, we also insert control rods:

A BWR core with no fuel bundles and no control rods inserted is just a metal structures that create specific places (channels?? not sure just which word to use) for the bundles and rods to fit. Those structures guide, and they define exactly how far apart the bundles and rods are from each other. I tried but failed to find a picture of a core with no fuel and no rods inserted.

Is that what you were asking about?

 

[girts]

So I take that as a yes, that in a BWR reactor the rectangular fuel bundles with slightly rounded corners to be exact are not fixed in place by anything other than the very support structure mounted inside an empty BWR vessel that houses the fuel assemblies and also acts as a spacer so that each fuel assembly is an even distance apart from each next one right?

Now I was asking about whether it takes a bit of force to push in the new bundles because if you watch at the end of the video you can see two bundles bolted together by a rail being lifted together from the fuel pool then dragged to the core and put in, but if you look closely you can see that at the very end when the bundles are almost inserted the operator kind of jerks the crane multiple times in order to get the fuel assemblies in, it almost feels like there is some resistance and they would not want to go all the way in simply from their weight under gravity alone.
Yes I know that the spent fuel pool is located next to the reactor pressure vessel and separated by a concrete wall but under the same silo type containment building that the vessel is at, right?Seems that the difference in containment between a PWR and a BWR is that the BWR has the reinforced concrete shell right around the pressure vessel while the rectangular reactor building itself is simply industrial and not specifically designed to whitstand extra pressures or other overloads, while the PWR silo type reactor building has its very shell made of thick reinforced concrete and the reactor vessel with I suppose the spent fuel pool is inside the silo, so from the viewpoint of an external attack or inner hydrogen explosion the PWR seems more safe as in the BWR the spent fuel pools are uncovered to atmosphere if the roof is blown away like in Fukushima?

I assume that in a refuel they take the rectangular bundles that were at the outermost periphery and insert them in the middle of the core and take the middle fuel bundles and discard them in the spent fuel pool, because in the middle is the highest burnup rate? Then the fresh fuel is firstly introduced in the periphery of the core and only after it has spent its time in the outer locations it is then relocated in the center?

Yes anorlunda I imagine that an empty BWR vessel just from factory is just an empty cylindrical vessel with a rectangular spacer grid for the rectangular fuel assemblies in the middle of the vessel and a support plate underneath and probably some control rod drives and mechanism at the bottom and holes for fresh water supply at the bottom and steam exit at the top.

I wonder why they have control rods coming from bottom, it kind of feels more intuitive to have them at the top and if all power is lost they can fall in by gravity to stop the reaction.
Oh one more thing, well if they install the rectangular hollow tube for the fuel rods at the reactor site once the fuel rods arrive then once they have served their full burn up lifetime they then let them "cool" in the spent fuel pond for some years and then ship to fuel recycling or storage together with the rectangular tubes or do they take them apart and reuse the rectangular Zircaloy tubes? I assume it depends on whether they have become radioactive or not but given the fact that the core is the highest neutron flux area they must be radioactive after they have been there?thanks

 

[anorlunda]

You're kind of rambling a bit. It is best to try to make one thread stick to one question more or less. But to answer one of your questions:

I wonder why they have control rods coming from bottom, it kind of feels more intuitive to have them at the top and if all power is lost they can fall in by gravity to stop the reaction.

Above the core in a BWR, are steam dryers. They reduce the moisture content of the steam. They also make it impossible to make the rods come down from the top. In a PWR, steam drying can be done in the steam generators away from the reactor vessel.

 

[sandy stone]

If you don't mind following up on that, surely there must be some kind of mechanism to return the control rods to the fail-safe position on the event of control system failure. What would that be in this case, if the control rods must be lifted against gravity?

 

[Astronuc]

If you don't mind following up on that, surely there must be some kind of mechanism to return the control rods to the fail-safe position on the event of control system failure. What would that be in this case, if the control rods must be lifted against gravity?

BWR control rods are inserted hydraulically. There is a locking mechanism that locks the control rod in the desired position, including full insertion.

 

[anorlunda]

If you don't mind following up on that, surely there must be some kind of mechanism to return the control rods to the fail-safe position on the event of control system failure. What would that be in this case, if the control rods must be lifted against gravity?

Safety
In most reactor designs, as a safety measure, control rods are attached to the lifting machinery by electromagnets, rather than direct mechanical linkage. This means that in the event of power failure, or if manually invoked due to failure of the lifting machinery, the control rods fall automatically, under gravity, all the way into the pile to stop the reaction. A notable exception to this fail-safe mode of operation is the BWR, which requires hydraulic insertion in the event of an emergency shut-down, using water from a special tank under high pressure. Quickly shutting down a reactor in this way is called scramming.

You should do a little effort to research questions yourself before posting here. Wikipedia is frequently the easiest source.

 

[Astronuc]

Now I was asking about whether it takes a bit of force to push in the new bundles because if you watch at the end of the video you can see two bundles bolted together by a rail being lifted together from the fuel pool then dragged to the core and put in, but if you look closely you can see that at the very end when the bundles are almost inserted the operator kind of jerks the crane multiple times in order to get the fuel assemblies in, it almost feels like there is some resistance and they would not want to go all the way in simply from their weight under gravity alone.

The 'two bundles bolted together' are not fuel bundles/assemblies. They are simply tubes to take up space and provide lateral support to the assembly being retrieve from a location where two face adjacent assemblies have been removed.

There is an upper guide structure that is removed for refueling, so assemblies need some support in addition to the other assemblies.

Channel reuse was considered in order to save money. However, the neutron flux in a BWR fuel assembly has a gradient, so one side of the channel may grow more than the other. The differential growth causes a channel to bow (distortion) with consequences of interference with a control blade, and possible impact on the moderation around a given fuel rod (some rods over-moderated and others, on the opposite side of the assembly, under-moderated).

I'll address other questions later.

 

[gmax137]

You're kind of rambling a bit.

So we can take turns answering the parts we like to talk about

...while the PWR silo type reactor building has its very shell made of thick reinforced concrete and the reactor vessel with I suppose the spent fuel pool is inside the silo,

Designs differ, but the PWR concrete "silo" you can see from the outside has a steel containment vessel inside it. The concrete part may play a structural role, or a shielding role, or both. Sometimes there is a space between the steel wall and the inside of the concrete, with fans to force the in between air through charcoal and filters. Sometimes the space acts as a chimney to draw outside air up around the steel, for cooling.

Also, the PWR spent fuel pool is outside the containment. The spent fuel bundles are moved out through a "transfer tube" that is closed during plant power operation. When the plant is running, the fuel inside containment is all in the reactor vessel, in the core.

 

[gmax137]

I assume that in a refuel they take the rectangular bundles that were at the outermost periphery and insert them in the middle of the core and take the middle fuel bundles and discard them in the spent fuel pool, because in the middle is the highest burnup rate? Then the fresh fuel is firstly introduced in the periphery of the core and only after it has spent its time in the outer locations it is then relocated in the center?

The core designers consider different refueling patterns and try to come up with the scheme that meets the core power requirements (meaning, will run at power until the next refueling) for the least cost. It turns out to be a little different for each refueling. Fuel cost for a nuclear unit is very low (compared to gas or coal) but even so, a good core design can save millions of dollars over a bad design. So it is worth the time for the designers to figure it out.

 

[sandy stone]

You should do a little effort to research questions yourself before posting here. Wikipedia is frequently the easiest source.

I suppose you're right. It was a spur-of-the-moment thing.
That being said, a safety device that requires positive mechanical action (hydraulic insertion), as opposed to passive (fall when released), seems a little less robust.

 

[gmax137]

... as opposed to passive ...

I'm not a BWR guy, but my understanding is that the hydraulic accumulators are normally pressurized. So the motive force is available. BTW, are they really hydraulic? I thought it was compressed air. Maybe @Hiddencamper can tell us what happens on decreasing accumulator pressure.

Neither system (gravity nor hydraulic) is really passive; something has to change state to release the grip and allow insertion.

The PWRs scram by de-energizing the holding coils. This is via relays, or loss of power to the M-G sets. That same degree of "passivity" on loss of power could be designed into the hydraulics (e.g., pressure holds the valves closed against springs).

My biggest philosophical issue with the BWR system is all the penetrations in the vessel lower head. Of course, many or even most PWRs have bottom-mounted instruments, so the argument doesn't necessarily hold. But I do like the designs that have the nice smooth continuous pressure boundary on the bottom.

 

[Hiddencamper]

If you don't mind following up on that, surely there must be some kind of mechanism to return the control rods to the fail-safe position on the event of control system failure. What would that be in this case, if the control rods must be lifted against gravity?

The control rods in a BWR are hydraulic.

There are three pressurized water sources to insert the control rod. The first is every rod has a charged hydraulic accumulator which uses high pressure water to insert the rod in 1-3 seconds.

The second is the reactor’s own pressurized water supply. There is a ball check valve that shuttles if the scram exhaust valve opens and the supply pressure isn’t sufficient to drive the rod in. The rods generally insert within 7 seconds in this case.

The third is the control rod hydraulic pumps. They will insert the rod even if you had a failed accumulator and failed check valve. You can also manually drive the rod in with the rod drive controls using the drive pumps.

If that all fails you have boron injection and the ATWS/ARI system which is an independent scram system that also trips the reactor recirculation pumps to lower core flow to drop core power and preserve vessel integrity. Afterwards the operators will terminate all injection to the core to eliminate subcooling, and lower water level to reduce natural circulation, until power is less than 5% or you are 2 feet below the feedwater spargers and the containment is not challenged. Then you wait until boron is injected and you can then start cooling down.

 

[Hiddencamper]

I suppose you're right. It was a spur-of-the-moment thing.
That being said, a safety device that requires positive mechanical action (hydraulic insertion), as opposed to passive (fall when released), seems a little less robust.

In both cases you are using prestored energy.

For a pwr, when a scram signal is initiated, your scram contactors and trip breakers need to open. This removes energy from the grippers. The grippers need to release (spring force), and prestored energy (gravity) drives the rods into the core.

For a bwr, when the scram occurs, the scram trip relays open which deenergize the scram pilot solenoids on each control rod. With no power, the solenoids reposition to the “block and bleed” position to rapidly depressurize the scram pilot air header. With no air pressure the scram inlet and outlet air operated valves rapidly open under spring pressure. Finally pressurized water from your accumulator goes into the underpiston (and overpiston water is exhausted to the discharge volume, or for newer designs to the reactor itself). If an accumulator is inoperable, a ball check valve repositions and reactor pressure drives the rod in using its own pressurized water supply. If reactor pressure isn’t sufficient then the rod drive pumps complete the scram.

In both causes you have a system where prestored energy is trying to insert the rod, and we hold the rods out using electricity and other forces. Losing power causes all of those things to reposition and that prestored energy puts the rod in. The only real weakness of the BWR design is that you can get hydraulically locked on the discharge header (see Browns ferry), and this is mitigated by level sensors in the discharge volume that generate reactor trip signals.

 

[sandy stone]

Thank you @Hiddencamper for the detailed response. I'm not sure I could have found that much information from an internet search. I used to work in an oil refinery, and I'm interested to see the level of redundancy built into the safety systems.

 

[Hiddencamper]

Thank you @Hiddencamper for the detailed response. I'm not sure I could have found that much information from an internet search. I used to work in an oil refinery, and I'm interested to see the level of redundancy built into the safety systems.

No problem!

I operate one of these after all.

 

[girts]

@Hiddencamper, could you please take a look at my post #4 and maybe answer some of the questions from your perspective?thanks

 

[Hiddencamper]

Question: So I take that as a yes, that in a BWR reactor the rectangular fuel bundles with slightly rounded corners to be exact are not fixed in place by anything other than the very support structure mounted inside an empty BWR vessel that houses the fuel assemblies and also acts as a spacer so that each fuel assembly is an even distance apart from each next one right?

Now I was asking about whether it takes a bit of force to push in the new bundles because if you watch at the end of the video you can see two bundles bolted together by a rail being lifted together from the fuel pool then dragged to the core and put in, but if you look closely you can see that at the very end when the bundles are almost inserted the operator kind of jerks the crane multiple times in order to get the fuel assemblies in, it almost feels like there is some resistance and they would not want to go all the way in simply from their weight under gravity alone.

The fuel bundles weight 1000 LBs or more. Their weight acts as the seating force. They have an orificed lower piece that is circular which slots into a hole in the core plate. There is an upper support plate that really only acts for positioning. Sometimes you need to move things around a little to get them to settle into place, but the bundle's weight is the seating force. The fuel bundles themselves act as the lateral support for the cruciform control rod blades, which have rollers on them so that they can lean on the fuel rods while inserting without impacting them. When we go to defuel a cell, we remove 2 fuel bundles, then place a dummy piece called a "Blade guide" in that provides lateral support for the control rod in that cell, then we pull the other 2 out.

Question: Yes I know that the spent fuel pool is located next to the reactor pressure vessel and separated by a concrete wall but under the same silo type containment building that the vessel is at, right?

Seems that the difference in containment between a PWR and a BWR is that the BWR has the reinforced concrete shell right around the pressure vessel while the rectangular reactor building itself is simply industrial and not specifically designed to whitstand extra pressures or other overloads, while the PWR silo type reactor building has its very shell made of thick reinforced concrete and the reactor vessel with I suppose the spent fuel pool is inside the silo, so from the viewpoint of an external attack or inner hydrogen explosion the PWR seems more safe as in the BWR the spent fuel pools are uncovered to atmosphere if the roof is blown away like in Fukushima?

Mark I and II BWR containments have a spent fuel pool on the upper floor of the reactor building. The reactor building is the "Secondary Containment" and is the square shaped building that the reactor auxiliaries are in. The secondary containment is not designed for pressure, it's only designed for leakage from the containment and to deal with small steam leaks from interfacing systems. Mark III BWRs do have an upper spent fuel pool, but only for refueling operations. You are not allowed to operate with spent fuel up there. So you have to transfer it to the spent fuel pool in the fuel building using a fuel transfer system. During Accident operation, the Mark III's upper containment pool water becomes a gravity feed to refill the suppression pool during post LOCA conditions.

Both PWR and BWR generally keep their fuel outside of the containment in a separate spent fuel pool. For PWRs and Mark III BWR containments, there may be a temporary storage pool, but it is not the norm to keep spent fuel in containment.

The roof blown away at Fukushima is a design feature. The upper roof is designed to blow out in the event of a steam or hydrogen explosion to prevent catastrophic failure of the building's structure and a collapse which could cause significant radiation release to the environment and local personnel. They did their job at Fukushima as far as I'm concerned. Neither plant is more safe than the other for spent fuel. However, BWR spent fuel has less power density, so it has less decay heat and becomes passively coolable far earlier than PWR fuel. Studies have shown that BWR fuel is coolable during total loss of coolant conditions in a few months, while PWR fuel requires a year or more. That said, these are complicated accidents and not all the data has been vetted and released yet (even if some people got to see it at Sandia labs last november during an owners group meeting)Question: I assume that in a refuel they take the rectangular bundles that were at the outermost periphery and insert them in the middle of the core and take the middle fuel bundles and discard them in the spent fuel pool, because in the middle is the highest burnup rate? Then the fresh fuel is firstly introduced in the periphery of the core and only after it has spent its time in the outer locations it is then relocated in the center?

Current core designs for BWRs generally use the "Improved Low Leakage Core Design" or something similar. You use the oldest fuel on the outer periphery, because it has very low LHGR limits (5-6 kw/ft), and power on the periphery is low. This also creates an effect of the periphery acting as a neutron reflector and multiplier for the fuel directly inside the periphery, helping to boost it's power level. The fresh fuel gets scattered throughout different rings. Fresh fuel has lower MCPR and LHGR penalties, so you tend to scatter them in high power portions of the core to help with thermal limits, especially if the bundles are to be loaded with gadolinium for burnup and axial power control. Each fuel bundle and location is specifically chosen for that fuel cycle. Some fuel bundles use part length rods, or have extra water rods or dummy rods, or gadolinium, or have lower peak enrichment. So there's all sorts of factors when you design a core. You typically do not use new fuel in the center, for a number of reasons. Putting fresh fuel in the periphery is a huge waste, especially because the uneven heating of the fuel can cause bowing and other issues (NEXRAT), while providing no support to the rest of the core to maintain power or flux shape.Question: Yes anorlunda I imagine that an empty BWR vessel just from factory is just an empty cylindrical vessel with a rectangular spacer grid for the rectangular fuel assemblies in the middle of the vessel and a support plate underneath and probably some control rod drives and mechanism at the bottom and holes for fresh water supply at the bottom and steam exit at the top.

I wonder why they have control rods coming from bottom, it kind of feels more intuitive to have them at the top and if all power is lost they can fall in by gravity to stop the reaction.

I have other responses on the safety of bottom entry control rods. So in this response I'll talk about the reasons we use bottom entry rods. The first, is that we have the steam separator, dryer, and other steam equipment on the top of the core. The second is it allows us to do some interesting things for flux shaping. Peak power is often at the bottom 1/3rd of the fuel, meaning that during a scram, the rods do not have to travel very far before they suppress peak power. By varying the position of the control rods we can adjust the axial flux shape if necessary, and push the power profile to the top of the core (if we are bottom peaked). Today you typically use gadolinium poisons in the fuel to control shape, so this is more of a post-transient recovery option. Bottom entry rods are just easier to put in. The guide tubes that the rods travel through also act as additional support for core internals.

Question: Oh one more thing, well if they install the rectangular hollow tube for the fuel rods at the reactor site once the fuel rods arrive then once they have served their full burn up lifetime they then let them "cool" in the spent fuel pond for some years and then ship to fuel recycling or storage together with the rectangular tubes or do they take them apart and reuse the rectangular Zircaloy tubes? I assume it depends on whether they have become radioactive or not but given the fact that the core is the highest neutron flux area they must be radioactive after they have been there?

The fuel rods themselves are full of fission products and gaseous radioactive materials. You cannot disassemble them without a reprocessing system designed for it, which currently does not exist in most countries (basically France and Japan and maybe Russia?) The fuel rods are not reused. They are extremely radioactive (although the fission products are more-so).

 

[rpp]

Here is a picture of 4 BWR fuel bundles with a cruciform control blade between them.
The uranium fuel is in zircaloy clad rods, arranged in an approximate 10x10 grid (at least for modern designs).
Each fuel bundle is surrounded by a zircaloy channel box, which you refer to as "rectangular fuel bundles with slightly rounded corners"

http://www.world-nuclear.org/uploadedImages/org/info/GE BWR nuclear fuel assembly 2.jpg

The fuel bundles are inserted into lower nozzles, which sit on top of the lower core plate. The flow moves upwards through the core plate,
through the nozzles, and into the fuel assemblies. The channel boxes are designed to isolate the void/vapor produced by the boiling water.
There is also some flow (called bypass flow) that goes on the outside of the channel boxes.

As was mentioned before, the fuel assemblies are quite heavy. however, there is also a top core plate that is placed on top of the assemblies that stops them from "lifting off" during normal operations.

You might also find this interesting. This is a small book with a description of the GE BWR/6 reactor. There are several diagrams to help you understand how all of the systems go together.
http://www4.ncsu.edu/~doster/NE405/Manuals/BWR6GeneralDescription.pdf

The control blades are cruciform in shape and travel between the fuel bundles. The control blades enter from the bottom of the core because there is no room at the top of the core to put them. The top of the core has the steam separators and steam dryers.

 

[girts]

I am reading the BWR general description brochure and a few question arose.
I was reading about the control rod insert mechanism and the means by which the operators know how far a rod is inserted, the brochure mentioned the use of reed switches, now I do know what a reed switch/herkon is and how it works, the thing that wasn't explicitly stated but I think I understood it is that I think the reed switch is located on a fixed position along the rod so called "index" tube and there are magnets located on the index tube with identical spacing and as the tube moves up or down the magnets connect/disconnect the reed switch contacts and so a computer or maybe in older designs an analog/mechanical device can count the times a magnet has passed and so the operator can know the exact position of the rod? Is this about correct? From what I understood each control rod index tube has a given set of points at which a horizontal electrically activated gear can latch the tube in place so that it doesn't fall back so does that mean that each control rod can only be stopped and held at specific heights which are predetermined by the design of the rod pushing tube? I assume other reactors like the PWR or RBMK and other can stop the control rods at any specific moment as their holding mechanisms differ?
Now as I understand the BWR control rod housing tubes are essentially open to the reactor core and it's water if the rod and pushtube itself would be taken out correct? So are these housing tubes fixed into place by welding them into the lower reactor vessel head?
It was also kind of interesting to read that the piston that pushed the index tube and control rod up or down and its seal is essentially the only thing separating the reactor core water pressure and the water pressure applied to the piston from below by the control rod water pressure mechanism?

It seems kind of safe to use the same water as used in the reactor for the control rod drive as if a small amount leaks past the piston seal which I believe does happen it simply adds to the reactor core water.Can there realistically be a case where the reactor vessel pressure exceeds the rod driving piston pressure and the control rod drive mechanism cannot supply enough force to move the piston and the rod upwards ? or what happens if a seal or multiple seals get damage as to they don't seal the pressure anymore between the core and outside what would happen? would the rod fall at its outermost/lowermost position or would it be held in place by its latching mechanism? I assume water could leak from the core into the rod drive mechanism or a control valve would block such water movement?
thanks.

 

[rpp]

A lot of questions...

I can't answer the question about the reed switches. This isn't my area of expertise. I do know that it can be common for blades to move two notches at a time, or to not catch and "settle" back to their original movement.

You are correct that there are only discrete axial "notches" where the control blade can be located. There are 24 notches and they are located in 6 inch increments. For some reason, the notches are labeled as even numbers, so the positions go from 0, 2, 4,.. 48.

The BWR control blade movements are completely different than the PWR designs. The PWR designs do not have notches, but they may have discrete axial positions.

A six inch control rod movement can add/remove a lot of reactivity, so you typically do not move control rods at full power. They decrease the power in the core, then move the rods, in order to minimize the local power change around the blade tip. To improve on this, newer reactor designs (like the ABWR and ESBWR) use fine motion control rods that do not have the "ratcheting" blade movement.

Yes, the control blade housing tubes are welded to the bottom of the reactor. This can be seen in Figure 2-11 where it shows the bottom of the reactor vessel.

The piping diagram is shown in Figure 2-10. I can't remember the exact sequence, but opening and closing the valves is what causes the blades to move up and down. Everything is run off of the system pressure (except for the scram tanks). It is the opening of valves and "bleeding" of the pressure that causes a pressure differential and makes things move. Opening valves will cause different pressure differentials at different places. You can move the collet spring, and/or move the blades up and down.

If there is not enough pressure in the system to move the blades, there are high pressure nitrogen scram tanks attached to each control rod drive which have enough pressure to drive the blades in. These tanks are not used for normal operations. The tanks are shown in Figure 2-13. The scram tanks are also shown on the left side of Figure 2-10.

I had to take a test on how all of this works, but it was a long time ago :)

 

[Hiddencamper]

I am reading the BWR general description brochure and a few question arose.

Question: I was reading about the control rod insert mechanism and the means by which the operators know how far a rod is inserted, the brochure mentioned the use of reed switches, now I do know what a reed switch/herkon is and how it works, the thing that wasn't explicitly stated but I think I understood it is that I think the reed switch is located on a fixed position along the rod so called "index" tube and there are magnets located on the index tube with identical spacing and as the tube moves up or down the magnets connect/disconnect the reed switch contacts and so a computer or maybe in older designs an analog/mechanical device can count the times a magnet has passed and so the operator can know the exact position of the rod? Is this about correct?

There is a PIP (Position Indication Probe) that is inside the center of the CRD mechanical drive unit. It sits still. Inside the PIP are reed switches. These are little low voltage switches that, when a magnetic field is applied, they close and bring in a signal. There is a reed switch for every numerical position (00 through 48), along with additional reed switches at Full In (same as 00), Full In 2 (3 inches further than full in, only achieved during a scram with pressure on the under piston), Full Out, and Full Out Overtravel (6 inches more withdrawn than maximum, its a sign that the rod decoupled from the drive and you are pulling the drive out with no rod attached to it, and the rod may be lost somewhere in the core). Around the moving portion of the control rod shaft is a permanent magnet. As the shaft goes in or out of the core, the magnet moves and picks up different reed switches to bring position indication in by opening and closing those switch contacts. The odd reed switches are all tied together. Since the rods cannot "settle" at an odd number, the only real information the odd switches give you is to let you know the rod is in motion, or failing to settle, or drifting (if an odd reed switch engages while the rod has no active drive signal, it means the rod is moving due to a non-operator demand, such as a scram or a equipment failure, and brings in an alarm for the operator to take action or scram the reactor). These reed switches feed back to a multiplexer cabinet, which feeds to the main control room Rod Position Indication System (RPIS). The RPIS decodes the multiplexed signal, and gives you a reed switch matrix display, which feeds into the other rod position display systems (full core display, 4 rod display, plant process computer OD-7 function, rod worth minimizer, rod sequence control system, etc).

Question: From what I understood each control rod index tube has a given set of points at which a horizontal electrically activated gear can latch the tube in place so that it doesn't fall back so does that mean that each control rod can only be stopped and held at specific heights which are predetermined by the design of the rod pushing tube? I assume other reactors like the PWR or RBMK and other can stop the control rods at any specific moment as their holding mechanisms differ?

The latch is NOT electronic. The latches are called "Collet Fingers". The collet fingers are engaged/closed by spring force, and will engage through springs to hold the rod in place. In order to insert a rod, the drive system simply has to push the rod in past the next set of collet fingers, apply a settle signal to relieve hydraulic pressure, and gravity drops the rod down onto the fingers where they latch in place. To withdraw a rod, the rod drive system actually applies a short insert signal to raise the rod off of the fingers, then the withdraw signal, which applies pressure to the top of the drive piston, also applies pressure which holds the collet fingers OUT (disengaged), and allows the rod to drive out. Once the rod reaches the target position, the settle function engages which removes pressure from the drive unit, allows the collet finger springs to push the fingers in place, and the rod settles onto the fingers.

If the collet fingers fail, the rod will keep slowly dropping out of the core, and the operator will use the emergency insert push button to drive the rod to the full in position. Emergency insert bypasses all logic (not counting rod sequence controller under the Low Power Set Point), the rod drive timers, and the settle function, and forces the rod to drive in. This isn't good on the rod's piston seals because there is no settle function, and pressure is relieved through the seals over time, however it allows the operator to drive the rod in. If the rod does not settle after the first emergency insert signal is applied and the collet fingers don't engage, then the operator drives it back to full in and field operators will hydraulically disarm the rod which holds it in place.

The rods can only settle in 6" increments. The positions are 00 through 48. Each number is 3". So 00 is full in. 01 is 3 inches out. 02 is 6 inches out. The collet fingers only allow the rod to latch on the even positions, so if you hit withdraw once, the rod moves from 00 to 02.

Question: Now as I understand the BWR control rod housing tubes are essentially open to the reactor core and it's water if the rod and pushtube itself would be taken out correct? So are these housing tubes fixed into place by welding them into the lower reactor vessel head?
It was also kind of interesting to read that the piston that pushed the index tube and control rod up or down and its seal is essentially the only thing separating the reactor core water pressure and the water pressure applied to the piston from below by the control rod water pressure mechanism?

The drive utilizes guide tubes for the drive shaft to move the rod with. During outages when we need to change out the drive units, we fully withdraw the control rod. The bottom of the control rod itself is a conical shape that is designed to create a nearly perfect sealing surface over the bottom penetration, which prevents reactor water from leaking out. So the bottom of the control rod creates a seal. You typically have to remove the fuel from that fuel cell first withdraw the rod to the full out position, disengage the drive unit, and withdraw the drive to the full out overtravel position. Then, underneath the vessel on the service platform, you have a jacking tool that engages the bottom of the drive unit while your service men unbolt the flange and all slowly lower the drive unit out the bottom. There is a little water drippage, but it's a nearly perfect seal.

The seals for the rod drive pistons aren't really an issue. Normally you apply a 40-50 gallon per minute flow across all CRDs in the core. This purges the seals and keeps them flushed, cleaned, and cooled. The seals are made of graphitar, and can degrade if they are too hot for too long. During a scram, reactor water can blow by the seal and out the scram discharge lines, and this is understood. This is why there is a dedicated scram discharge volume. The scram discharge volume receives the hydraulic water from the scram, and becomes part of the reactor coolant pressure boundary when a scram signal is initiated. When a scram occurs, the scram discharge volume vents and drains automatically isolate under spring pressure to the closed position because you are now part of the pressure boundary.

Question: It seems kind of safe to use the same water as used in the reactor for the control rod drive as if a small amount leaks past the piston seal which I believe does happen it simply adds to the reactor core water.

The purge flow does provide cooling, but you also can supply water to the vessel. Under normal operation you supply 40-50 gallons per minute. When decay heat is low enough or when you are not steaming, this is all the makeup supply water you need. You use the reactor water cleanup system to let down any excess water to keep level steady. During a scram, the rod drive hydraulic charging header opens up directly to the reactor vessel and your CRD pumps go up to runout on their pump curve. For my BWR this supplies about 200 gpm to the reactor vessel via seal injection to the bottom head. Immediately after a scram, this allows RCIC to makeup sufficient water inventory with a minute or two (normally RCIC needs 15 minutes of decay time before it can makeup for boiling losses). After a few hours, this 200 gpm is more than enough water to keep the core submerged if pressure is held steady. So this is a useful means of adding extra water. For the isolation condenser plants, this may be their only high pressure makeup source. This also exceeds the maximum leakage expected from the reactor recirculation pump seals for most failure conditions. The only downside to this, is that the cold water can cause stratification in the bottom head of the reactor. This isn't an issue directly for reactor safety, but it can cause you to exceed cooldown limits on the bottom head if you have no forced flow (at least 1 recirculation pump running at any operating speed/flow, or the reactor water cleanup system running with increased suction from the bottom head drain). Operator training used to have us reset the scram signal as soon as possible to prevent bottom head cold water stratification, but we've seen it's really not an issue in most cases and the extra water can be useful. Taking another scram if the transient isn't over is not acceptable in the eyes of the industry or regulator. Now we just deal with the stratification and raise water level up high enough to establish liquid natural circulation (in contrast to boiling natural circulation).

Question: Can there realistically be a case where the reactor vessel pressure exceeds the rod driving piston pressure and the control rod drive mechanism cannot supply enough force to move the piston and the rod upwards ? or what happens if a seal or multiple seals get damage as to they don't seal the pressure anymore between the core and outside what would happen? would the rod fall at its outermost/lowermost position or would it be held in place by its latching mechanism? I assume water could leak from the core into the rod drive mechanism or a control valve would block such water movement?

No there isn't. The reactor pressure isn't an issue because during a scram, the scram exhaust valve opens up to the scram discharge volume which is at 0 PSIG. The insert side of the drive piston has either CRD accumulator pressure on it (1000ish PSIG for older plants, 1700 PSIG for newer plants), rod drive pump pressure (up to 2000 PSIG), or there is a ball check valve that supplies the reactor's own pressurized water as the driving force (1000 PSIG ish). The over piston side gets exhausted to the discharge volume, so you essentially create at least a 1000 PSIG differential pressure across the piston and the rod drives in. Seal leakage isn't sufficient to overcome the exhaust flow to the discharge volume. The only issue with damaged or degraded seals, is they raise the temperature of the top piston area, so that when the water exhausts to the scram discharge volume it can flash to steam, causing back pressure and reduced scram times. It does not cause a loss of scram capability on its own though, it just penalizes your insertion speed, which may require declaring the rod SLOW. The rod would not fall out of the core, because the spring loaded collet fingers will engage any time the rod falls down to one of its notches on the drive shaft.

 

[Hiddencamper]

A lot of questions...

A six inch control rod movement can add/remove a lot of reactivity, so you typically do not move control rods at full power. They decrease the power in the core, then move the rods, in order to minimize the local power change around the blade tip. To improve on this, newer reactor designs (like the ABWR and ESBWR) use fine motion control rods that do not have the "ratcheting" blade movement.

You can and do move rods at power, but you need to do a lot of analysis using the core monitoring system. You have 4 primary thermal limits you monitor, local critical power ratio, linear heat generation rate, average planar power, and PCIOMR (Preconditioning). Typically you'll be able to pull your power rods out to position 8 to 12 to maintain full power operation. Sometimes you have to take a small down power to provide additional margin.

The normal mode of operation is to set your rod line, then raise core flow to reach target power and use flow to maintain. But as fuel depletes, you have to withdraw rods, and there's typically a number of notches you can withdraw without requiring a down power. I move rods at power all the time because my plant has a very limited maximum rod line and limited core flow capability. So we have to use rods to control power more than most BWRs out there.

 

[girts]

Thanks for such in depth replies, much appreciated.

So if I am reading correctly what you said then not only does the water leak past the rod drive piston seals but it is intended to do so in order to keep them in a healthy condition and as another way by which to supply water to the reactor , although a rather small amount as compared to the other means still significant as you said enough to keep the decay heat in check.
By natural circulation you meant the heat given off by the fuel after scram is not enough to cause boiling under the right circumstances but enough to cause water heating and such heated water travels upwards like in a house boiler and so as it is over the core boundary it can then fall back and cool off via the sides of the reactor between the so called "core shroud" and reactor vessel outer wall where the inlet water pipes are located and the jet pumps through which the inlet pipes enter?
I understand this would be highly unlikely but can't there be a case where if the valves or other parts brake in the case of scram reactor water can leak out through the control rod pipes right past the seals if you say the seals are made such that they permit water flow past them? I assume if the pressure from below the CRD mechanism is lost there would be some pressure coming from the high stack of water which is in the vessel and since the control rods are at the bottom all that pressure concentrates there?

Also as you said when the men go to service the CRD mechanism below the reactor vessel say during a refueling, how can one remove a rod with it's piston if the reactor core is under water for biological shielding and heat removal purposes during refueling, wouldn't the water start flowing out through the rod guide tube if the rod and its piston would be removed or say taken out for most of its length, I assume it depends on where the seals are located but I'm not sure about that one, from the drawings in the brochure it seems the piston with seals is located at the down most position of the rod?
I had a bit of hard time imagining how the collet fingers work in terms of applying opposite pressure to them but I assume that each rod is held in place by more tan just one such finger as they are multiple for each rod spaced equal distance apart?

Also I understand from what you said that in the case of scram condition when the rods are fully inserted and position is read 00 a valve opens so that no excess pressure is accumulated in the reactor vessel due to decay heat slowly heating up the water and rising pressure? Because if the turbine steam outlets are shut off then another bypass must be opened correct?

Lastly speaking about your answer to @rpp answer about rod movement, I read in the "general description" manual that after years of operation about 1980 they made up a new way to use the BWR core called the "CCC" in order to achieve better fuel burnup and lower operator need to adjust control rods for power during operation or fuel lifecycle, now I want to make sure I got it correctly, so the CCC (control cell management) is basically a way to reorganize the BWR core so that after one cycle of the fuel they relocate the most burned assemblies around the control rod blade positions forming arrays, so the highest reactivity assemblies (less burnt fuel) is further away from control blades and lowest reactivity assemblies (most burnt fuel) around the control blades, I understand that by this they achieve the moment where changing the control blade insertion height they get a more smoothed out and less steep response in the core neutron flux and also since the control blades affect the neutron background more to the closest assemblies to the blades the generated heat difference is less if the assemblies closest to the blades are with less reactivity?
and besides this they achieve higher and more even burnup in the fresh assemblies further away from the blades?

is this strategy still used in BWR's and their successors like ABWR etc?

PS. from your last reply about the necessity to move the rods a lot etc I assume you are operating one of the older BWR plants with a lower output capacity?
is you facility using the 7x7 fuel assembly dimensions which were then changed to 8x8 referring to rod count on each side of the bundle? Atleast that's what I read int he manual.

thanks

 

[Hiddencamper]

Question: So if I am reading correctly what you said then not only does the water leak past the rod drive piston seals but it is intended to do so in order to keep them in a healthy condition and as another way by which to supply water to the reactor , although a rather small amount as compared to the other means still significant as you said enough to keep the decay heat in check.

This is correct. After a few hours the rod drive pumps can supply enough water to manage decay heat losses. For plants with isolation condensers, the control rod drive pumps are the ONLY injection after a main steam isolation until you cool the reactor down (typically below 400 PSIG).

Question: By natural circulation you meant the heat given off by the fuel after scram is not enough to cause boiling under the right circumstances but enough to cause water heating and such heated water travels upwards like in a house boiler and so as it is over the core boundary it can then fall back and cool off via the sides of the reactor between the so called "core shroud" and reactor vessel outer wall where the inlet water pipes are located and the jet pumps through which the inlet pipes enter?

BWR natural circulation is a bit different than a PWR. In a PWR, natural circulation is required to move heat from the core to the steam generators, so that it can be removed from the reactor coolant system. In a BWR, natural circulation is about minimizing stratification and not impacting your cooldown limits. When you have no forced flow, you have a small amount of natural circulation due to the boiling effect. Decay heat WILL boil water, and create sufficient flow to keep the core safe. But it doesn't create enough flow to mix the bottom head of the reactor, and with < 100 degF water being injected there you will have bottom head stratification. You can minimize this stratification by raising reactor water level to the steam dryer (typically 2-3 feet above normal operating level), and whenever the core is subcooled this is a requirement to make sure your reactor coolant temperature measurements are valid.

The concern is if you have bottom head stratification, you could violate the 100degF/hr cooldown limit. Or even worse, you could try to start up a reactor recirculation pump and thermal shock the reactor when that cold water now rapidly moves to a 400+ degF section of the vessel. Tech specs, procedures, and thermal shock interlocks should prevent this, but it's still not a good thing.

The area between the core shroud and the outer vessel wall is called the "Downcomer". The downcomer is where condensation from the steam separators/dryers drains to, where the reactor recirculation pumps and jet pumps take suction from, and is where feedwater is injected to.Question: I understand this would be highly unlikely but can't there be a case where if the valves or other parts brake in the case of scram reactor water can leak out through the control rod pipes right past the seals if you say the seals are made such that they permit water flow past them? I assume if the pressure from below the CRD mechanism is lost there would be some pressure coming from the high stack of water which is in the vessel and since the control rods are at the bottom all that pressure concentrates there?

The piping for the CRD system is ASME nuclear boiler classified piping. I believe it's class 1 because it is part of the Reactor Coolant Pressure Boundary (RCPB). If you had a failure of one of those pipes, it would potentially be a small break LOCA and would require a unit shutdown and cooldown. It's unlikely though as it is ASME code

Question: Also as you said when the men go to service the CRD mechanism below the reactor vessel say during a refueling, how can one remove a rod with it's piston if the reactor core is under water for biological shielding and heat removal purposes during refueling, wouldn't the water start flowing out through the rod guide tube if the rod and its piston would be removed or say taken out for most of its length, I assume it depends on where the seals are located but I'm not sure about that one, from the drawings in the brochure it seems the piston with seals is located at the down most position of the rod?

The bottom of the control rod has a cone shaped velocity limiter. This cone shape also acts as a sealing surface. When we go to remove a drive mech unit from the bottom of the reactor, we first fully withdraw the control rod and let is sit over the bottom head penetration. The cone shaped velocity limiter creates a seal which prevents water from leaking out of the bottom of the reactor. When we remove the drive units, we unbolt the flanges and pull them out completely, creating a hole to the reactor, but the control rod bottom seals it.Question: I had a bit of hard time imagining how the collet fingers work in terms of applying opposite pressure to them but I assume that each rod is held in place by more tan just one such finger as they are multiple for each rod spaced equal distance apart?

Take a look at this link: http://patentimages.storage.googleapis.com/pages/US5446774-1.png
29a is the collet finger and 31 is the collet finger springs. The springs push the fingers in place to latch the rod. If you look at the notches (55) each notch in the side is the even number positions (00, 02, 04 ... 48). When you insert a rod, the notches are tapered to allow the collet fingers to be pushed out of the way for an insert signal. When you go to withdraw a rod, the system applies a momentary insert signal to push the fingers out of the way, then pressure on the withdraw header holds the fingers in place. 81 is the withdraw header supply line. If you follow that down, you'll see 69, which is the withdraw pressure to the collet fingers. Follow that up, and you see a small piston below the collet fingers which pushes them up into little slots after they are retracted. Those slots hold the fingers in place so you can continue to withdraw the rod. When the withdraw line is closed, the fingers push back into place and the rod notch settles on the fingers.

Question: Also I understand from what you said that in the case of scram condition when the rods are fully inserted and position is read 00 a valve opens so that no excess pressure is accumulated in the reactor vessel due to decay heat slowly heating up the water and rising pressure? Because if the turbine steam outlets are shut off then another bypass must be opened correct?

This has nothing to do with steam dumping. If you look at that link again, 80 is the insert line, and 81 is the withdraw line. The insert line has two parallel sources of hydraulic fluid, the rod drive system and the control rod accumulator. When a scram occurs the scram inlet valve opens allowing the accumulator to inject to the under piston region and force the rod in.

On the withdraw line you have the normal withdraw valves, then you have a scram outlet valve which opens up. The scram outlet valve directs the over-piston water to the scram discharge volume, which is empty and at atmospheric pressure when the scram occurs. This has nothing to do with decay heat or vessel pressure, it's completely independent from it.

If your turbine steam outlets shut off, there are steam dumps (called turbine bypass valves), which will auto open to control pressure on the main steam header. If the main steamline isolation valves shut, your relief valves will automatically or manually be operated, and operators can also place other systems in service to help control pressure (isolation condenser, RCIC, HPCI, reactor water cleanup, etc).

Question: Lastly speaking about your answer to @rpp answer about rod movement, I read in the "general description" manual that after years of operation about 1980 they made up a new way to use the BWR core called the "CCC" in order to achieve better fuel burnup and lower operator need to adjust control rods for power during operation or fuel lifecycle, now I want to make sure I got it correctly, so the CCC (control cell management) is basically a way to reorganize the BWR core so that after one cycle of the fuel they relocate the most burned assemblies around the control rod blade positions forming arrays, so the highest reactivity assemblies (less burnt fuel) is further away from control blades and lowest reactivity assemblies (most burnt fuel) around the control blades, I understand that by this they achieve the moment where changing the control blade insertion height they get a more smoothed out and less steep response in the core neutron flux and also since the control blades affect the neutron background more to the closest assemblies to the blades the generated heat difference is less if the assemblies closest to the blades are with less reactivity?
and besides this they achieve higher and more even burnup in the fresh assemblies further away from the blades?

is this strategy still used in BWR's and their successors like ABWR etc?

Some background here. The control rods are broken into 4 groups, "A1", "A2", "B1", and "B2". These are the four rod sequences that you can use for operation. In the original core designs, you would start up on one group, and do sequence exchanges every couple of months to equalize fuel burnup. Sequence exchanges take a long time to perform, and require heavily reduced power levels. It was not very efficient, and was a pain on the operators and for capacity factor.

The control cell core is a different fuel loading strategy where you utilize only 1 or 2 rod sequences. The CCC was an improvement. Today we use the "Improved Low Leakage Control Cell Core" design or something similar. We put the oldest fuel on the outer periphery, and it acts like a neutron reflector to the next innermost ring, boosting it's reactivity output. We typically only use 1 rod sequence (my plant is almost always on A2), and we just swap which rods we use in that sequence. You have different "rings" in the core based on fuel loading, and the control rods which are used are based on controlling the reactor's power radially.

We use control rods in the core to help control both reactor power, as well as flux shape. Shallow rods (typically position 32 to 48, that are barely inserted) only impact the axial flux shape of the core. Deep rods (position 16 to 00) affect radial power and full core power. In the middle it depends on the spot in core life and which ring the rods are in. We use rods in combination with burnable poisons in the fuel to shape the flux profile and optimize burnup. Typically we do not use shallow rods anymore. At most, we use them during initial startup/heatup until the fuel is conditioned and xenon is built in sufficiently to allow us to switch to only deep rods (this is why BWRs require a downpower after a day or two of operating, to swap to the normal full power rod pattern).

I do not know what ABWR uses, or the other plants with fine motion control rod drives. I suspect it is similar.

Question: PS. from your last reply about the necessity to move the rods a lot etc I assume you are operating one of the older BWR plants with a lower output capacity?
is you facility using the 7x7 fuel assembly dimensions which were then changed to 8x8 referring to rod count on each side of the bundle? Atleast that's what I read int he manual.

Actually I operate a BWR/6, one with a high power output and a very high power density. We are using GNF-2 fuel, which is 10x10 with 92 fuel rods and 2 water rods.