Thorium Accelerator Driven Nuclear Power - Why not ?

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  • #26
Morbius
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Ok. Overstated a bit. A meltdown is possible in a Candu if the core loses light water coolant as well (ie. if rods are removed from the heavy water and not placed in light water). Although a Candu is designed so that this will not happen, in a catastrophic situation it conceivably could happen. The Candu has an emergency gravity system that will ensure the core is immersed in light water and this will ensure that fission ends without loss of cooling. Since Candus are built adjacent to some body of water, there is no lack of available coolant.

Since in a light water reactor you cannot immediately shut down the reactor by immersing the core in light water (it already is), you need to have some means of absorbing neutrons while in the water (such as control rods) or immerse it in some other coolant that is not a good moderator. If your control rods aren't working you have a big problem. In the Candu you just pour water on it.
Andrew,

Afraid not. Just having the core immersed is not sufficient. You need a method
of getting the heat out.

Actually, the CANDU is NOT safer than a US reactor; if fact a CANDU does NOT
meet the U.S. Nuclear Regulatory Commission standards to be licensed in the USA.

The CANDU NEEDS the extra shutdown mechanism of a D2O dump.

In a US reactor, the light water coolant temperature coefficient is negative. That is, if
the reactor coolant gets hot - it is a negative reactivity insertion.

A CANDU is precisely the opposite - if you lose coolant flow to the CANDU, and the
light water coolant temperature increases, and thus the light water gets less dense;
that is a POSITIVE reactivity insertion in a CANDU because all the moderation that
is needed is provided by the D2O.

U.S. reactors are "under-moderated"; which provides an inherent safety mechanism.
CANDUs are "over-moderated" - thus the D2O dump.

This is one of the reasons why the CANDU fails to meet U.S. safety requirements.

I wouldn't worry too much about losing control rods; they drop by gravity. All that's
required is to cut current to the electromagnets that hold them up and they drop.

Dr. Gregory Greenman
Physicist
 
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  • #27
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To clarify Dr. Greenman's statement regarding controlling the reactor with coolant temperature:

The LWRs I worked on generate heat that is transfered via a heat exchanger to a secondary boiler system. As the coolant passes through the heat exchanger, it cools an amount directly proportional to the energy output.

In a steady state (constant load) system, the reactor is generating the same amount of heat that is taken out of the steam generator in the form of steam.

If the power demand increases, more steam (more heat) is taken out of the steam generator, causing the coolant will give up more of it's heat energy to the heat exchanger to compensate.

The colder coolant is more dense, so when it returns to the reactor, the moderator/coolant slows (or thermalizes) more neutrons because it is more dense, causing an increase in fission rate, which causes an increase in heat production, which is then fed to the heat exchanger.

Eventually the reactor will be producing more heat than is being required to compensate for the steam being removed. Because of that excess heat, the coolant will not be giving up all the heat energy in the steam generator, leaving the coolant warmer (less dense). The moderator/coolant thermalizes fewer neutrons, causing the fission rate to slow, reducing heat output.

After several progressively smaller cycles, the system will settle at the new, higher power output with the coolant giving up more heat than it was at the lower power level. The mean temperature of the coolant, however, will have remained virtually unchanged steady-state to steady-state.

I don't know if that helped, now that I read it, but there you are.
 
  • #28
Astronuc
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In a US reactor, the light water coolant temperature coefficient is negative. That is, if the reactor coolant gets hot - it is a negative reactivity insertion.
I've actually seen some core designs with positive MTC, which depends in enrichment, burnable absorber distribution and cycle length.

I wouldn't worry too much about losing control rods; they drop by gravity. All that's required is to cut current to the electromagnets that hold them up and they drop.
This is the case for PWRs, but BWR control blades must be hydraulically inserted from the bottom.

Dewey2k said:
The LWRs I worked on generate heat that is transfered via a heat exchanger to a secondary boiler system.
That would be a PWR. BWRs generate steam in the core, which passes through a dryer above core before being sent through the main steam line directly to the high pressure turbine. That can be a major disadvantage when there is one or more fuel failures and the Xe and Kr fission gases are transported to the turbine. Another disadvantage is that N-16 can be transported from the core to the turbine if a plant uses hydrogen water chemistry (HWC) in order to protect the stainless steel from IGSCC/IASCC. N-16 has an energetic gamma.

If the power demand increases, more steam (more heat) is taken out of the steam generator, causing the coolant will give up more of it's heat energy to the heat exchanger to compensate.

The colder coolant is more dense, so when it returns to the reactor, the moderator/coolant slows (or thermalizes) more neutrons because it is more dense, causing an increase in fission rate, which causes an increase in heat production, which is then fed to the heat exchanger.
Actually, Morbius has been discussing decay heat removal which is after reactor shutdown.

During operation, the primary coolant system temperature does not vary all that much. Normal core inlet temperature in a PWR is ~290°C and the exit temperature is around 325-330°C. The reactor coolant pumps (RCPs) for coolant through the core. A significant change in load would like cause insertion of the control rods, but I'd have to confirm that. Some plants do have grey rods for axial power distribution control or load follow, however in the US load following is not a practice, although some plants have done it.
 
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  • #29
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That would be a PWR. BWRs generate steam in the core, which passes through a dryer above core before being sent through the main steam line directly to the high pressure turbine. That can be a major disadvantage when there is one or more fuel failures and the Xe and Kr fission gases are transported to the turbine. Another disadvantage is that N-16 can be transported from the core to the turbine if a plant uses hydrogen water chemistry (HWC) in order to protect the stainless steel from IGSCC/IASCC. N-16 has an energetic gamma.
I thought Nickel (from the crud) was the main gamma problem with HWC. Where is the Nitrogen from, a fission product?
 
  • #30
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Actually, Morbius has been discussing decay heat removal which is after reactor shutdown.

During operation, the primary coolant system temperature does not vary all that much. Normal core inlet temperature in a PWR is ~290°C and the exit temperature is around 325-330°C. The reactor coolant pumps (RCPs) for coolant through the core. A significant change in load would like cause insertion of the control rods, but I'd have to confirm that. Some plants do have grey rods for axial power distribution control or load follow, however in the US load following is not a practice, although some plants have done it.
The reactor plants I worked on were on an aircraft carrier, which regularly received power transients as high as 80% reactor power (self sustaining power was approximately 20% in that mode of operation for RCPs and other equipment). Extreme transients could drop average coolant temperature by 2C or 3C initially before Reactor Power stabilized.

Average coolant temperature was adjusted by rod height only to meet minimum requirements for steam plant operation, temperature compensation for Xe-135 and other fission product neutron absorbers, or for operational readiness.

Unfortunately (or fortunately), my ship was relatively new so we didn't really have a huge issue with decay heat. Even without RCPs we could establish enough flow to keep the reactor cooled by thermal flow if necessary.

As an aside, the last time I touched a control rod shim switch was in 1995, so I'm a little rusty.
 
  • #31
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I thought Nickel (from the crud) was the main gamma problem with HWC. Where is the Nitrogen from, a fission product?
This is taken out of context, but here you go:

http://www.ans.org/pubs/journals/nt/va-147-2-269-283"

As a water-cooled nuclear system with a direct thermal cycle, the supercritical-water-cooled reactor (SCWR) shares with the boiling water reactor (BWR) the issue of coolant activation and transport of the coolant activation products to the turbine and balance of plant (BOP). Consistent with the BWR experience, the dominant nuclide contributing to the SCWR coolant radioactivity at full power is N-16, which is produced by an (n,p) reaction on O-16.
A neutron is absorbed by an O-16 atom, but it kicks out a proton becoming N-16.
 
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  • #32
Astronuc
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I thought Nickel (from the crud) was the main gamma problem with HWC. Where is the Nitrogen from, a fission product?
Nickel in crud is an issue when it gets into the condensate in a BWR or in reactor cavity in LWRs during refueling. LWRs have filter/demins to remove crud as much as possible, but then the demins get hot.

N-16 comes from the n,p reaction of O-16 and in a reducing environment volatile nitrogen compounds form (IIRC, amines), which can be carried to the turbine in the steam.
 
  • #33
Astronuc
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The reactor plants I worked on were on an aircraft carrier, which regularly received power transients as high as 80% reactor power (self sustaining power was approximately 20% in that mode of operation for RCPs and other equipment). Extreme transients could drop average coolant temperature by 2C or 3C initially before Reactor Power stabilized.
Well yes, naval reactors are quite different animals than commercial power reactors. Enrichments are higher and that certainly leads to differences in reactivity management/control.
 
  • #34
Morbius
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After several progressively smaller cycles, the system will settle at the new, higher power output with the coolant giving up more heat than it was at the lower power level. The mean temperature of the coolant, however, will have remained virtually unchanged steady-state to steady-state.
Dewey,

What you are describing is the ability of a Pressurized Water Reator or PWR [ which I guess
is probably what you were working with] to "load follow".

In a PWR, the turbine throttle valve is opened/closed to match turbine power to the plant's
electrical load. The reactor power will adjust automatically via the coolant/moderator
temperature coefficient - to match reactor power to that needed to drive the turbine at
the desired throttle setting.

Dr. Gregory Greenman
Physicist
 
  • #35
Morbius
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This is the case for PWRs, but BWR control blades must be hydraulically inserted from the bottom.
Astronuc,

Yes - in the case of a BWR; there is a hydraulic cylinder with a piston connected to the
control rod drive shaft. The bottom side of the piston is connected to the inside of the
pressure vessel so that the bottom side of the piston is at reactor vessel pressure.

The volume of the cylinder above the piston can be vented to ambient pressure.

If a fast shutdown or SCRAM is desired; a valve is opened and the pressure above the
piston drops to essentially normal atmospheric pressure. It is the pressure differential
between the reactor pressure and the ambient pressure that drives the control rod up
and into the core.

So as long as the reactor is at pressure; there is drive pressure to force the control rods
into the core.

Dr. Gregory Greenman
Physicist
 
  • #36
Morbius
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The reactor plants I worked on were on an aircraft carrier,
Dewey,

Just to add; all naval reactors are PWRs - even the ones designed by General Electric
[KAPL]. GE's commercial power reactors are BWRs. However, one doesn't want a
"free-surface" i.e. an interface between water and steam; in a reactor that is moving,
where the water and the free-surface can "slosh" around.

Unfortunately (or fortunately), my ship was relatively new so we didn't really have a huge issue with decay heat. Even without RCPs we could establish enough flow to keep the reactor cooled by thermal flow if necessary.
The fission products that give you a shutdown decay heat that is 7% of nominal power
are all short lived fission products. So they equilibrate VERY quickly to an equilibrium
value. If the aircraft carrier reactor has been operating for a few days, or even less;
you essential have reached the equilibrium level of fission products that give the
decay heat that is of concern from a meltdown point of view.

Dr. Gregory Greenman
Physicist
 
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  • #37
Morbius
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Well yes, naval reactors are quite different animals than commercial power reactors. Enrichments are higher and that certainly leads to differences in reactivity management/control.
Astronuc,

Yes - naval reactors use HEU - Highly Enriched Uranium for fuel. In fact, the HEU that
is used in naval reactors has a higher enrichment than "weapons-grade" uranium.

Ther primary reason was so that one could override a "Xenon transient". In a normal
power reactor; if the reactor is shutdown, one has to wait a day or so for the Xenon
to decay in order to restart the reactor. That would be unacceptable as the power plant
of a naval vessel. A warship has to be able to move anytime the captain orders it to
move. They can't sit around and wait for Xenon to decay. So naval reactors are fueled
with HEU. Naval reactors are also not refueled on an annual basis. The current
generation of reactors used in Trident subs and Nimitz-class carriers will go for about
20 years between refuelings. In fact, in the subs; there's no hatch over the reactor to
facilitate refueling. When a Trident is refueled, they cut a hole in the hull; and repair
it after refueling. A Trident will probably be refueled only once in its operating lifetime.

From the late '50s to the '80s; the Ford Nuclear Reactor at the University of Michigan
used naval reactor fuel. If you toured the FNR, they would show you a mock fuel
assembly. That mock assembly had "Property of U.S. Navy" embossed on the side.

There were many other university reactors that used highly enriched uranium as fuel.
There was concern in the late '70s that the use of HEU in university research reactors
was a proliferation risk - someone could hijack HEU meant for a university research
reactor.

Starting in the '80s there was a program run out of Argonne National Lab called RERTR -
Reduced Enrichment Research and Test Reactors. The goal was to redesign the fuel
for research reactors so that they were no more than 20% in enrichment. The FNR at
the University of Michigan was the test-bed for that effort. The FNR operated for many
years on the reduced enrichment, and was shutdown a few years ago.

At present, practically all university research reactors use fuel that is about 20%
enriched. I know of one university reactor, my alma mater's; that still uses HEU.
That's because the core is very small; and one needs the high enrichment to operate.
However, that also means that when it is refueled; the incoming charge of fresh fuel
is less than that needed to make a bomb. So the bad guys can't get their hands on
enough HEU in a single hijack.

If that reactor is redesigned, [ it's in its second incarnation having been designed with
this very compact core in the early '70s ], the redesign would also entail being able
to operate at lower enrichment.

Dr. Gregory Greenman
Physicist
 
  • #38
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All,
Thanks for the feedback. I kind of feel like I trespassed into hallowed ground, at it were. :redface: :biggrin:
 

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