Thorium Accelerator Driven Nuclear Power - Why not ?

[Dewey2k]

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.

 

[Astronuc]

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.

 

[Astronuc]

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.

 

[Morbius]

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

 

[Morbius]

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

 

[Morbius]

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

 

[Morbius]

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

 

[Dewey2k]

All,
Thanks for the feedback. I kind of feel like I trespassed into hallowed ground, at it were.