India Nuclear News and Discussion 4 July 2011

[A_Gupta]

This is what someone on LinkedIn writes:

The main engineering challenge with FBRs is the liquid metal coolant. There is no other industrial application for liquid metal. So, all the major components of this type of reactor are custom-built and very expensive. Most FBRs use liquid sodium, which has intrinsic safety issues. Sodium burns in contact with air and explodes in contact with water. If the liquid metal coolant solidifies for any reason, the reactor can never be repaired. Corrosion is an issue in Lead Bismuth-cooled reactors.

[sanjaykumar]

Interesting. Why Na+ specifically? Lithium cooling?

[Amber G.]

My Take - A Few Point:

While the PFBR’s headlines are about “500 MWe” and “fuel loading,” the deeper significance is that it’s a large-scale physics experiment in fast neutron chain reactions, liquid-metal thermodynamics, and radiation-material interactions — areas where theory and experiment meet under extreme conditions.

It’s one of the few projects in the world where basic nuclear and condensed-matter physics are being tested at power-plant scale.

I will say ( and nuclear physicist will agree) India’s PFBR do offer several deep physical insights and “new learning” opportunities in basic nuclear and thermal physics, not just applied engineering.


It is IMO one of the most significant milestones in India’s nuclear program in decades, and the above story hides some truly deep engineering and scientific achievement beneath its bureaucratic language.

Note that AFAIK only Russia currently operates the large commercial fast breeder, the (BN-800 at Beloyarsk (800 MWe)). Other countries — the U.S., France (Superphénix), Japan (Monju) — all tried and eventually shut down their fast breeders,

That’s why India’s PFBR achieving full fuel loading is globally rare!, It is:

Technically: One of the world’s most advanced nuclear projects — mastering fast-neutron and sodium-coolant technology.

Strategically: The gateway to India’s thorium-based future and nuclear self-sufficiency.

Scientifically: A culmination of decades of indigenous R&D in materials, reactor physics, and sodium engineering.

Symbolically: Proof that India has persisted where many advanced nations gave up.

FWIW - My take - continue reading if interested in scientific details:
More of this cont.. in next post(s)

[Amber G.]

... India’s PFBR do offer several deep physical insights and “new learning” opportunities in basic nuclear and thermal physics, not just applied engineering.

This is dealing with more of Physics - Next post Engineering - Skip if not interested.

Here are the key ones:

- Fast Neutron Physics — Cross-sections and Kinetics

(Most of the world’s operating reactors are thermal reactors, meaning they use moderators to slow neutrons down to ~0.025 eV.
In contrast, fast reactors operate in the MeV range (10⁶ eV).

That changes the basic nuclear physics in important ways:
The neutron transport equation and reactor kinetics equations behave differently.

- - Neutron Economy and Breeding Ratios
A “breeder” reactor relies on balancing fission, capture, and leakage in such a way that: Breeding ratio = new fissile atoms created/ fissile atoms consumed >1

Understanding and measuring this ratio in real time is a highly nontrivial physics .
(PFBR operation contributes fundamental nuclear data to the world’s knowledge base) not just power.

- Thermal Hydraulics of Liquid Metals

(From a heat-transfer physics point of view:

Sodium has very high thermal conductivity, Low viscosity,
No phase change up to ~880 °C.

(So the governing equations of convection and turbulence inside a sodium loop are quite different from water systems.)

many similar things.. just to name a few important ones..
- Materials Physics under Fast Neutron Irradiation
- Neutron Feedback and Nonlinear Dynamics
- Gamma Transport and Radiation Shielding

These are all real, experimentally testable physics processes — not purely engineering.

[srin]

I don't know, the fact that it uses molten sodium makes me shudder. Lots of what ifs come to mind - what if it comes in contact with steam or even air.

I'm sure the reactor would be over engineered with multiple redundancies for just this scenario, but the discomfort doesn't go away

[Amber G.]

..This is dealing with more of Physics - Next post Engineering - Skip if not interested.

Cont ..Engineering challenges ..

** Long post - skip if not interested - comments, observations (from somebody who studied this since 60's), and technical unpacking of the AshokK article **

Yes, FBRs are technically challenging — sodium coolant fires, complex control systems, and high neutron flux make them difficult to operate safely and economically. Historically built but shut down: > 20 research and demo units across the world.

However, their potential to breed more fuel than they consume and to burn long-lived actinides keeps interest alive. Especially in India

So this development is indeed one of the most significant milestones in India’s nuclear program in decades.

Few points:

- Context: India’s “Three-Stage Nuclear Program

The PFBR (Prototype Fast Breeder Reactor) at Kalpakkam is the keystone of India’s long-planned three-stage nuclear fuel cycle conceived by Dr. Homi Bhabha in the 1950s:


Stage - I→ PHWRs) → Natural uranium→Electricity + plutonium (in spent fuel) → Pu-239
Stage-II→FBRs →Plutonium–uranium mixed oxide (MOX) →More Pu + electricity→Breed more fissile fuel from U-238
Stage- III→ Advanced Thorium Reactors →Thorium-U-233 cycle →Sustainable, self-replenishing fuel→Utilise India’s abundant thorium reserves

- PFBR = the bridge between Stage I and Stage III.

So when the PFBR becomes operational, it marks the entry into Stage II, i.e., plutonium recycling and breeding new fuel from U-238.

- Why it’s called “fast breeder”

“Fast” → Unlike conventional reactors, there’s no neutron moderator (like water or graphite). The reactor uses fast neutrons, which allow U-238 to convert to Pu-239 efficiently.

“Breeder” → It produces (breeds) more fissile material than it consumes — turning otherwise non-fissile uranium (U-238) into useful plutonium (Pu-239).

“Prototype” → It’s a full-scale demonstration of the technology before deploying future commercial breeder reactors.

- Why the engineering is so demanding

- Sodium coolant — both brilliant and dangerous

The PFBR uses liquid sodium instead of water as a coolant because water would slow (moderate) neutrons.

Sodium transfers heat extremely well and stays liquid over a wide temperature range — perfect for fast reactors.

But: sodium reacts violently with water and air. Even a pinhole leak could cause fires or explosions. Designing and welding hundreds of Kms of sodium piping without leaks is immensely challenging.


- Fuel handling at high temperatures

The fuel is MOX (mixed oxide) — uranium-plutonium oxide.

→ requiring specialized cladding materials (like D9 austenitic steel) that can tolerate >550°C and high neutron flux.


- Reactor kinetics and safety

The reactor has no moderator and little inherent negative reactivity feedback, so control must be ultra-precise.

The physics is governed by fast neutron spectra — with different delayed neutron fractions and feedbacks — making stability control far more delicate than in water-cooled designs.

- Materials science challenges

Fast neutrons cause intense damage and swelling in steel structures.


- Seismic, thermal and structural design

The massive pool of liquid sodium (~1000 Tons) must be kept pure…

Global context — AFAIK only the second of its kind


- Strategic and scientific implications

India’s known uranium reserves are limited, but it has vast thorium reserves. Breeding plutonium from U-238 and later converting thorium to U-233 is the long-term solution.

If PFBR succeeds, India can close the fuel cycle — reusing spent fuel rather than discarding it as waste.

Breeders “burn” long-lived actinides, reducing high-level waste inventory.

If commercial FBRs follow, India becomes the breeder cycle end-to-end.


(Original criticality target was around 2010–2012, but safety reviews, sodium-system qualification, and fuel-handling issues delayed it.

The fuel-loading step now is the most crucial milestone before “first criticality”
After low-power physics tests, AERB will permit power ascension, likely over months, before grid connection.

Some subtle but interesting (!) points

The PFBR is not under international safeguards, since it uses indigenous plutonium and is part of India’s strategic fuel cycle.

The coolant systems and sodium test loops at IGCAR (Indira Gandhi Centre for Atomic Research) are themselves major R&D achievements.

The plant is designed for 70% load factor, with doubling time of plutonium inventory ~10 years — meaning every 10 years, it breeds enough extra fuel for another similar reactor.



India’s PFBR is:

Technically: One of the world’s most advanced nuclear projects — mastering fast-neutron and sodium-coolant technology.

Strategically: The gateway to India’s thorium-based future and nuclear self-sufficiency.

Scientifically: A culmination of decades of indigenous R&D in materials, reactor physics, and sodium engineering.

.Amber G. - Breeding optimism one neutron at a time.

[Amber G.]

Interesting. Why Na+ specifically? Lithium cooling?

..the fact that it uses molten sodium makes me shudder...

Interesting points but see details in above post.. Sodium has excellent heat transfer properties, a low neutron absorption cross-section, and stays liquid over a wide temperature range — ideal for fast reactors. Lithium’s melting point is too high and it reacts even more violently with air and water..(IOW..Li melts too hot and is far nastier to handle chemically.)
(Lithium (or lithium–lead alloys) has been used in fusion research, not in power-producing fission breeders.. except may be some *old* military type classified ones)

[Tanaji]

Many thanks AmberG for the posts. A couple of questions:

• Web searches say that some FBRs have used lead as a coolant. Lead is used for shielding, so doesn’t it slow or absorb electrons as water does?
• Once sufficient Pu239 is built up, stage 2 is supposed to generate U233 by burning thorium as a blanket. Stage 3 is then supposed to be another breeder that burns U233 and Thorium to create more U233. Since you can burn Th in stage 2 itself, is the advantage of stage 3 that you dont need Pu239 to burn Th?
• Stage 3 sounds magical - burn Th + U233 to get more U233 and then keep refuelling using Th alone. Can you please explain the physics behind this if possible?

[Amber G.]

Many thanks AmberG for the posts. A couple of questions:

• Web searches say that some FBRs have used lead as a coolant. Lead is used for shielding, so doesn’t it slow or absorb electrons as water does?

[ Clarification: A typo - you wrote electron/meant neutron - FBRs deal with neutrons, not electrons.)
Nice Question - (Short answer - for India's perspective it is not used)..
Lead's function as a coolant in a Lead-cooled Fast Reactor (LFR) is based on its specific interactions with neutrons, which are different from its interactions with the gamma rays and other charged particles it shields.

"Shielding” usually means stopping gamma rays or charged particles, but lead’s role in FBRs is different. Liquid lead (or lead–bismuth) coolants have a very low neutron moderation power compared with hydrogen in water, so they preserve the fast neutron spectrum needed for breeding. They’re heavier, less reactive, and much more radiation-resistant than sodium, though harder to pump and manage due to corrosion and high density.

All said - Sodium–potassium (NaK) is by far the most common alternative coolant.
- Lithium (or lithium–lead alloys) has been used in fusion research, not in power-producing fission breeders.
- As said *old* USSR’s Alfa-class submarines (1970s) used liquid-metal cooled reactors with lead–bismuth or lithium–lead mixture..
- Very few (IMHO) use lead. (AFAIK Russia uses/plans to use mixed nitride fuel (U-Pu-N). Key step in Russia’s closed fuel-cycle program... China's CLEAR etc..

All said: Only Russia’s Brest-OD-300 is actually nearing operation as a lead-cooled fast reactor...China and the EU are next in line with prototypes.

India, for now, sticks with sodium FBRs — though its researchers keep an eye on lead-bismuth technology for future hybrid or ADS systems.

So: India’s three-stage nuclear program is tightly focused on sodium-cooled fast breeders, not lead. So PFBR at Kalpakkam uses liquid sodium because:

-India has decades of sodium technology experience (FBTR, IGCAR heritage).
-The PFBR is designed to breed plutonium-239 for the thorium cycle, and the physics/design base assumes sodium’s neutron economy.

(Lead or lead-bismuth systems have been discussed at IGCAR mostly for materials testing and ADS (Accelerator Driven Systems) —
-- BARC-ADS project has mentioned lead-bismuth as a potential coolant, mainly for spallation neutron target systems — again, experimental only.)

[Amber G.]

Once sufficient Pu-239 is built up, stage 2 is supposed to generate U-233 by burning thorium as a blanket. Stage 3 is then supposed to be another breeder that burns U-233 and Thorium to create more U-233. Since you can burn Th in stage 2 itself, is the advantage of stage 3 that you don’t need Pu-239 to burn Th?

Exactly — that’s the main idea.
In Stage 2 (FBR), plutonium (Pu-239 or Pu-241) from spent PHWR fuel is used as the driver fuel. Thorium-232 in the surrounding blanket absorbs fast neutrons and breeds fissile U-233, but the reactor itself still depends on plutonium for fission power.

In Stage 3, the reactor is designed to run directly on U-233 as the main fissile fuel, using thorium as the fertile material. So yes — the key advantage is self-sufficiency: once enough U-233 has been bred, India can theoretically sustain its reactors with thorium alone, without needing more plutonium.

That marks the transition from a Pu-based cycle to a Th-U233 cycle — which is the whole strategic goal, given India’s limited uranium but vast thorium reserves.
-----

Stage 3 sounds magical — burn Th + U233 to get more U233 and then keep refuelling using Th alone. Can you please explain the physics behind this if possible?

It’s not quite perpetual motion — it’s a clever balance of neutron economy.

Here’s the nuclear physics logic:

Fissile reaction:-> U-233 + n → Fission → Energy + 2 to 3 neutrons

Breeding reaction: -> Th-232 + n → Th-233 → Pa-233 → U-233 (after β-decays)

So each fission of U-233 releases a few neutrons...
To sustain breeding, the system must:
- Use 1 neutron to keep the chain reaction going (fission another U-233).
-Use 1 neutron to convert Th-232 → U-233 (to replace what was burned).
-Lose a few neutrons to leakage and non-productive captures — so a high neutron yield and low parasitic losses are crucial.

--- Added later: (Some more physics for experts :
(The U-233 isotope is special because its neutron yield per neutron absorbed (η) is around 2.28 in the thermal spectrum, significantly higher than U235 or Pu 239. This crucial margin above the minimum of 2 is what makes thermal breeding possible

Here U-233 is used both as the fuel that fissions and the product that replaces the consumed fuel would reinforce the closed-loop concept.)
----

In practice, Stage 3 would likely be a Molten Salt Breeder Reactor (MSBR) or an Advanced Heavy Water Reactor (AHWR) using Th-U233 fuel, carefully optimized to achieve near-breeding conditions.

[A_Gupta]

Has a proof of concept Stage 3 reactor been built?

[Amber G.]

Has a proof of concept Stage 3 reactor been built?

As you may know - Short answer: No – there is still no full-scale, operational “Stage-3” Th–U-233 commercial reactor. India has a mature design (the AHWR) and small research demos (KAMINI).. as we have discussed before..

Other countries (notably China) are doing molten-salt/thorium experiments — but a proof-of-concept commercial Stage-3 plant has not been built yet. (see note in the end)

(AHWR is a completed design (≈300 MWe) intended to demonstrate thorium use and many passive-safety features; BARC describes it as the demonstrator for Stage-3

KAMINI at Kalpakkam is a small research reactor that does run on U-233 fuel (a lab-scale proof that U-233 can be used in reactors), but it is tiny (tens of kW thermal) and not a power-plant demonstrator.

Added later: Note: A small, experimental Molten Salt Reactor (MSR) in China MSR-LF1 achieved criticality in late 2023 and has demonstrated the use of thorium salt, marking a significant step toward an MSR-based U233 breeding cycle.