X-posting from India-Russia realtions thread.
Just for reference. I would edit a detailed reply.
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Nuclear Energy in India : Updates
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X-posting from India-Russia realtions thread.
Just for reference. I would edit a detailed reply.
Apologies, seems I had overshot the time limit for editing the post
Now, as the Russian reactors currently in production are mainly lightwater reactors [LWR] of Pressurised Water Reactor [PWR] type, let me compare similar third generation PWR reactors using open source info, as I dont have a ringside insights on them, neither am I a nuclear reactor engineer. So I am open to corrections as always and the below is just my understanding of such a vast and complex field of study.
Caveats apart,
The current Russian reactor model belongs to VVER family (Water-Water energy reactor) that uses light water for both moderation and cooling. The family traces its origins to 1970 when Soviet Union [SU] went about to develop nuclear power industry in large scale for cheap power production. This was one of the three leading designs then in USSR, the most advanced and safest of all three, but the costliest too, within the SU. Starting as a small experimental design of 70 MWe power, there are VVER designs of FIRST, SECOND and THRID gen and current VVER-1200 , VVER-TOI and our own KKNPP models are Gen 3+ at 1000 to 1300 MWe capacity.
Now, the contemporaries of current VVER are offerings from France and US (if those companies are still around after the crisis in them).
The french design is EPR-1600 (Evolutionary Power Reactor formerly European Pressurised Reactor) of 1650 MWe from Areva of yesteryears.
The US design is AP-1000 from Westinghouse.
The Chinese have bought designs from all (US, RUSSIA, FRANCE) and have a series of variants and modifications based on them going, but hasn't reached maturity.
Without going to their developmental history, suffice to say mainstream Power Reactors are truly evolutionary in nature, with each variant building upon its predecessor.
Without much ado, let me address the primary question, that is if the VVER are the best design in industry.
Fundamentally, all designs are choices and trade-offs, so calling one the best in absolute terms is impossible, especially in Power reactors, where the fundamental physics and engineering hasn't exploded, it is just a slow and gradual evolution in this industry.
Having said that the thrust of all the variants of designs has been to anticipate maximum of the potential accidents and malfunctions in design stage itself (called Design Basis Accident [DBA]) and build in mitigations if not prevention of such unfavourable operating conditions. Also the evolutionary improvement in efficiency and better burn-up of fuel is another thrust.
When it comes to DBA mitigation, the emphasis has been on utilising/incorporating in the design maximum number of redundant, diverse acting and preferably passive systems to effect say emergency core cooling in times of Loss Of Coolant Accident [LOCA] and such.
The VVER design in its various variants in prduction now incorporates elements of this. The Gravity Driven Water Pool [GDWP] and passive decay heat removal system of KKNPP are examples. But the EPR and AP1000 designs incorporate these into their designs from initial stages and are better integrated than VVER-1000. This has been the result of them being designed and developed later in time than the russian ones.
The reduction in components in the operation of reactor, especially the coolant loops, and reduction in active components( components that require power to operate ) is better in EPR and APR with 3 and 2 primary cooling circuits versus 4 in VVER. Also the VVER hasn't given as an important emphasis on passive operation as the other two.
In summary, better integration of third gen features give a slight edge to the others than VVER. Do note reduction in power consuming systems and complexities also improve efficiency, although not by a big margin. Let me add, none of this is really a deal-breaker and Nuclear safety systems are over-engineered to the hilt as we all know what happens if anything were to go wrong and there happens to be a weak link in the chain of systems. The above is the purely technical aspect of the question, as I understand it.
But my disagreement with @randomradio is not there. Designing a good reactor is a job half done. Building that design is the other half, where Russians doesnt have an enviable record. Consider the units 1 and 2 of Koodankulam [KKNPP]. The multiple reactor trips in its start-up and frequent outages doesn't inspire confidence. Though unrelated to nuclear island, the brand new Turbo-generator [TG] at unit 1 had to be refurbished within a year or two of commissioning. This paints a bad picture of Russian industry. I believe you have noted the move to replace Russian TG with Jap ones. Also the general decay of Russian industry after the collapse of Soviet Union and subsequent collapse of supplychains, economy etc is not a confidence generating situation. Even during Soviet times, implementation has been their achilees heel. Though not a contributor, there had been KGB reports about construction non-conformities in Chernobyl plant. Add to it the revelation of corruption and mal-practices in suppliers to Atommash. Though I would like to add, officially Russia has clarified none of those suspect components has made its way to India.
Inspite of these, the operational record of KKNPP 1 is not one to go to town about. And worst is the Russian insistence that the frequent trips are due to Indian Operator incompetence, after they being trained in Russia (Novovoronezh is where I remember, but it just is a recollection from my memory, may be wildly off the mark). Now consider the operational record of Kalpakkam reactors. The MAPS 1 unit was commissioned under great stress as Canadians left us in the dark after the '74 tests and its been basically a systematic and scientific jugaad. Still it compares much better than KKNPP.
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Regarding comparison of various reactors, I'm collecting data. Efficiency, Burn-ups etc, is a little difficult to get. Also the wide variety of reactor designs add to this. I shall pen one response when I get enough quantitative data to do an analysis.
I don't get what you mean here. The reactors are operating at well over 95% capacity factor.
And even 100%.
Kudankulam nuclear power plant generates full 2,000 MW capacity for the first time - Firstpost
"Unit 2 of the Kudankulam nuclear plant in Tamil Nadu reached full capacity of 1,000 MW today (Tuesday). For the first time both units 1 and 2 attained full generation capacity and KNPP and became the first nuclear plant in India to generate 2,000 MW of power," a Rosatom release said.
So the starting trouble, whatever it actually was, with KKNPP-1 was fixed a long time ago. And it's difficult to go by media reports about reactors in India is because the govt likes to keep things quiet and there are a lot of vested interests writing whatever they want.
As for AP1000, it is a very safe reactor, but cost overruns have been ridiculously high to the point where they are not even telling the actual cost of the reactor anymore. They are potentially talking about mroe than $8B for each.
As for EPR, the French themselves agree that their reactor is not up to the mark for the global market.
French regulator to EDF: don't assume new reactor model is accident-proof - Reuters
https://www.nirs.org/wp-content/uploads/reactorwatch/newreactors/eprfukushima_summary.pdf
The EPR cannot handle a Fukushima type situation at all. That's why they are going around developing the EPR-2.
So, as of right now, the EPR-1 is not safe, AP1000 is too expensive, whereas the VVER-1000 is safe, cheaper and works.
Due to a combination of generous credit terms and diplomatic compulsion, we may end up building 6 reactors each of the EPR (hopefully EPR-2) and AP1000, but I don't see a significant future beyond that. The current plan is to build 2 of each by 2031. Otoh, the VVER-1200, VVER-TOI and the more advanced SCWR called VVER-1700 could end up defining our future reactor inventory, apart from our own designs of course.
From 2030 onwards, we will need at least 10-15GW every year.
Sir, I have nothing to say, if you account those as just starting trouble/teething trouble.
IIRC the problems were :-
a)Multiple tripping of reactor by core safety systems and emergency protection systems
b)Inability to operate at designated power ( specified during startup regime ) for a specified time, happening multiple times, and reactor again tripping by emergency systems
As I said even MAPS 1 didnot face these much issues during startup. Again, if you believe the regimes of operation during startup is trivial, I am at a loss of words as the various dynamics of reactor comes into play and create a complex operation condition during certain dP/dt regimes and startup is pretty important there. If you agree with me there, then I would like to know if you would have rejected these deficiencies as just teething issues, when you read the same in conjuction with startup of Unit 1 of Leningrad NPP & Ingalina Unit 1 in 1970s. Suffice to say, the last time Russians/Soviets decided to address these shortfalls in design, the world did sit up and took notice.
Again, I urge you to look at the date line of the article you quoted, and it reads IANS Dec 05, 2017 22:38:34 IST . The date of criticality of Unit 2 is 10 July 2016 . There lies my proof. And Im not quoting sme hearsay from Indian media, who know nothing about Nuclear Reactor except that it is a giant kettle and the spelling "RADIOACTIVE". I used to follow the ops of the said reactor almost daily from its grid synchronisation at the Load Despatch data put out in their website.
Had the secondary power supply in case of total power blackout of Fukushima Dai-ichi had been at a higher ground, they too would have rode the waves out. There isn't any certain pre-ordained method though the International community became pretty stringent after March 2011.
I never said VVER isn't cheap. It definitely is.
First of all I cant fathom Indian obsession with LWRs. For a economy with capabilities like India and the Uranium reserves, and its constraints that we face, it was not for nothing, we embarked upon 3 tier programme and PHWR. The PHWR and the requisite industry complexity is manageable within Indian context. Now having mastered that path, we, are as usual going for those shiny new LWRs.. Typically Indian in strategic foresight.
Right now Supercritical reactor with single coolant loop, I guess how effective it is anybody's guess.Radiation shielding etc will be an issue, if we look from DBA perspective etc. But my information is limited, so I have nothing to say really there.
If I recall, the reactor started operations long before it was ready. Probably political pressure was playing its part. Even then I wouldn't go by media reports for anything to do with nuclear. Regardless of whatever plagued the reactor in the beginning, it's now functioning at 95-100% capacity factor, can't expected anything better. What matters now is more important.
The MAPS reactor already had cousins in the form of CANDU and RAPS-2. MAPS is actually a slightly modified clone of RAPS-2. Even then MAPS-1 took 3 years after constuction was done before it could generate full power. And full power meant 170MW out of its estimated 220-235MW. So it actually fell short of its design goal.
CANDU, RAPS-2 and MAPS-1/2 are all the same reactors basically. Our first true-blue reactor with indigenous technology was the NAPS-1 and everything that followed after. So I don't know why you have given so much importance to MAPS. And NAPS-1 started working well only from its 5th year.
Otoh, AP-1000 suffered a lot of teething problems and cost overruns to get to where it is today, just barely 1 year in operation. And with just 93% capacity factor, it generates as much power as our VVER even though it has a higher power rating. The VVER is a far more successful design, including our Unit 1.
Anyway this is the performance of Unit-1 in terms of electricity supplied (in GWh):
2014 - 2542
2015 - 3213
2016 - 5823
2017 - 2889
2018 - 4373
Unit 2:
2017 - 4148
2018 - 2761
As for AP1000, it's supplied 2746 GWh in 2018-19.
Russian VVER-1000s have been generating 7000 to 8000 GWh every year. So it's a pretty well-established reactor design. Kalinin 4 achieved 100% capacity factor and produced 8700 GWh last year. So they are simply much better than us in running the reactor. But they also have much more experience. The Chinese are also no different. Their Tianwan VVER-1000s have been delivering 7000-8000 GWh every year with 90% capacity factor as well.
So in time our VVERs will also consistently deliver near full power with 80-100% capacity factors.
Even then Fukushima couldn't handle being cut off from the grid for 11 days at a stretch.
But the problem with EPR is even the French admit to safety lapses and are working on EPR-2 instead. So they themselves agree it's a problem.
It's not an obsession. There are various reasons for it.
1. Nuclear deal - Hence diplomatic compulsion. We received a lot of help getting an NSG waiver and recognition as a nuclear power. So we are going to have to pay for it through reactor orders.
2. The promise of buying LWRs have allowed us access to unlimited supplies of uranium imports. This is of direct benefit to our overall nuclear power situation.
3. Importing reactors allow us to bring in the best global practices, which we can then apply in our own programs.
4. Investment - It allows foreign companies related to nuclear research to invest in India.
5. It allows India to collaborate with other countries in future generation programs, like the ITER.
6. 4% enriched uranium is cheap. And we need nuclear power far beyond what our first two stages (only 50+GW) can deliver.
Our plan under the three stage program is to build 10GW of PHWRs in order to sustain the 2nd stage and we will achieve that goal by 2027.
Without the nuclear deal and the waiver from NSG, none of the LWRs would have come to India.
Exactly the reason - MAPS was not our design, or our tinkering of CANDU. Similar to KKNPP in 2013. Had well established counterparts. And secondly what I am comparing is the frequent reactor triping and not the maturity of design. During startup multiple systems can trip up the system, from conventional island. At times from Nuclear side of things too. But this sort of frequent outages caused by triping of Emergency Core protection systems of the reactor - essentially we couldn't control the chain reaction properly and we were testing the envelope of safe operation. You may very well compare NAPS and its bad record, but please keep in mind, they had design changes made by us(integral end shield-calendria iirc, etc) and it was part of a learning curve, whereas VVER are a fully tested out design as you mentioned and Indian operators were trained in a Russian facility, so I was just asking, why the glitch, at such a crucial phase. Feel free to compare with RAPS 2, if that suits the discussion better and If it proves Im wrong, well I am ready to eat the humble pie. MAPS because, simply I looked at its data during those days, so purely arbitrary.
BTW let me be clear, I have nothing against VVER, they are a good design, though the excess of active systems make it less better than others from my POV, thats it. And I question the quality aspects based on initial operational experience in India. I am more than happy they are running OK now. While doing so, if you could plot the outages also along with net production, i believe it will give you a better view.
Did you just guess that from Net production? I believe we have been running nuclear reactors for years before. the records at kaiga and raps are testimony to our competence and training. Thereby, if we are bad at running the VVER may I blame that on Russians, who trained us poorly? Or it is just that Indians are not bad operators, but the reactor was tripping? because, we had passed the training from Russia, perfectly well.
Please note Chinese used a set of Russian operators in their initial ops to gain on-site training of their particular version, with modified cntrols from west.
1. Agreed
2. We could have doubled down on AHWR, though risky proposition.
3. So our practices are not by international standards? And especially where? In operations? Design? This is tenuous at best. Why deride our capability?
4. Foreign FDI in nuclear research in india? Can you name any ? No body shares their crown jewels, which you do know better than me
5. Except ITER in which design exercise are we collaborating? In operations aren't we already a part of Candu Owner's Group and havent our reactors at times adjudged the best performer before?
6. I would say the availability of this has lulled us into complacency. And again , given all this where is the French or US scale of Nuclear power adaption in India? What has been our progress in closing the fuel cycle since '08?
This is the same argument used against the T-90 when they failed a lot for about 2-3 years after induction. Quite literally everything has starting troubles, what matters more is when and how the fixes are made. Look at the AP1000 itself. Cost overruns do not happen on their own.
There were most definitely outages during its running, even today. The VVER should best case run for 8500 hours every year and deliver almost an equal amount in power. Ideally 7500 hours. But our VVERs have been running at between 3000 and 5000 hours every year. The problem is we do not know what's the actual cause. And you should stop going by media reports, most of it is not written by Indians or Russians in the first place.
More likely starting troubles. Chinese VVERs are 10 years older than ours, so they have a whole lot of experience. But I think they did a better job considering their average yearly operating time was 7000+ hours compared to our 3000-5000. And I don't see how having experience on our reactors will automatially mean we can run someone else's reactor. Foreign reactors will obviously have their own SOPs.
All reactors trip. That's why no reactors work 100% of the time all the time. And that's why claiming less number of tripping is actually a part of the advertisement for reactors. For example, AP1000's goal is 1 trip per year at 93% capacity factor.
Now we all know VVER's in India have been tripping a lot more, but the reasons for it won't be found in the media, especially in one-off articles.
2. Was not possible. We need many FBRs to sustain AHWR, and we have barely begun with FBRs. It's going to be 2040-50 before we go whole hog on AHWRs and more advanced designs.
3. Why do you think we are better than everyone else? I'm sure everybody uses at least some unique solutions, and it would be of great benefit to us if the best of solutions from all reactors are used in our future reactors. We can't think of everything ourselves.
4. This is a long process. With the introduction of the LWRs, we will see technology input and indigenization of the supply chain. This will gradually give birth to cooperation. The fact is the world will need our human resources surplus. So you can even expect R&D to be outsourced to India in the future, just like IT.
5. New stuff will always keep coming up. For example, we will need international cooperation in order to tap Helium-3 from the moon.
6. The opposite. We are currently building 10 indigenous PHWRs. It wouldn't be possible if we didn't have access to foreign supplies of uranium. So this has accelerated our three stage program.
India’s quest for nuclear fuel
By Saurav Jha.
Published : 10 September 2019
India has always had ambitions to become self-sufficient in nuclear fuel production. Saurav Jha reviews recent developments in the country, including plans to increase uranium production ten-fold by 2032.
The heavy water plant at Hazira achieved 101.12MT of production during 2018-19, recording its lowest ever specific energy consumption of 22.92 GJ/kg D2O (Photo Credit: DAE)
As Part of its Three-Stage Nuclear Program(TSNP), India’s Department of Atomic Energy(DAE) has been progressively scaling up front-end activities related to a uranium-238/plutonium-239 based closed cycle.
The eventual goal is al ‘self-sustaining’ thorium-232/uranium-233 closed cycle. Since India’s TSNP grew amidst years of isolation from international nuclear trade, self-reliance in the techno-industrial aspects of closed fuel cycles is a core aim. Today DAE is one of the few nuclear estates that has expertise across the range of activities that constitute the front end of the nuclear fuel cycle, including mineral exploration, mining and processing, heavy water production and fuel fabrication. India is self- sufficient in production of heavy water, zirconium alloy components and other materials, and supplies for PHWRs. India also has some uranium enrichment capability, which it is now looking to enhance via a new facility. The scaling up of front-end fuel activities is alongside the aim of putting in place at least 22.5GWe of nuclear power generation capacity by 2031.
Uranium resources
Fuel cycle facilities in India (source: DAE)
India has access to uranium imports from abroad and fourteen Nuclear Power Corporation of India Limited (NPCIL) reactors, which have a total installed capacity of 4380MWe, are currently operating on imported fuel. Units running on imported fuel are all safeguarded reactors.
Only the remaining eight ‘out of safeguards’ reactors in NPCIL’s fleet, with a total capacity of 2400MWe, are fuelled with domestically mined uranium.
Since 2008 fuel imports have gone a long way in restoring NPCIL’s power plant performance in terms of their plant load factors. Nevertheless, DAE is seeking to reduce dependence on foreign sources for fuel. Instead, it wants to increase domestic uranium production ten-fold by 2031/32, while simultaneously building up a strategic uranium reserve. In the past, DAE said that a stockpile of 15,000t uranium would suffice to provide security of supply for India’s nuclear reactor fleet in the medium term.
DAE’s Atomic Minerals Directorate for Exploration and Research (AMD) has been carrying out extensive radiometric, geo-chemical and geo-physical surveys across India to scour for new uranium deposits. Among the latest prospects being explored are a small tonnage deposit in Naktu, Uttar Pradesh and potential deposits in Betul, Madhya Pradesh. The Uranium Corporation of India Limited (UCIL), as part of agreements with AMD, is also conducting activities related to exploratory mining at Rohil in Rajasthan, Singridungri Banadungri in Jharkhand and Peddagattu in Telangana. As of May 2018, AMD had established 300,034t of in situ U3O8 reserves in 44 low- grade uranium deposits across India.
Mining and milling
As new U3O8 reserves are being explored, DAE has been looking to step up mining and ore processing activities. UCIL, which is responsible for mining and milling of uranium in India, has accordingly been on a path of expansion. In June 2019, the company revealed plans to invest in 13 new domestic mining projects with a total outlay of Rs105.70 billion ($1.52 billion). These new projects, once online, are expected to quadruple total domestic uranium production. Apart from these new projects, UCIL is also working towards capacity expansion at some existing units.
UCIL presently operates seven mines in the state of Jharkhand at Jaduguda, Bhatin, Narwapahar, Turamdih, Bagjata, Banduhurang and Mohuldih. It also has two processing plants in Jharkhand co-located with the mines at Jaduguda and Turamdih. Capacity expansion and ‘debottlenecking activities’ are under way in some of its oldest mines in Jharkhand. The company’s relatively new mine at Tummalapalle, Andhra Pradesh, is also being augmented for greater efficiency. This mine has achieved full production capacity and a co-located processing plant has also been delivering output.
UCIL’s processing plants at Jaduguda and Turamdih use standard acid leach techniques for the production of yellow cake or magnesium diuranate (MDU). However, the plant at Tummalapalle, which has a peak capacity of about 3000t/d, uses an alkaline pressure leach process technology to produce sodium diuranate (SDU). Recently, a re-dissolution system (RDS) facility at the Tummalapalle processing plant became operational. Commissioning marked the culmination of several years of work by DAE scientists seeking a useful technique for settling and complete recovery of precipitated product by sending a part of it to precipitation tanks. A new leaching process that can be used to obtain yellowcake from crude SDU- phosphoric acid leach solution has also been developed.
Nuclear Fuel Complex
The output of UCIL’s processing plants, whether MDU or SDU, ends up at the Nuclear Fuel Complex (NFC), Hyderabad, which is India’s key fuel fabrication facility. Any uranium imports, whether in the form of MDU, enriched uranium hexafluoride (UF6) or uranium dioxide (UO2) pellets (both enriched and unenriched) are also sent to NFC for fuel fabrication. This has put several of its units under safeguards since 2008. NFC first converts and refines MDU into UO2 powder which is then pelletised. The pellets are put into elements which are then assembled to form PHWR fuel bundles by using contemporary welding, machining and assembly techniques. In the past, NFC has fabricated PHWR fuel of varying designs, such as the 19-element wire wrap, 19-element split spacer, 22-element split spacer and a 37-element split spacer, all of which contain natural UO2 pellets.
In recent times, NFC’s main unit at Hyderabad has been churning out record amounts of PHWR fuel on a yearly basis. In 2018, NFC produced and handed over its millionth PHWR fuel-bundle to NPCIL. The fiscal year 2018/19 also saw NFC recording its highest ever production of 37-element production — 365t. For NFC, another notable achievement was the supply of 37-element fuel bundles of modified bearing pad design to the 700MWe Kakrapar 3, which is scheduled to begin commercial operation by 2022.
By the mid-2020s, NFC seems set to further augment its PHWR fuel-fabrication capacity by commissioning a new complex at Kota, Rajasthan. This greenfield facility, called NFC-Kota, will initially be able to fabricate 500t/yr of UO2 pellets and 65t/yr of Zircaloy products. Foundation works for this new facility have been completed and work on plant and administrative buildings is underway. Meanwhile, the Enriched Fuel Fabrication Plant at NFC, Hyderabad, which is used to fabricate 36 and 49-rod fuel assemblies with enrichment levels of 2.66%, 2.1% and 1.6%, respectively, for NPCIL’s two BWRs, Tarapur 1&2, has begun delivering about a hundred fuel-assemblies on an annual basis. This facility is also used for de-canning rejected PHWR fuel elements to recover UO2 pellets to be loaded into fresh fuel tubes.
Uranium enrichment capability
Over time, DAE intends to supply this facility with domestically enriched uranium. India currently operates a small enrichment plant at Ratenhalli near Mysore in Karnataka, which uses gas centrifuge technology, and is primarily meant for military purposes. However, its capacity is being expanded to 25,000 separative work units per year (SWU/yr) and it does provide limited quantities of enriched compounds to the research and power generation programmes. In the near future, it is likely to supply some slightly enriched uranium (SEU) for India’s PHWRs as well.
A much larger enrichment plant, called the Special Material Enrichment Facility (SMEF) is under construction at Challakere, Karnataka. It too will use mature gas centrifuge technology. SMEF is expected to contribute in a major way to India’s power generation programme. DAE is also pursuing R&D on laser enrichment techniques.
Zirconium
The steady growth in PHWR fuel production mentioned above necessitates a commensurate increase in output from NFC’s other major stream of nuclear related activity, the so-called ‘zirconium stream’. This is manufacturing of various zircaloy clad tubes and components used to create fuel assemblies, through a series of steps that begins with the conversion of zirconium sand, supplied by the DAE controlled Indian Rare Earths Limited (IREL) to nuclear grade zirconium oxide (ZrO2) powder.
NFC’s Zirconium Oxide Plant (ZOP) in Hyderabad has been undergoing steady modernisation and recently completed the commissioning of a high-capacity pulveriser unit, which has enhanced its capacity for grinding ZrO2 to 500kg/hr. ZOP’s control and instrumentation systems are also being upgraded. Meanwhile, NFC’s zirconium unit, the so-called Zirconium Complex (ZC), at Pazhayakayal, Tamil Nadu, has been steadily scaling up output of zirconium sponge and is likely to attain its design capacity of 250t/yr in the near future.
Heavy water production
A critical aspect for the PHWR fleet expansion is the uninterrupted supply of heavy water (D2O). DAE’s Heavy Water Board (HWB) is today the world’s largest producer of D2O and its plants have been exceeding annual production targets. In 2018, its largest plant, at Manuguru, completed a major turnaround, as did the plant at Kota. An important factor that determines viability of HWB’s plants is their specific energy consumption, which was found to be 27.9GJ/kg D2O in 2017-18 — below what HWB had budgeted for. At Tuticorin, work is currently underway on a solvent production plant and a solvent extraction plant both of which will be dovetailed to the production of special solvents including organ-phosphorous compounds.
HWB also produces critical inputs for India’s FBR programmes, such as nuclear-grade sodium for coolant purposes and enriched boron. Its boron enrichment and boron carbide pellet facilities, at HWB’s Talcher and Manuguru sites, have already delivered the entire requirement for the first core of the 500MWe Prototype Fast Breeder Reactor located at Kalpakkam. HWB has also developed closed-cell technology to produce sodium, which is used as a coolant in the PFBR. At the moment, work on a 24kA prototype cell is under way. The experience will be used to eventually set up a 600t/yr sodium production plant at Baroda.
PFBR fuel
NFC’s Fast Reactor Facility (FRF) is currently building ten different types of core sub-assemblies for the PFBR. D9 austenitic stainless steel has been used for the fuel clad tubes and the hexagonal channels, while SS 316 low nitrogen has been used to manufacture bulk components. One noteworthy feature of the hexagonal channels manufactured at NFC is that they are of the seamless variety, while elsewhere in the world these channels are usually seam welded. NFC Hyderabad has also been developing metallic fuels with short doubling time for use in India’s future FBRs. Engineering scale production of Nat-U metal power has already been demonstrated and a uranium metal production facility (UMPF) is set to be built in Vishakapatnam in the 2020s.
Author information: Saurav Jha, Author and commentator on energy and security, based in New Delhi.
India’s quest for nuclear fuel - Nuclear Engineering International
By Saurav Jha.
Published : 10 September 2019
India has always had ambitions to become self-sufficient in nuclear fuel production. Saurav Jha reviews recent developments in the country, including plans to increase uranium production ten-fold by 2032.
The heavy water plant at Hazira achieved 101.12MT of production during 2018-19, recording its lowest ever specific energy consumption of 22.92 GJ/kg D2O (Photo Credit: DAE)
As Part of its Three-Stage Nuclear Program(TSNP), India’s Department of Atomic Energy(DAE) has been progressively scaling up front-end activities related to a uranium-238/plutonium-239 based closed cycle.
The eventual goal is al ‘self-sustaining’ thorium-232/uranium-233 closed cycle. Since India’s TSNP grew amidst years of isolation from international nuclear trade, self-reliance in the techno-industrial aspects of closed fuel cycles is a core aim. Today DAE is one of the few nuclear estates that has expertise across the range of activities that constitute the front end of the nuclear fuel cycle, including mineral exploration, mining and processing, heavy water production and fuel fabrication. India is self- sufficient in production of heavy water, zirconium alloy components and other materials, and supplies for PHWRs. India also has some uranium enrichment capability, which it is now looking to enhance via a new facility. The scaling up of front-end fuel activities is alongside the aim of putting in place at least 22.5GWe of nuclear power generation capacity by 2031.
Uranium resources
Fuel cycle facilities in India (source: DAE)
India has access to uranium imports from abroad and fourteen Nuclear Power Corporation of India Limited (NPCIL) reactors, which have a total installed capacity of 4380MWe, are currently operating on imported fuel. Units running on imported fuel are all safeguarded reactors.
Only the remaining eight ‘out of safeguards’ reactors in NPCIL’s fleet, with a total capacity of 2400MWe, are fuelled with domestically mined uranium.
Since 2008 fuel imports have gone a long way in restoring NPCIL’s power plant performance in terms of their plant load factors. Nevertheless, DAE is seeking to reduce dependence on foreign sources for fuel. Instead, it wants to increase domestic uranium production ten-fold by 2031/32, while simultaneously building up a strategic uranium reserve. In the past, DAE said that a stockpile of 15,000t uranium would suffice to provide security of supply for India’s nuclear reactor fleet in the medium term.
DAE’s Atomic Minerals Directorate for Exploration and Research (AMD) has been carrying out extensive radiometric, geo-chemical and geo-physical surveys across India to scour for new uranium deposits. Among the latest prospects being explored are a small tonnage deposit in Naktu, Uttar Pradesh and potential deposits in Betul, Madhya Pradesh. The Uranium Corporation of India Limited (UCIL), as part of agreements with AMD, is also conducting activities related to exploratory mining at Rohil in Rajasthan, Singridungri Banadungri in Jharkhand and Peddagattu in Telangana. As of May 2018, AMD had established 300,034t of in situ U3O8 reserves in 44 low- grade uranium deposits across India.
Mining and milling
As new U3O8 reserves are being explored, DAE has been looking to step up mining and ore processing activities. UCIL, which is responsible for mining and milling of uranium in India, has accordingly been on a path of expansion. In June 2019, the company revealed plans to invest in 13 new domestic mining projects with a total outlay of Rs105.70 billion ($1.52 billion). These new projects, once online, are expected to quadruple total domestic uranium production. Apart from these new projects, UCIL is also working towards capacity expansion at some existing units.
UCIL presently operates seven mines in the state of Jharkhand at Jaduguda, Bhatin, Narwapahar, Turamdih, Bagjata, Banduhurang and Mohuldih. It also has two processing plants in Jharkhand co-located with the mines at Jaduguda and Turamdih. Capacity expansion and ‘debottlenecking activities’ are under way in some of its oldest mines in Jharkhand. The company’s relatively new mine at Tummalapalle, Andhra Pradesh, is also being augmented for greater efficiency. This mine has achieved full production capacity and a co-located processing plant has also been delivering output.
UCIL’s processing plants at Jaduguda and Turamdih use standard acid leach techniques for the production of yellow cake or magnesium diuranate (MDU). However, the plant at Tummalapalle, which has a peak capacity of about 3000t/d, uses an alkaline pressure leach process technology to produce sodium diuranate (SDU). Recently, a re-dissolution system (RDS) facility at the Tummalapalle processing plant became operational. Commissioning marked the culmination of several years of work by DAE scientists seeking a useful technique for settling and complete recovery of precipitated product by sending a part of it to precipitation tanks. A new leaching process that can be used to obtain yellowcake from crude SDU- phosphoric acid leach solution has also been developed.
Nuclear Fuel Complex
The output of UCIL’s processing plants, whether MDU or SDU, ends up at the Nuclear Fuel Complex (NFC), Hyderabad, which is India’s key fuel fabrication facility. Any uranium imports, whether in the form of MDU, enriched uranium hexafluoride (UF6) or uranium dioxide (UO2) pellets (both enriched and unenriched) are also sent to NFC for fuel fabrication. This has put several of its units under safeguards since 2008. NFC first converts and refines MDU into UO2 powder which is then pelletised. The pellets are put into elements which are then assembled to form PHWR fuel bundles by using contemporary welding, machining and assembly techniques. In the past, NFC has fabricated PHWR fuel of varying designs, such as the 19-element wire wrap, 19-element split spacer, 22-element split spacer and a 37-element split spacer, all of which contain natural UO2 pellets.
In recent times, NFC’s main unit at Hyderabad has been churning out record amounts of PHWR fuel on a yearly basis. In 2018, NFC produced and handed over its millionth PHWR fuel-bundle to NPCIL. The fiscal year 2018/19 also saw NFC recording its highest ever production of 37-element production — 365t. For NFC, another notable achievement was the supply of 37-element fuel bundles of modified bearing pad design to the 700MWe Kakrapar 3, which is scheduled to begin commercial operation by 2022.
By the mid-2020s, NFC seems set to further augment its PHWR fuel-fabrication capacity by commissioning a new complex at Kota, Rajasthan. This greenfield facility, called NFC-Kota, will initially be able to fabricate 500t/yr of UO2 pellets and 65t/yr of Zircaloy products. Foundation works for this new facility have been completed and work on plant and administrative buildings is underway. Meanwhile, the Enriched Fuel Fabrication Plant at NFC, Hyderabad, which is used to fabricate 36 and 49-rod fuel assemblies with enrichment levels of 2.66%, 2.1% and 1.6%, respectively, for NPCIL’s two BWRs, Tarapur 1&2, has begun delivering about a hundred fuel-assemblies on an annual basis. This facility is also used for de-canning rejected PHWR fuel elements to recover UO2 pellets to be loaded into fresh fuel tubes.
Uranium enrichment capability
Over time, DAE intends to supply this facility with domestically enriched uranium. India currently operates a small enrichment plant at Ratenhalli near Mysore in Karnataka, which uses gas centrifuge technology, and is primarily meant for military purposes. However, its capacity is being expanded to 25,000 separative work units per year (SWU/yr) and it does provide limited quantities of enriched compounds to the research and power generation programmes. In the near future, it is likely to supply some slightly enriched uranium (SEU) for India’s PHWRs as well.
A much larger enrichment plant, called the Special Material Enrichment Facility (SMEF) is under construction at Challakere, Karnataka. It too will use mature gas centrifuge technology. SMEF is expected to contribute in a major way to India’s power generation programme. DAE is also pursuing R&D on laser enrichment techniques.
Zirconium
The steady growth in PHWR fuel production mentioned above necessitates a commensurate increase in output from NFC’s other major stream of nuclear related activity, the so-called ‘zirconium stream’. This is manufacturing of various zircaloy clad tubes and components used to create fuel assemblies, through a series of steps that begins with the conversion of zirconium sand, supplied by the DAE controlled Indian Rare Earths Limited (IREL) to nuclear grade zirconium oxide (ZrO2) powder.
NFC’s Zirconium Oxide Plant (ZOP) in Hyderabad has been undergoing steady modernisation and recently completed the commissioning of a high-capacity pulveriser unit, which has enhanced its capacity for grinding ZrO2 to 500kg/hr. ZOP’s control and instrumentation systems are also being upgraded. Meanwhile, NFC’s zirconium unit, the so-called Zirconium Complex (ZC), at Pazhayakayal, Tamil Nadu, has been steadily scaling up output of zirconium sponge and is likely to attain its design capacity of 250t/yr in the near future.
Heavy water production
A critical aspect for the PHWR fleet expansion is the uninterrupted supply of heavy water (D2O). DAE’s Heavy Water Board (HWB) is today the world’s largest producer of D2O and its plants have been exceeding annual production targets. In 2018, its largest plant, at Manuguru, completed a major turnaround, as did the plant at Kota. An important factor that determines viability of HWB’s plants is their specific energy consumption, which was found to be 27.9GJ/kg D2O in 2017-18 — below what HWB had budgeted for. At Tuticorin, work is currently underway on a solvent production plant and a solvent extraction plant both of which will be dovetailed to the production of special solvents including organ-phosphorous compounds.
HWB also produces critical inputs for India’s FBR programmes, such as nuclear-grade sodium for coolant purposes and enriched boron. Its boron enrichment and boron carbide pellet facilities, at HWB’s Talcher and Manuguru sites, have already delivered the entire requirement for the first core of the 500MWe Prototype Fast Breeder Reactor located at Kalpakkam. HWB has also developed closed-cell technology to produce sodium, which is used as a coolant in the PFBR. At the moment, work on a 24kA prototype cell is under way. The experience will be used to eventually set up a 600t/yr sodium production plant at Baroda.
PFBR fuel
NFC’s Fast Reactor Facility (FRF) is currently building ten different types of core sub-assemblies for the PFBR. D9 austenitic stainless steel has been used for the fuel clad tubes and the hexagonal channels, while SS 316 low nitrogen has been used to manufacture bulk components. One noteworthy feature of the hexagonal channels manufactured at NFC is that they are of the seamless variety, while elsewhere in the world these channels are usually seam welded. NFC Hyderabad has also been developing metallic fuels with short doubling time for use in India’s future FBRs. Engineering scale production of Nat-U metal power has already been demonstrated and a uranium metal production facility (UMPF) is set to be built in Vishakapatnam in the 2020s.
Author information: Saurav Jha, Author and commentator on energy and security, based in New Delhi.
India’s quest for nuclear fuel - Nuclear Engineering International
Another Sterlite waiting to happen :
Opposition parties in Andhra gear up to protest against uranium mining in Kurnool
Opposition parties in Andhra gear up to protest against uranium mining in Kurnool
Unit 1 had frequent trips, unit 2 comparitively few, but yes. Let me search the old load dispatch data, if available, to post here, as I lost my collection LDC spreadsheets from days back
It is a tedious task to dig out unit wise data regarding unit outrages as it is buried deep inside "daily reports" updated by Southey Region Load Dispatch Centre". I am adding the link below, if at all anyone is curious.
SRLDC Daily Reports
@Gautam sir, here is a news report to give a general idea, though report quality leaves much to be desired.
Pioneer article
SRLDC Daily Reports
@Gautam sir, here is a news report to give a general idea, though report quality leaves much to be desired.
Pioneer article
Delay and decay
30 October 2019
Saurav Jha looks at India’s approach to radioactive waste management.
INDIA’S PECULIAR CLOSED NUCLEAR FUEL cycle guides its approach towards radioactive waste management. It aims to recover as much useful nuclear material (such as fissile fuel and other radionuclides) as possible from the waste generated at various stages of the fuel cycle. This is in keeping with the primary aim of minimising the amount of high-level waste (HLW) designated for final disposal. The overall idea is to ‘delay and decay’ HLW for as long as possible.
The Nuclear Recycle Board at the Bhabha Atomic Research Centre (BARC), Trombay, (the premier research establishment of India’s Department of Atomic Energy, DAE), says that India follows a “coherent, comprehensive and consistent set of principles and standards, in line with international standards” with respect to radioactive waste management systems. DAE has developed credible indigenous capability for dealing with low, intermediate and high-level waste obtained from various stages of the nuclear fuel cycle (see Figure 1). Low and intermediate level wastes There are clear practices and mature processes to deal with low and intermediate level waste (LLW and ILW). For liquid waste, the typical method is storage if it contains short half-life radionuclides, before it is diluted and discharged into selected water bodies. In waste streams that require treatment, processes such as chemical precipitation, ion exchange, evaporation and reverse osmosis are employed, either on a standalone basis or in combination, depending on the nature, composition and level of contamination. The objective is to concentrate the bulk of the activity in a small volume, before discharging the remainder into a water body. The discharge is only a small fraction of permissible limits. The radioactive concentrate is conditioned and immobilised in highly durable matrices. For instance, ILW generated during spent fuel reprocessing, is first stored in underground tanks. It is alkaline in nature, with more than 99% of the activity contributed by Caesium-137, so this waste is treated using a Cs-selective ion-exchange process using an indigenously developed Resorcinol formaldehyde resin. The process partitions the ILW into two streams: a Cs-rich ‘eluate’ stream and an effluent stream. The eluate, being HLW, is immobilised in a glass matrix, while the LLW effluent is pumped to an effluent treatment plant. This is done at a permanent facility established in 2013 at the Waste Immobilization Plant (WIP) in Trombay. DAE says that substantial decontamination and volume reduction factors (VRFs) have been achieved by this process, compared with an earlier method using bituminisation and cementation which proved unsuitable for the increasing amount of ILW generated.
India also has several solid waste management facilities with facilities for segregation, repacking and processing. For combustible LLW incineration is used wherever possible. For instance, in 2018 DAE announced that it had developed a 30kW hafnium electrode-based air-plasma torch for solid LLW treatment, which had achieved a VRF of about 30. This torch was also used to dispose of a combination of cellulosic and rubber waste.
For solid LLW and ILW that cannot be incinerated, mechanical compaction is used to reduce the waste volumes and it is then placed in near-surface disposal facilities (NSDFs) which, as a matter of national policy, are co-located with nuclear installations in India. These use a multi-barrier approach to isolate and confine the wastes, which are placed in reinforced concrete trenches or tile bores. India has considerable experience in setting up such facilities for varied geological and climatological conditions. They all have 24-hour monitoring and surveillance systems.
For gaseous LLW and ILW, various types of scrubbers have been used. Off-gases are brought into contact with suitable liquid media so as to retain the activity in the liquid phase, according to BARC. Adsorbers or absorbers are also used to remove radionuclides such as ruthenium from the gases. Prior to final release, irrespective of their initial treatment, off-gases are routed through highefficiency particulate air filters which are designed with an efficiency of greater than 99.9% for sub-micron particles.
Strategy for HLW management
Dealing with LLW and ILW is obviously very important, but 99% of the activity is in HLW streams. According to DAE, India manages HLW by immobilising liquid HLW into vitrified borosilicate glass. It is placed in engineered interim storage with other HLW with passive cooling and surveillance and eventually will be moved to final disposal in a deep geological repository. But spent fuel is first subject to reprocessing, so it is in practice a key part of India’s radioactive waste management architecture.
Spent fuel management
The journey for managing HLW generated from reactor operations in India, as elsewhere, begins with spent fuel storage facilities. India has a sizeable spent fuel inventory commensurate with the growth and vintage of nuclear fuel use, which is stored either ‘at-reactor’ or ‘away-from reactor’ before being sent for reprocessing. DAE has more than forty years of experience in operating such facilities and their design, construction and in meeting international safety standards — especially ‘wet-type’ facilities.
All new PHWRs will have spent fuel pools with a storage capacity of 10 reactor years. Stores at reprocessing plants are much smaller than the PHWR fuel pools, as they are meant to meet reprocessing operational requirements. The design of the store is based on the guidelines given in IAEA’s TECDOC-1250, which lays out safety classifications of system and components for nuclear fuel cycle facilities. All Indian spent fuel stores are designed for operating basis earthquake and the design life of the civil structure is expected to be 50 calendar years. Each has a singlefailure-proof EOT crane of 75Mt capacity, which can handle 70Mt shipping casks.
Reprocessing
Spent fuel is considered for reprocessing after three years in storage. It is transferred to a spent fuel storage pool in the fuel handling area of a reprocessing plant. India has reprocessed over 250t of spent fuel using the wellestablished hydrometallurgical Purex method, which has been used by DAE for more than 40 years. The three main operating plants are at Trombay, Tarapur and Kalpakkam.
The 60t/yr Trombay facility reprocesses aluminium-clad spent fuel from research reactors and has traditionally been used for military purposes. Tarapur and Kalpakkam, each with a capacity of 100t/yr, process zircaloy-clad oxide fuels from PHWRs. The legacy plant at Tarapur called Power Reactor Fuel Reprocessing (Prepfre) was replaced by a new facility called Prefre-2 in 2010, which shares the spent fuel pool, ADU conversion facility and utility services with its predecessor. Prefre-2 has a row of five process cells and is designed to process spent fuel from 220MWe PHWRs with an average burnup of 7000MWd/t and a cooling period of more than three years. This new unit has redundancy in safety related equipment and components, defence in depth philosophy, fail-safe logic and uses remote operation and maintenance.
Prefre-2 builds on both the design maturity of the Kalpakkam Reprocessing Plant (KARP) and also the safety lessons learnt from the accident which put KARP out of commission in the period 2003-2009. Now refurbished, KARP is back in operation and its capacity has been doubled by the addition of Prefre-3A (now called KARP-II), with a matching increment in the capacity of the adjacent waste plant WIP-3A. The head-end systems for KARP-II were designed to operate the plant at higher throughput, be more operator-friendly and allow for easier remote maintenance. Overall, India continues to make efforts further improve its reprocessing plants in terms of process flow, equipment and automation in order to increase throughput while enhancing safety.
DAE says that its reprocessing units have achieved substantial reduction in waste volume over the years by using salt-free reagents. These plants also use evaporation followed by acid reduction by formaldehyde to reduce the volume of HLW. India’s experience with the Purex process has given DAE the confidence that this technology can be successfully employed for the recovery of both uranium and plutonium with yields exceeding 99.5% — in line with international performance benchmarks.
To treble India’s current reprocessing capability and move things to an industrial scale, construction has begun at Tarapur on an integrated nuclear recycle plant (INRP) encompassing reprocessing and waste management. It will reprocess spent fuel from PHWRs and LWRs. Meanwhile, at Kalpakkam the construction of a fast reactor fuel cycle facility is gaining momentum. Its fuel reprocessing plant will have a plutonium processing section comprising eight concrete-shielded process cells.
India has also developed a method for efficient recycling of rejected MOX fuel by ‘microwave direct de-nitration’. In the recent past, more than 3t of rejected PFBR MOX fuel was recycled using this technique.
An engineering-scale facility at Trombay, the Uranium Thorium Separation Facility, is used on a regular basis to recover U-233 from ThO2 rods irradiated in the Dhruva research reactor. A much larger Power Reactor Thoria Reprocessing Facility, designed to handle the high gamma radiation associated with U-232 is also operating at Trombay and it recently completed reprocessing its second batch of ThO2. The recovered U-233 was used in the AHWR Critical Facility.
Minor actinide partitioning and transmutation
DAE sees the liquid HLW generated from reprocessing as a resource. Accordingly, at Tarapur BARC has an engineering scale Actinide Separation Demonstration Facility (ASDF) in operation. It has already demonstrated alpha separation from HLW to an extent of more than 99.9 percent at a throughput of 35 litres per hour (l/hr).
ASDF uses three distinct solvent extraction cycles. Purex is used first to separate uranium and plutonium from concentrated HLW, then the Truex-CMPO process is used to separate the bulk of MAs along with rare earths. Finally an indigenously modified Talspeak process removes trivalent actinides from lanthanides. The modified Talspeak process solved the issue of solvent extraction of americium-241 (Am-241) which TBP is unable to remove. An integrated spent solvent management facility manages the solvents spent in the partitioning process. ASDF’s operating conditions can be maintained for runs lasting 48–115 hours at an average throughput of 35 l/hr of liquid HLW.
The WIP in Trombay also uses the partitioning technology developed at ASDF. India’s actinide separation strategy recovers useful fission products such as Cs-137 and Sr-90 before final disposal.
Vitrification
India has proven industrial scale capability for vitrification and interim storage of liquid HLW from its reprocessing facilities. Vitrification is carried out in the three WIPs at Trombay, Tarapur and Kalpakkam.
Prior to commencing the vitrification process, liquid HLW is concentrated via evaporation and stored in underground stainless steel tanks, which are actively cooled and under continuous surveillance. Subsequently, the pre-concentrated liquid HLW is immobilised in various types of borosilicate glass matrix, depending on compositional changes in the waste. DAE prefers borosilicate glass for its optimal waste loading, leach resistance and long-term stability. Having said that, research into phosphate-based vitrification for HLW discharged by fast breeder reactors is also under way.
Both metallic and ceramic melters have been successfully deployed on an industrial scale by DAE. Joule Heated Ceramic Melters (JHCMs) are currently the mainstay but cold crucible induction melting (CCIM) technology has also been developed and an engineering-scale facility with a throughput of 15 l/hr is in operation at BARC, Trombay. CCIM is expected to address requirements such as high temperature availability, high waste loading and compatibility with new matrices like glass-ceramic.
India’s WIPs are being progressively upgraded with better remote handling facilities. Recently, a new second-generation advanced servo manipulator was installed at a hot cell in the vitrification bay of Trombay’s WIP. This has improvements such as force reflection capabilities, reconfigurable arm configuration, higher payload and digital control, all of which have made the cell’s remote handling operations safer and more effective.
Vitrified waste from the WIPs is stored at India’s main solid storage and surveillance facility (SSSF) at Tarapur, which has been operation for the last 20 years. Within the facility, the vitrified waste is in overpacks consisting of two stainless steel canisters. Each stainless steel canister holds about 100kg of product. Long term studies, including leach-rate experiments under simulated conditions, have been conducted in specially designed hot cells at the facility. Some of the experiments lasted for more than 700 days. The data thus generated are being used to develop models predicting the release of radionuclides from the vitrified waste over time.
These experiments are part of the R&D being undertaken in India for creation of a geological disposal facility (GDF) for final disposal of HLW. With regards to the GDF, India has made progress with respect to natural barrier characterisation, numerical modelling, conceptual design, associated instrumentation for measurements and monitoring and characterising natural analogues of waste forms and repository processes. Granite has been evaluated as a host rock and may have emerged as a leading candidate. The GDF will have a system of multiple barriers for waste isolation.
Overall, India’s GDF-related R&D has made it an active international partner in shaping the discourse on site-selection criteria for geological repositories.
Delay and decay - Nuclear Engineering International
30 October 2019
Saurav Jha looks at India’s approach to radioactive waste management.
INDIA’S PECULIAR CLOSED NUCLEAR FUEL cycle guides its approach towards radioactive waste management. It aims to recover as much useful nuclear material (such as fissile fuel and other radionuclides) as possible from the waste generated at various stages of the fuel cycle. This is in keeping with the primary aim of minimising the amount of high-level waste (HLW) designated for final disposal. The overall idea is to ‘delay and decay’ HLW for as long as possible.
The Nuclear Recycle Board at the Bhabha Atomic Research Centre (BARC), Trombay, (the premier research establishment of India’s Department of Atomic Energy, DAE), says that India follows a “coherent, comprehensive and consistent set of principles and standards, in line with international standards” with respect to radioactive waste management systems. DAE has developed credible indigenous capability for dealing with low, intermediate and high-level waste obtained from various stages of the nuclear fuel cycle (see Figure 1). Low and intermediate level wastes There are clear practices and mature processes to deal with low and intermediate level waste (LLW and ILW). For liquid waste, the typical method is storage if it contains short half-life radionuclides, before it is diluted and discharged into selected water bodies. In waste streams that require treatment, processes such as chemical precipitation, ion exchange, evaporation and reverse osmosis are employed, either on a standalone basis or in combination, depending on the nature, composition and level of contamination. The objective is to concentrate the bulk of the activity in a small volume, before discharging the remainder into a water body. The discharge is only a small fraction of permissible limits. The radioactive concentrate is conditioned and immobilised in highly durable matrices. For instance, ILW generated during spent fuel reprocessing, is first stored in underground tanks. It is alkaline in nature, with more than 99% of the activity contributed by Caesium-137, so this waste is treated using a Cs-selective ion-exchange process using an indigenously developed Resorcinol formaldehyde resin. The process partitions the ILW into two streams: a Cs-rich ‘eluate’ stream and an effluent stream. The eluate, being HLW, is immobilised in a glass matrix, while the LLW effluent is pumped to an effluent treatment plant. This is done at a permanent facility established in 2013 at the Waste Immobilization Plant (WIP) in Trombay. DAE says that substantial decontamination and volume reduction factors (VRFs) have been achieved by this process, compared with an earlier method using bituminisation and cementation which proved unsuitable for the increasing amount of ILW generated.
India also has several solid waste management facilities with facilities for segregation, repacking and processing. For combustible LLW incineration is used wherever possible. For instance, in 2018 DAE announced that it had developed a 30kW hafnium electrode-based air-plasma torch for solid LLW treatment, which had achieved a VRF of about 30. This torch was also used to dispose of a combination of cellulosic and rubber waste.
For solid LLW and ILW that cannot be incinerated, mechanical compaction is used to reduce the waste volumes and it is then placed in near-surface disposal facilities (NSDFs) which, as a matter of national policy, are co-located with nuclear installations in India. These use a multi-barrier approach to isolate and confine the wastes, which are placed in reinforced concrete trenches or tile bores. India has considerable experience in setting up such facilities for varied geological and climatological conditions. They all have 24-hour monitoring and surveillance systems.
For gaseous LLW and ILW, various types of scrubbers have been used. Off-gases are brought into contact with suitable liquid media so as to retain the activity in the liquid phase, according to BARC. Adsorbers or absorbers are also used to remove radionuclides such as ruthenium from the gases. Prior to final release, irrespective of their initial treatment, off-gases are routed through highefficiency particulate air filters which are designed with an efficiency of greater than 99.9% for sub-micron particles.
Strategy for HLW management
Dealing with LLW and ILW is obviously very important, but 99% of the activity is in HLW streams. According to DAE, India manages HLW by immobilising liquid HLW into vitrified borosilicate glass. It is placed in engineered interim storage with other HLW with passive cooling and surveillance and eventually will be moved to final disposal in a deep geological repository. But spent fuel is first subject to reprocessing, so it is in practice a key part of India’s radioactive waste management architecture.
Spent fuel management
The journey for managing HLW generated from reactor operations in India, as elsewhere, begins with spent fuel storage facilities. India has a sizeable spent fuel inventory commensurate with the growth and vintage of nuclear fuel use, which is stored either ‘at-reactor’ or ‘away-from reactor’ before being sent for reprocessing. DAE has more than forty years of experience in operating such facilities and their design, construction and in meeting international safety standards — especially ‘wet-type’ facilities.
All new PHWRs will have spent fuel pools with a storage capacity of 10 reactor years. Stores at reprocessing plants are much smaller than the PHWR fuel pools, as they are meant to meet reprocessing operational requirements. The design of the store is based on the guidelines given in IAEA’s TECDOC-1250, which lays out safety classifications of system and components for nuclear fuel cycle facilities. All Indian spent fuel stores are designed for operating basis earthquake and the design life of the civil structure is expected to be 50 calendar years. Each has a singlefailure-proof EOT crane of 75Mt capacity, which can handle 70Mt shipping casks.
Reprocessing
Spent fuel is considered for reprocessing after three years in storage. It is transferred to a spent fuel storage pool in the fuel handling area of a reprocessing plant. India has reprocessed over 250t of spent fuel using the wellestablished hydrometallurgical Purex method, which has been used by DAE for more than 40 years. The three main operating plants are at Trombay, Tarapur and Kalpakkam.
The 60t/yr Trombay facility reprocesses aluminium-clad spent fuel from research reactors and has traditionally been used for military purposes. Tarapur and Kalpakkam, each with a capacity of 100t/yr, process zircaloy-clad oxide fuels from PHWRs. The legacy plant at Tarapur called Power Reactor Fuel Reprocessing (Prepfre) was replaced by a new facility called Prefre-2 in 2010, which shares the spent fuel pool, ADU conversion facility and utility services with its predecessor. Prefre-2 has a row of five process cells and is designed to process spent fuel from 220MWe PHWRs with an average burnup of 7000MWd/t and a cooling period of more than three years. This new unit has redundancy in safety related equipment and components, defence in depth philosophy, fail-safe logic and uses remote operation and maintenance.
Prefre-2 builds on both the design maturity of the Kalpakkam Reprocessing Plant (KARP) and also the safety lessons learnt from the accident which put KARP out of commission in the period 2003-2009. Now refurbished, KARP is back in operation and its capacity has been doubled by the addition of Prefre-3A (now called KARP-II), with a matching increment in the capacity of the adjacent waste plant WIP-3A. The head-end systems for KARP-II were designed to operate the plant at higher throughput, be more operator-friendly and allow for easier remote maintenance. Overall, India continues to make efforts further improve its reprocessing plants in terms of process flow, equipment and automation in order to increase throughput while enhancing safety.
DAE says that its reprocessing units have achieved substantial reduction in waste volume over the years by using salt-free reagents. These plants also use evaporation followed by acid reduction by formaldehyde to reduce the volume of HLW. India’s experience with the Purex process has given DAE the confidence that this technology can be successfully employed for the recovery of both uranium and plutonium with yields exceeding 99.5% — in line with international performance benchmarks.
To treble India’s current reprocessing capability and move things to an industrial scale, construction has begun at Tarapur on an integrated nuclear recycle plant (INRP) encompassing reprocessing and waste management. It will reprocess spent fuel from PHWRs and LWRs. Meanwhile, at Kalpakkam the construction of a fast reactor fuel cycle facility is gaining momentum. Its fuel reprocessing plant will have a plutonium processing section comprising eight concrete-shielded process cells.
India has also developed a method for efficient recycling of rejected MOX fuel by ‘microwave direct de-nitration’. In the recent past, more than 3t of rejected PFBR MOX fuel was recycled using this technique.
An engineering-scale facility at Trombay, the Uranium Thorium Separation Facility, is used on a regular basis to recover U-233 from ThO2 rods irradiated in the Dhruva research reactor. A much larger Power Reactor Thoria Reprocessing Facility, designed to handle the high gamma radiation associated with U-232 is also operating at Trombay and it recently completed reprocessing its second batch of ThO2. The recovered U-233 was used in the AHWR Critical Facility.
Minor actinide partitioning and transmutation
DAE sees the liquid HLW generated from reprocessing as a resource. Accordingly, at Tarapur BARC has an engineering scale Actinide Separation Demonstration Facility (ASDF) in operation. It has already demonstrated alpha separation from HLW to an extent of more than 99.9 percent at a throughput of 35 litres per hour (l/hr).
ASDF uses three distinct solvent extraction cycles. Purex is used first to separate uranium and plutonium from concentrated HLW, then the Truex-CMPO process is used to separate the bulk of MAs along with rare earths. Finally an indigenously modified Talspeak process removes trivalent actinides from lanthanides. The modified Talspeak process solved the issue of solvent extraction of americium-241 (Am-241) which TBP is unable to remove. An integrated spent solvent management facility manages the solvents spent in the partitioning process. ASDF’s operating conditions can be maintained for runs lasting 48–115 hours at an average throughput of 35 l/hr of liquid HLW.
The WIP in Trombay also uses the partitioning technology developed at ASDF. India’s actinide separation strategy recovers useful fission products such as Cs-137 and Sr-90 before final disposal.
Vitrification
India has proven industrial scale capability for vitrification and interim storage of liquid HLW from its reprocessing facilities. Vitrification is carried out in the three WIPs at Trombay, Tarapur and Kalpakkam.
Prior to commencing the vitrification process, liquid HLW is concentrated via evaporation and stored in underground stainless steel tanks, which are actively cooled and under continuous surveillance. Subsequently, the pre-concentrated liquid HLW is immobilised in various types of borosilicate glass matrix, depending on compositional changes in the waste. DAE prefers borosilicate glass for its optimal waste loading, leach resistance and long-term stability. Having said that, research into phosphate-based vitrification for HLW discharged by fast breeder reactors is also under way.
Both metallic and ceramic melters have been successfully deployed on an industrial scale by DAE. Joule Heated Ceramic Melters (JHCMs) are currently the mainstay but cold crucible induction melting (CCIM) technology has also been developed and an engineering-scale facility with a throughput of 15 l/hr is in operation at BARC, Trombay. CCIM is expected to address requirements such as high temperature availability, high waste loading and compatibility with new matrices like glass-ceramic.
India’s WIPs are being progressively upgraded with better remote handling facilities. Recently, a new second-generation advanced servo manipulator was installed at a hot cell in the vitrification bay of Trombay’s WIP. This has improvements such as force reflection capabilities, reconfigurable arm configuration, higher payload and digital control, all of which have made the cell’s remote handling operations safer and more effective.
Vitrified waste from the WIPs is stored at India’s main solid storage and surveillance facility (SSSF) at Tarapur, which has been operation for the last 20 years. Within the facility, the vitrified waste is in overpacks consisting of two stainless steel canisters. Each stainless steel canister holds about 100kg of product. Long term studies, including leach-rate experiments under simulated conditions, have been conducted in specially designed hot cells at the facility. Some of the experiments lasted for more than 700 days. The data thus generated are being used to develop models predicting the release of radionuclides from the vitrified waste over time.
These experiments are part of the R&D being undertaken in India for creation of a geological disposal facility (GDF) for final disposal of HLW. With regards to the GDF, India has made progress with respect to natural barrier characterisation, numerical modelling, conceptual design, associated instrumentation for measurements and monitoring and characterising natural analogues of waste forms and repository processes. Granite has been evaluated as a host rock and may have emerged as a leading candidate. The GDF will have a system of multiple barriers for waste isolation.
Overall, India’s GDF-related R&D has made it an active international partner in shaping the discourse on site-selection criteria for geological repositories.
Delay and decay - Nuclear Engineering International
Wasn't RP Taleyarkhan, thrown out of Oakridge for want of proof and independent verification? Or is this some other methodology of attaining the "unattainable"?
Third Unit of Kakrapar Atomic Reactor to be Commissioned in April
The 700-MW pressurised heavy water reactor is likely to be commissioned by April while the fourth unit would likely to be commissioned by 2021.
By PTI, Updated: December 31, 2019, 5:57 PM IST
The 700-MW pressurised heavy water reactor is likely to be commissioned by April while the fourth unit would likely to be commissioned by 2021.
New Delhi: The Department of Atomic Energy will commission one nuclear reactor every year from 2020, starting with the third unit of the Kakrapar Atomic Power Station in Gujarat, Union minister Jitendra Singh said.
The 700-MW pressurised heavy water reactor is likely to be commissioned by April, Singh, who is a minister of state in the Prime Minister's Office, said in an interaction with reporters in New Delhi. "We will commission one nuclear reactor every year from 2020. Kakrapar-3 should be commissioned in 2020," Singh said.
A senior Department of Atomic Energy (DAE) official said the fourth unit of the power station would likely to be commissioned by 2021. The Nuclear Power Corporation of India Ltd (NPCIL) has 22 reactors.
Shrikrishna Gupta, a senior DAE official, said the Tarapur Atomic Power Reactors -- units 1 and 2 -- completed 50 years of operation this year. The two boiling water reactors -- the first in the country -- were commissioned in October 1969.
He said the Kaiga Power Station unit 1 also created a world record by operating for 941 days.
Third Unit of Kakrapar Atomic Reactor to be Commissioned in April
The 700-MW pressurised heavy water reactor is likely to be commissioned by April while the fourth unit would likely to be commissioned by 2021.
By PTI, Updated: December 31, 2019, 5:57 PM IST
The 700-MW pressurised heavy water reactor is likely to be commissioned by April while the fourth unit would likely to be commissioned by 2021.
New Delhi: The Department of Atomic Energy will commission one nuclear reactor every year from 2020, starting with the third unit of the Kakrapar Atomic Power Station in Gujarat, Union minister Jitendra Singh said.
The 700-MW pressurised heavy water reactor is likely to be commissioned by April, Singh, who is a minister of state in the Prime Minister's Office, said in an interaction with reporters in New Delhi. "We will commission one nuclear reactor every year from 2020. Kakrapar-3 should be commissioned in 2020," Singh said.
A senior Department of Atomic Energy (DAE) official said the fourth unit of the power station would likely to be commissioned by 2021. The Nuclear Power Corporation of India Ltd (NPCIL) has 22 reactors.
Shrikrishna Gupta, a senior DAE official, said the Tarapur Atomic Power Reactors -- units 1 and 2 -- completed 50 years of operation this year. The two boiling water reactors -- the first in the country -- were commissioned in October 1969.
He said the Kaiga Power Station unit 1 also created a world record by operating for 941 days.
Third Unit of Kakrapar Atomic Reactor to be Commissioned in April
I came here to post the same, but you beat me to it.
Finally the 700 mwe is nearing completion after an arduous journey.
If one observes, the 540 mwe Tarapur 3 & 4 was completed in record time. The 700 mwe is a slight modification of that same design with almost the same outer dimensions. There are changes in fuel pin config etc, but it took like ever for this set of reactors to complete construction.
This reflects poorly on the project management. Progress on indigenous reactors has been one area Modi government has been found lacking. May be they were banking on the operationalisation of nuclear deal and imports, in the truest tradition of our nation.
