Delay and decay30 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.
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.
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.
Author information: Saurav Jha is an author and commentator on energy and security, based in New Delhi