India’s research fleet
12 December 2017
Saurav Jha explains how India is using its research reactor fleet to support its three-stage nuclear programme
Research reactors are an integral part of the support infrastructure of any country with significant nuclear fuel cycle activities on its soil. India, being no exception to this rule, has built a number of research reactors over the years to support
its unique three-stage nuclear programme (TNSP). The mix of reactors operated by India’s Department of Atomic Energy (DAE) reflects the peculiar aims of the programme, which seeks to industrialise innovative closed fuel cycles with the aim of utilising India’s vast thorium reserves.
While Indian research reactors have been used in all the standard roles, including neutron radiography, neutron activation analysis, radioisotope production for medical and industrial use, neutron irradiation for materials characterisation and testing, neutron beam research and applications etc, they have also been instrumental in helping DAE validate computer codes and perform elemental analysis for prototyping next generation reactors that depend on U-238/ Pu-239 and Th-232/U-233 fuel cycles. Indian research reactors have also been useful in training scientific, maintenance, operation, radiation protection and regulatory staff.
With India’s existing nuclear fleet getting older and new reactors being built to last for much longer, the demand for material testing is set to grow and so will the need for radioisotope production, as well as new requirements such as silicon doping. As a result, DAE is involved in the design and creation of new reactor designs that will meet demand for established neutron- source-based R&D activities and emerging needs. DAE is also retrofitting its existing research reactors with a view to bringing their safety and security standards in line with contemporary requirements and making them ready to serve the dynamic demands being placed on R&D-oriented neutron sources in India.
Apsara renewed
Asia’s oldest research reactor, the now 61- year old Apsara pool-type reactor, located at the Bhabha Atomic Research Centre (BARC), Trombay, is currently undergoing a major upgrade and life-extension programme, to keep it viable as a neutron source for its traditional roles (radioisotope production, neutron activation analysis, neutron radiography, shielding studies, material irradiation and the development and testing of neutron detectors). Apsara has proved rather useful in ratifying the design adequacy of many reactors that came later, including the Dhruva and the 500MWe Prototype Fast Breeder Reactor (PFBR) in Kalpakkam that is expected to reach criticality in the coming months.
Apsara’s extensive refurbishment involves new control systems, shielding and core cooling structures and components, in line with current safety standards and codes. Post-refurbishment, Apsara’s maximum rated power will double to 2MWt and the maximum thermal neutron flux at the rated power of the reactor will rise to 6.1×1013 n/cm2/s from the earlier 1x1013 n/cm2/s. The maximum fast neutron flux of the renovated Apsara will be 1.4×1013 n/cm2/s.
Changes to Apsara’s core design have been made to ensure better neutron economy. Beryllium oxide (BeO) reflector assemblies clad in aluminium will now surround the core, in order to provide the desired level of core reactivity while sustaining high thermal neutron flux levels over a large radial distance around the reactor core, for material studies and isotope production. The use of four fast- acting hafnium shut-off rods, two of which double-up as the reactor’s control rods (and are supplemented by a hafnium fine control rod) has enhanced safety and maintained the availability of spots in the core matrix for irradiation purposes. For instance, an in-core hollow BeO plug is being provided for high neutron flux experiments in addition to seven other irradiation positions in the reflector assembly. Despite the uprating, the new reactor core design ensures negative reactivity coefficient from zero to full power levels. The earlier ‘single-channel’ reactor power regulating system is being replaced by a triple-channel proportional control system using neutron power and reactor period signals.
In keeping with global non-proliferation trends, the refitted Apsara will use low- enriched uranium (LEU) fuel instead of highly-enriched uranium (HEU). In particular, LEU (17 percent) uranium silicide dispersed in aluminium (U3Si2-Al) plate type fuel has been chosen. Its favourable features include high uranium loading density in the fuel meat, good thermal conductivity, an excellent blister resistance threshold, stable swelling behaviour under irradiation, high fission gas- retention capability and easy fabrication.
Construction of the refurbished Apsara reactor pool, annex building, pump house and dump tank is now complete, and DAE is satisfied that it meets seismic and shielding adequacy standards. Construction of the reactor hall and electrical substation is nearing completion and the renovated Apsara is likely to be re-commissioned in 2018.
Dhruva for isotopes and more
Even as Apsara’s rebirth beckons, India’s premier neutron beam research facility, the 100MWt Dhruva reactor, operational at BARC, Trombay soldiers on. It had an availability factor of around 72% and a capacity factor of around 61%, respectively in 2016. This natural uranium metal-fuelled vertical tank- type thermal reactor, commissioned in 1985, is of indigenous design and uses heavy water as coolant and moderator.
It has a large number of neutron beam tube facilities with diameters of up to 300mm and the plant has regularly operated up to its maximum rated power with a maximum thermal neutron flux of 1.8×1014 n/cm2/s. It is DAE’s chief isotope production facility. Last year alone, over 700 samples were irradiated at Dhruva for radioisotope production. At the moment, over a thousand ‘user’ institutions receive isotopic preparations (Mo-99, I-131, I-125, P-32, S-35, Cr-51, Co-60, Au-198, Br-82, Ir-192 and others) from Dhruva. The plant’s workload has gone up significantly since the 2010 decommissioning of the Cirus reactor.
Beyond radioisotope production, Dhruva is also India’s chief facility for neutron radiography. In 2016, a new Neutron Radiography and Tomography Facility (NRTF) was commissioned at Dhruva to boost reactor use. A dedicated neutron imaging beam line has been set up at Dhruva Beam-hole HS-3018 for real-time neutron imaging and neutron tomography.
NRTF consists of a dual-purpose collimator for neutron tomography and phase imaging experiments that may require high spatial coherence. Given Dhruva’s high neutron flux levels, it should be quicker to produce imaging data of superior resolution and better signal-to-noise ratio than at other facilities, meanwhile while reducing the time taken for image data acquisition.
NRTF is available for various studies including hydrogen ingress in zircaloy, examination of pressurised heavy water reactor (PHWR) fuel pins, cracks in failed turbine blades, real-time investigation of lead melting. It will meet neutron imaging demand from both DAE users as well as such as the Indian Space Research Organization (ISRO). DAE says, ‘the tomography set-up will be extended to include new imaging techniques such as magnetic phase contrast imaging using polarized neutrons’.
Dhruva is also being used to develop new fuel types via its in-pile loop facility which has higher heat removal capacity than
was available in the now decommissioned Cirus. This can accommodate fuel arrays from India’s emerging medium-sized PHWR fleet as well as that of the Advanced Heavy Water Reactor (AHWR) that uses thorium for testing purposes. In the recent past, a twisted pin geometry cluster, with dispersion-type fuel was successfully irradiated up to its research target burn-up to study of fuel performance, such as the reaction between dispersant and diluents, swelling characteristics, clad deformation behaviour, etc.
Dhruva’s existing pneumatic carrier facility, which can provide a thermal neutron flux of 9.7x1013 n/cm2/s for the irradiation of samples that yield short-lived isotopes on irradiation, continues to be in demand. Dhruva also has a prompt gamma ray neutron activation analysis system that is used for the online analysis of various materials. Dhruva’s small angle neutron scattering facilities have also been used for studies on soft condensed matter and large inhomogeneities in the past.
Supporting the FBR
While not a research reactor in the strictest sense, the Fast Breeder Test Reactor (FBTR) at the Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, which attained first criticality in 1985, has been the mothership for FBR-related R&D in India. Identical to the French Rapsodie- Fortissimo except for the addition of steam generators and a turbo-generator (the unit was connected to the grid in 1997), this 40MWt (13.5MWe) sodium-cooled loop- type reactor has been used to train a cadre of personnel. It also provided experience in fast reactor operation and large-scale sodium handling, while serving as a means to test fuels, structural materials and special neutron detectors.
Given a 20-year life-extension in 2011, FBTR is currently into its 25th irradiation campaign, which has seen the power raised to 27.3MWt for the first time in its operating history. This campaign, like the ones immediately preceding it, involved the loading of yttria-72 thoria subassemblies for irradiation in order to produce uranium-233. This campaign involves irradiation of yttria, natural U-Zr sodium-bonded metal fuel pins, uranium (14.3% enriched) metal pins, and ternary fuel pin U-Pu-Zr, as well as impact specimens of low nitrogen austenitic stainless steels that make up the grid-plate of FBTR itself.
These sub-assemblies are in addition to the fuel sub-assemblies that are at any time a part of the FBTR core. They include indigenously developed Mark-I (70% PuC + 30%UC) and Mark-II (55% PuC + 45%UC) mixed carbine fuels; and MOX (44% PuO2 + 56% UO2). More than 1200 Mark-I fuel pins have reached a burnup of 155GWd/t in FBTR over the years, and the record burnup is 165GWd/t – the highest ever attained internationally. Only one Mk-I fuel pin failure event has been observed up to now.
FBTR’s fuel-cycle has been ‘closed’; new mixed carbide fuel was fabricated after fuel, discharged at up to 150GWd/t, was reprocessed in IGCAR’s CORAL facility. PFBR fuel under irradiation testing in FBTR has reached a burnup of 112GWd/t and has been discharged for post-irradiation examination. About a fifth of FBTR’s core is currently loaded with high-Pu MOX as part of efforts to develop higher-burnup fuel for PHWRs.
Past irradiation campaigns have included studies on D9 structural materials, testing of high-temperature fission chambers for PFBR up to 10MWt, and testing of an industrial version of a Kalman-filter-based instrument, meant for drop time measurement of the Diverse Safety Rod Drive Mechanisms (DSRDM) for PFBR.
FBTR operations have not been without incident, however. A major fuel handling incident has been recorded, as has a primary sodium leak, as well as reactivity transients. As a result, major structural modifications have been carried out in the past, including one on FBTR’s steam generator leak detection system and the other on its steam-water circuit. These changes have helped boost the reactor’s availability.
Refurbishment was carried out to support life extension. That included replacing steam generator rupture discs and main boiler feed pumps; commissioning a new 1MVA transformer along with associated switch gear to augment the existing power supply; and a new demineralised water plant. FBTR is also receiving post-Fukushima retrofits. Last year IGCAR installed a new flood-safe building for emergency diesels, and it installed and commissioned two mobile diesel generator sets for accident scenarios.
The 30kWt Kalapakkam Mini (Kamini) reactor, commissioned at IGCAR in 1996, has been used to test high-temperature fission chambers developed for PFBR up to a neutron flux of 5x109 n/cm2/s and gamma field of 5 x 105 R/hr at a temperature of 570°C, in a test assembly made up of Inconel tube which is installed on the east side of the reactor in the pool. Kamini is also used for neutron radiography of irradiated FBTR fuel.
This tank-in-pool type design is a reflector-moderated reactor where half of fission events are due to reflector-returned neutrons. Zircoloy-2 canned BeO is used as a reflector: it has high reflection efficiency, which results in lowering fuel inventory requirements. Kamini is a neutron source with a flux level of 8.0x1012 n/cm2/s at the centre of its core. It has facilities for carrying out neutron radiography of radioactive and non-radioactive objects, as well as the neutron activation analysis of a variety of samples.
Essentially a materials testing reactor, Kamini has been used by ISRO for the non- destructive examination of thousands of pyro devices used in India’s space programme.
It has also been used to irradiate space- borne sensors to test their reliability in the face of cosmic radiation. In recent years, due to the non-availability of Apsara during refurbishment, Kamini has also been used to test and calibrate neutron detectors. Owing to its design, Kamini has been unable to host major shielding experiments.
Kamini is also a test-bed for the third stage of India’s three-stage nuclear programme, given that it is the only uranium-233
fuelled reactor in use today. In the past, thoria fuel rods irradiated in Cirus have been reprocessed in the Uranium-Thorium Separation Facility (UTSF) at BARC and the recovered U-233 has been fabricated as fuel for the Kamini reactor at IGCAR. U-233 from thoria fuel bundles irradiated in FBTR by CORAL have also been used to fabricate fuel for Kamini. At the moment, Kamini is licensed for operation till June 2020, after a periodic safety review.
Critical Facility
Back in Trombay, third-stage activities are being supported by the 100W critical facility, commissioned in 2008, which can generate a thermal neutron flux of 108 n/cm2/s and has been designed to facilitate study of different core lattices based on various fuel types, moderator materials and reactivity control devices. It is being used to validate the reactor core physics for AHWR, which will utilise thorium fuel. AHWR studies began with a 54 pin (Th-Pu)MOX/(Th-233U) MOX cluster representative AHWR core. The critical facility is also being used for the testing and qualification of reactivity control devices and detectors, and in neutron activation analysis of various samples.
A new generation
India’s next generation research reactors will be constructed at BARC’s new campus in Vishakhapatnam, on the East Coast. Among the new reactors planned, the High Flux Research Reactor (HFRR) is currently in the most advanced stage of implementation. Preparation and review of preliminary safety analysis report, part-A and preliminary architectural drawings of the structures have been completed.
The 30MWt HTRR is a swimming pool type, thermal reactor design that will use U3Si2 dispersed fuel with enrichment levels of 19.75%. The reactor will be cooled and moderated by light water, with heavy water as a reflector. The maximum thermal neutron flux will be 6.7x1014 n/cm2/s and the maximum fast neutron flux will be 1.8×1014 n/cm2/s to cater to the requirements of radioisotope production, fuel and material testing as well as advanced neutron beam tube research. Laboratory-scale recovery of uranium from the silicide fuel has been established using a feed composition of reference burnup for 80GWd/t, with ten-year cooling.
Joining HFRR in the years ahead at Vishakapatnam will be the 125MWt Thermal Research Reactor (THRR), which will be
a vertical tank type reactor, rather similar in design to Dhruva. THRR will be fuelled by either natural uranium metal or slightly enriched uranium metal (1.25%) and will be able to generate a maximum thermal neutron flux of 2.2x1014 n/cm2/s. Its focus will be on irradiating various materials, although it will also have facilities for neutron beam tube research, radioisotope production, neutron activation analysis and neutron radiography.
Meanwhile, IGCAR has prepared a preliminary layout for a follow-on to the FBTR, dubbed FBTR-2, in line with its experience with PFBR. This is expected to be a 300MWt (150MWe) fast breeder reactor that will be a test-bed for metallic fuels that will power the future 1000MWe Metallic FBRs (MFBRs), which are currently at the design stage. FBTR-2 will use U-Pu alloy or U-Pu-Zr, with electrometallurgical reprocessing to close the fuel cycle.
India has already built a number of research reactors to support its three-stage nuclear programme. It is clear that this fleet will play a quiet but pivotal role in aiding India’s new nuclear programme.
12 December 2017
Saurav Jha explains how India is using its research reactor fleet to support its three-stage nuclear programme
Research reactors are an integral part of the support infrastructure of any country with significant nuclear fuel cycle activities on its soil. India, being no exception to this rule, has built a number of research reactors over the years to support
its unique three-stage nuclear programme (TNSP). The mix of reactors operated by India’s Department of Atomic Energy (DAE) reflects the peculiar aims of the programme, which seeks to industrialise innovative closed fuel cycles with the aim of utilising India’s vast thorium reserves.
While Indian research reactors have been used in all the standard roles, including neutron radiography, neutron activation analysis, radioisotope production for medical and industrial use, neutron irradiation for materials characterisation and testing, neutron beam research and applications etc, they have also been instrumental in helping DAE validate computer codes and perform elemental analysis for prototyping next generation reactors that depend on U-238/ Pu-239 and Th-232/U-233 fuel cycles. Indian research reactors have also been useful in training scientific, maintenance, operation, radiation protection and regulatory staff.
With India’s existing nuclear fleet getting older and new reactors being built to last for much longer, the demand for material testing is set to grow and so will the need for radioisotope production, as well as new requirements such as silicon doping. As a result, DAE is involved in the design and creation of new reactor designs that will meet demand for established neutron- source-based R&D activities and emerging needs. DAE is also retrofitting its existing research reactors with a view to bringing their safety and security standards in line with contemporary requirements and making them ready to serve the dynamic demands being placed on R&D-oriented neutron sources in India.
Apsara renewed
Asia’s oldest research reactor, the now 61- year old Apsara pool-type reactor, located at the Bhabha Atomic Research Centre (BARC), Trombay, is currently undergoing a major upgrade and life-extension programme, to keep it viable as a neutron source for its traditional roles (radioisotope production, neutron activation analysis, neutron radiography, shielding studies, material irradiation and the development and testing of neutron detectors). Apsara has proved rather useful in ratifying the design adequacy of many reactors that came later, including the Dhruva and the 500MWe Prototype Fast Breeder Reactor (PFBR) in Kalpakkam that is expected to reach criticality in the coming months.
Apsara’s extensive refurbishment involves new control systems, shielding and core cooling structures and components, in line with current safety standards and codes. Post-refurbishment, Apsara’s maximum rated power will double to 2MWt and the maximum thermal neutron flux at the rated power of the reactor will rise to 6.1×1013 n/cm2/s from the earlier 1x1013 n/cm2/s. The maximum fast neutron flux of the renovated Apsara will be 1.4×1013 n/cm2/s.
Changes to Apsara’s core design have been made to ensure better neutron economy. Beryllium oxide (BeO) reflector assemblies clad in aluminium will now surround the core, in order to provide the desired level of core reactivity while sustaining high thermal neutron flux levels over a large radial distance around the reactor core, for material studies and isotope production. The use of four fast- acting hafnium shut-off rods, two of which double-up as the reactor’s control rods (and are supplemented by a hafnium fine control rod) has enhanced safety and maintained the availability of spots in the core matrix for irradiation purposes. For instance, an in-core hollow BeO plug is being provided for high neutron flux experiments in addition to seven other irradiation positions in the reflector assembly. Despite the uprating, the new reactor core design ensures negative reactivity coefficient from zero to full power levels. The earlier ‘single-channel’ reactor power regulating system is being replaced by a triple-channel proportional control system using neutron power and reactor period signals.
In keeping with global non-proliferation trends, the refitted Apsara will use low- enriched uranium (LEU) fuel instead of highly-enriched uranium (HEU). In particular, LEU (17 percent) uranium silicide dispersed in aluminium (U3Si2-Al) plate type fuel has been chosen. Its favourable features include high uranium loading density in the fuel meat, good thermal conductivity, an excellent blister resistance threshold, stable swelling behaviour under irradiation, high fission gas- retention capability and easy fabrication.
Construction of the refurbished Apsara reactor pool, annex building, pump house and dump tank is now complete, and DAE is satisfied that it meets seismic and shielding adequacy standards. Construction of the reactor hall and electrical substation is nearing completion and the renovated Apsara is likely to be re-commissioned in 2018.
Dhruva for isotopes and more
Even as Apsara’s rebirth beckons, India’s premier neutron beam research facility, the 100MWt Dhruva reactor, operational at BARC, Trombay soldiers on. It had an availability factor of around 72% and a capacity factor of around 61%, respectively in 2016. This natural uranium metal-fuelled vertical tank- type thermal reactor, commissioned in 1985, is of indigenous design and uses heavy water as coolant and moderator.
It has a large number of neutron beam tube facilities with diameters of up to 300mm and the plant has regularly operated up to its maximum rated power with a maximum thermal neutron flux of 1.8×1014 n/cm2/s. It is DAE’s chief isotope production facility. Last year alone, over 700 samples were irradiated at Dhruva for radioisotope production. At the moment, over a thousand ‘user’ institutions receive isotopic preparations (Mo-99, I-131, I-125, P-32, S-35, Cr-51, Co-60, Au-198, Br-82, Ir-192 and others) from Dhruva. The plant’s workload has gone up significantly since the 2010 decommissioning of the Cirus reactor.
Beyond radioisotope production, Dhruva is also India’s chief facility for neutron radiography. In 2016, a new Neutron Radiography and Tomography Facility (NRTF) was commissioned at Dhruva to boost reactor use. A dedicated neutron imaging beam line has been set up at Dhruva Beam-hole HS-3018 for real-time neutron imaging and neutron tomography.
NRTF consists of a dual-purpose collimator for neutron tomography and phase imaging experiments that may require high spatial coherence. Given Dhruva’s high neutron flux levels, it should be quicker to produce imaging data of superior resolution and better signal-to-noise ratio than at other facilities, meanwhile while reducing the time taken for image data acquisition.
NRTF is available for various studies including hydrogen ingress in zircaloy, examination of pressurised heavy water reactor (PHWR) fuel pins, cracks in failed turbine blades, real-time investigation of lead melting. It will meet neutron imaging demand from both DAE users as well as such as the Indian Space Research Organization (ISRO). DAE says, ‘the tomography set-up will be extended to include new imaging techniques such as magnetic phase contrast imaging using polarized neutrons’.
Dhruva is also being used to develop new fuel types via its in-pile loop facility which has higher heat removal capacity than
was available in the now decommissioned Cirus. This can accommodate fuel arrays from India’s emerging medium-sized PHWR fleet as well as that of the Advanced Heavy Water Reactor (AHWR) that uses thorium for testing purposes. In the recent past, a twisted pin geometry cluster, with dispersion-type fuel was successfully irradiated up to its research target burn-up to study of fuel performance, such as the reaction between dispersant and diluents, swelling characteristics, clad deformation behaviour, etc.
Dhruva’s existing pneumatic carrier facility, which can provide a thermal neutron flux of 9.7x1013 n/cm2/s for the irradiation of samples that yield short-lived isotopes on irradiation, continues to be in demand. Dhruva also has a prompt gamma ray neutron activation analysis system that is used for the online analysis of various materials. Dhruva’s small angle neutron scattering facilities have also been used for studies on soft condensed matter and large inhomogeneities in the past.
Supporting the FBR
While not a research reactor in the strictest sense, the Fast Breeder Test Reactor (FBTR) at the Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, which attained first criticality in 1985, has been the mothership for FBR-related R&D in India. Identical to the French Rapsodie- Fortissimo except for the addition of steam generators and a turbo-generator (the unit was connected to the grid in 1997), this 40MWt (13.5MWe) sodium-cooled loop- type reactor has been used to train a cadre of personnel. It also provided experience in fast reactor operation and large-scale sodium handling, while serving as a means to test fuels, structural materials and special neutron detectors.
Given a 20-year life-extension in 2011, FBTR is currently into its 25th irradiation campaign, which has seen the power raised to 27.3MWt for the first time in its operating history. This campaign, like the ones immediately preceding it, involved the loading of yttria-72 thoria subassemblies for irradiation in order to produce uranium-233. This campaign involves irradiation of yttria, natural U-Zr sodium-bonded metal fuel pins, uranium (14.3% enriched) metal pins, and ternary fuel pin U-Pu-Zr, as well as impact specimens of low nitrogen austenitic stainless steels that make up the grid-plate of FBTR itself.
These sub-assemblies are in addition to the fuel sub-assemblies that are at any time a part of the FBTR core. They include indigenously developed Mark-I (70% PuC + 30%UC) and Mark-II (55% PuC + 45%UC) mixed carbine fuels; and MOX (44% PuO2 + 56% UO2). More than 1200 Mark-I fuel pins have reached a burnup of 155GWd/t in FBTR over the years, and the record burnup is 165GWd/t – the highest ever attained internationally. Only one Mk-I fuel pin failure event has been observed up to now.
FBTR’s fuel-cycle has been ‘closed’; new mixed carbide fuel was fabricated after fuel, discharged at up to 150GWd/t, was reprocessed in IGCAR’s CORAL facility. PFBR fuel under irradiation testing in FBTR has reached a burnup of 112GWd/t and has been discharged for post-irradiation examination. About a fifth of FBTR’s core is currently loaded with high-Pu MOX as part of efforts to develop higher-burnup fuel for PHWRs.
Past irradiation campaigns have included studies on D9 structural materials, testing of high-temperature fission chambers for PFBR up to 10MWt, and testing of an industrial version of a Kalman-filter-based instrument, meant for drop time measurement of the Diverse Safety Rod Drive Mechanisms (DSRDM) for PFBR.
FBTR operations have not been without incident, however. A major fuel handling incident has been recorded, as has a primary sodium leak, as well as reactivity transients. As a result, major structural modifications have been carried out in the past, including one on FBTR’s steam generator leak detection system and the other on its steam-water circuit. These changes have helped boost the reactor’s availability.
Refurbishment was carried out to support life extension. That included replacing steam generator rupture discs and main boiler feed pumps; commissioning a new 1MVA transformer along with associated switch gear to augment the existing power supply; and a new demineralised water plant. FBTR is also receiving post-Fukushima retrofits. Last year IGCAR installed a new flood-safe building for emergency diesels, and it installed and commissioned two mobile diesel generator sets for accident scenarios.
The 30kWt Kalapakkam Mini (Kamini) reactor, commissioned at IGCAR in 1996, has been used to test high-temperature fission chambers developed for PFBR up to a neutron flux of 5x109 n/cm2/s and gamma field of 5 x 105 R/hr at a temperature of 570°C, in a test assembly made up of Inconel tube which is installed on the east side of the reactor in the pool. Kamini is also used for neutron radiography of irradiated FBTR fuel.
This tank-in-pool type design is a reflector-moderated reactor where half of fission events are due to reflector-returned neutrons. Zircoloy-2 canned BeO is used as a reflector: it has high reflection efficiency, which results in lowering fuel inventory requirements. Kamini is a neutron source with a flux level of 8.0x1012 n/cm2/s at the centre of its core. It has facilities for carrying out neutron radiography of radioactive and non-radioactive objects, as well as the neutron activation analysis of a variety of samples.
Essentially a materials testing reactor, Kamini has been used by ISRO for the non- destructive examination of thousands of pyro devices used in India’s space programme.
It has also been used to irradiate space- borne sensors to test their reliability in the face of cosmic radiation. In recent years, due to the non-availability of Apsara during refurbishment, Kamini has also been used to test and calibrate neutron detectors. Owing to its design, Kamini has been unable to host major shielding experiments.
Kamini is also a test-bed for the third stage of India’s three-stage nuclear programme, given that it is the only uranium-233
fuelled reactor in use today. In the past, thoria fuel rods irradiated in Cirus have been reprocessed in the Uranium-Thorium Separation Facility (UTSF) at BARC and the recovered U-233 has been fabricated as fuel for the Kamini reactor at IGCAR. U-233 from thoria fuel bundles irradiated in FBTR by CORAL have also been used to fabricate fuel for Kamini. At the moment, Kamini is licensed for operation till June 2020, after a periodic safety review.
Critical Facility
Back in Trombay, third-stage activities are being supported by the 100W critical facility, commissioned in 2008, which can generate a thermal neutron flux of 108 n/cm2/s and has been designed to facilitate study of different core lattices based on various fuel types, moderator materials and reactivity control devices. It is being used to validate the reactor core physics for AHWR, which will utilise thorium fuel. AHWR studies began with a 54 pin (Th-Pu)MOX/(Th-233U) MOX cluster representative AHWR core. The critical facility is also being used for the testing and qualification of reactivity control devices and detectors, and in neutron activation analysis of various samples.
A new generation
India’s next generation research reactors will be constructed at BARC’s new campus in Vishakhapatnam, on the East Coast. Among the new reactors planned, the High Flux Research Reactor (HFRR) is currently in the most advanced stage of implementation. Preparation and review of preliminary safety analysis report, part-A and preliminary architectural drawings of the structures have been completed.
The 30MWt HTRR is a swimming pool type, thermal reactor design that will use U3Si2 dispersed fuel with enrichment levels of 19.75%. The reactor will be cooled and moderated by light water, with heavy water as a reflector. The maximum thermal neutron flux will be 6.7x1014 n/cm2/s and the maximum fast neutron flux will be 1.8×1014 n/cm2/s to cater to the requirements of radioisotope production, fuel and material testing as well as advanced neutron beam tube research. Laboratory-scale recovery of uranium from the silicide fuel has been established using a feed composition of reference burnup for 80GWd/t, with ten-year cooling.
Joining HFRR in the years ahead at Vishakapatnam will be the 125MWt Thermal Research Reactor (THRR), which will be
a vertical tank type reactor, rather similar in design to Dhruva. THRR will be fuelled by either natural uranium metal or slightly enriched uranium metal (1.25%) and will be able to generate a maximum thermal neutron flux of 2.2x1014 n/cm2/s. Its focus will be on irradiating various materials, although it will also have facilities for neutron beam tube research, radioisotope production, neutron activation analysis and neutron radiography.
Meanwhile, IGCAR has prepared a preliminary layout for a follow-on to the FBTR, dubbed FBTR-2, in line with its experience with PFBR. This is expected to be a 300MWt (150MWe) fast breeder reactor that will be a test-bed for metallic fuels that will power the future 1000MWe Metallic FBRs (MFBRs), which are currently at the design stage. FBTR-2 will use U-Pu alloy or U-Pu-Zr, with electrometallurgical reprocessing to close the fuel cycle.
India has already built a number of research reactors to support its three-stage nuclear programme. It is clear that this fleet will play a quiet but pivotal role in aiding India’s new nuclear programme.
