Improving resilience of India’s fleet

POST-FUKUSHIMA SAFETY REVIEWS CONDUCTED in India revealed that the country’s existing fleet of power generating reactors had sufficient ‘defence in depth’ (DiD) to withstand design basis events such as earthquakes and floods, even those that would result in extended station blackout (SBO) and loss of the heat sink.

New safety measures were agreed subsequent to these reviews to enhance the resilience of India’s plants to beyond design basis events, as well as to bolster severe accident mitigation capabilities. Deploying these measures across the fleet is in keeping with India’s ethos of embracing nuclear safety and regulation as continuously evolving, informed by national and international experience and strengthened through a large, active domestic research & development (R&D) programme on prevention, monitoring and mitigation.

India’s willingness and ability to incorporate worldwide operational experience is evidenced by the fact that indigenous designs currently under construction have anticipated the move towards inherent and passive engineered features to maintain fuel cooling and restrict radioactive release in the event of a severe accident. It continues to invest in dedicated safety development facilities, with modelling and simulation supplementing experimental analysis. The Nuclear Power Corporation of India (NPCIL) claims over 500 reactor years of safe operation. NPCIL stands proudly amongst its peers in the World Association of Nuclear Operators (WANO).

Keeping NPCIL and its parent the Department of Atomic Energy (DAE) on their toes is the Atomic Energy Regulatory Board (AERB), which is independent of DAE and is responsible for monitoring and enforcing safety in Indian reactors. AERB discharges these functions via an multi-tier review and authorisation process for nuclear, industrial and radiation safety.

Indian nuclear plants are typically licensed for a maximum period of five years by AERB with renewal subject to a comprehensive safety review (CSR) that takes place every five years and a periodic safety review (PSR) that is conducted every ten years. PSRs are designed to address ageing management concerns and evaluate the state of the plant with respect to current safety requirements and practices.

Since the second extraordinary meeting of the Convention of Nuclear Safety in 2012, AERB has been updating its regulatory requirements on reactor siting and design to fully incorporate the lessons learnt from Fukushima 2011. These requirements are in accord with the latest International Atomic Energy Agency (IAEA) standards and other international benchmarks. India is committed to the Vienna Declaration on Nuclear Safety and the IAEA’s Action Plan endorsed by member States in September 2011.

In addition to CSRs and PSRs, AERB’s regulatory framework also includes a system of special safety reviews (SSRs) undertaken after any major nuclear event. It is the SSR conducted by AERB after Fukushima, alongside NPCIL’s own in-house reviews, which formed the basis for the three distinct sets of short, medium and long-term safety measures that have been applied to Indian reactors since late 2011.

It is worth noting that the AERB and NPCIL reviews reconfirmed ‘the inherent strengths in design, practices and safety regulation followed in India’. In general, DAE believes that the pressurised heavy water reactors (IPHWRs), which account for 18 of India’s 22 power generating reactors, have advantages in being able to slow down accident progression. This is for two reasons. The steam generators are located above the core in IPHWRs and this enables the removal of decay heat through thermosiphoning of the primary system and the boiling water in the steam generators. In IPHWRs, water can be supplied to depressurised steam generators via diesel pumps even in SBO conditions. Moreover, steam discharge valves can depressurise the steam generators even during station blackout and with compressed air unavailable. Second, the core is always enveloped by water at low temperature and pressure water in the calandria and the calandria vault, which serves as a heat sink that can delay the progression of a severe accident.

The post-Fukushima reviews resulted in short, medium and long-term measures to ensure that certain minimum safety functions remain available even in the aftermath of extreme beyond design basis events, and to improve severe accident mitigation capabilities in general. The short-term measures were limited in scope: providing external hook-up points for adding water to reactor systems and the spent fuel bay; deploying emergency lighting backed up by solar cells; reviewing and revising emergency operating procedures and operator training.

Medium-term measures included: introducing automatic seismic trip functions in older reactors, providing additional backup diesel-generators at a higher elevation; strengthening the ability to monitor critical parameters under prolonged loss of station power using battery powered devices; deploying diesel-driven pumps to transfer water from the deaerator storage tank to the steam generators; additional mobile pumps and fire tenders; seismic strengthening; and increasing onsite water storage to cover decay heat removal for 30 days. NPCIL finished implementing the short and medium-term safety measures on a fleet-wide basis by 2015/16, so most upgrades have been in place for a while now.

Site-specific safety upgrades

Alongside these measures, additional plant-specific upgrades were also undertaken. For instance, flood defences were strengthen at the Madras IPHWR site to protect the plant against future tsunamis higher than that of 2004. The Tarapur 1&2 BWRs, which are older than Fukushima Daichi 1 and had already been upgraded to ensure continuous cooling during site blackouts, were modified to allow nitrogen injection into the containment in the event of a hydrogen build-up.

The long-term safety measures relate to severe accident management. They include: creating station-specific accident management guidelines approved by AERB; strengthening hydrogen management provisions; providing filtered venting of the containment; and setting up on-site emergency support centres capable of withstanding natural calamities such as earthquakes, cyclones and floods.

As of today, station-specific accident management guidelines are in place across the board at NPCIL and personnel are trained. Meanwhile, indigenously developed passive catalytic recombiner devices (PCRD) are now being installed alongside facilities for homogenising the containment atmosphere. These installations were greenlighted once the PCRD housing and the catalyst- bearing panels (CBPs) received seismic qualification on the basis of shake-table tests performed a couple of years ago. The PCRDs underwent development testing at a hydrogen recombiner test facility (HRTF), which has an instrumented 60m3 vessel in which high concentrations of hydrogen, steam or air can be contained safely. A new cordierite-based PCRD is under development.

An indigenously developed containment filtered venting system (CFVS), which operates on the wet-scrubbing principle, is also being deployed fleet-wide after it was granted approval by AERB. A full-scale system has been operational at Tarapur 3 (540MWe) for some time. NPCIL is also deploying iodine scrubbing through a containment spray system on which tests using different aerosols have been conducted and removal rate measurements recorded.

All the safety measures adopted after Fukushima are now part of the standard design of IPHWRs and will therefore feature on the six 700MWe IPHWRs (IPHWR-700s) now under construction, as well as the ten more IPHWR-700s to be built in ‘fleet mode’ over the next decade. In fact, once ready, the iodine scrubbing CSS mentioned above will be added to the IPHWR-700s under construction.

Tthe IPHWR-700 design is a step-up in safety over older IPHWRs. While the greater power output has been achieved by allowing partial boiling at the coolant channel outlet, the IPHWR-700 features the interleaving of primary heat transport system feeders, regional over-power protection, a containment spray system, a mobile fuel transfer machine, a steel liner on the inner containment wall and a passive decay heat removal system. The IPHWR-700’s ECCS consists of passive high-pressure injection, followed by an active long-term recirculation phase for removal of decay heat, using ‘all-headers’ injection for this purpose. The injection is initiated automatically upon sensing a fall in reactor inlet header pressure below 40kg/cm2 (g) along with the presence of high calandria level or high pump room pressure.

Bolstering safety at India’s PFBR

The sodium-cooled prototype fast breeder reactor (PFBR) in Kalpakkam, which is currently at its early commissioning stage, has a passive decay heat removal capability, by way of natural circulation through dedicated heat exchangers. The PFBR features the ‘warm roof’ concept in order to minimise the risk of sodium aerosol deposits, has two- independent fast-acting shutdown systems, applies ‘leak before break’ for the main vessel, sodium piping and steam generators; and a robotic device for in-service inspection of the main vessel.

Fuel failures will be detected through continuous monitoring of the cover gas fission product activity and delayed neutron detection in the primary coolant. Being built by Bhavini, the PFBR has also cleared post-Fukushima safety reviews. However, the PFBR’s designers (Indira Gandhi Centre for Atomic Research, a part of DAE) is currently developing and evolution of the design in the form of the 500MWe Commercial Fast Breeder Reactor (CFBR), in which the sodium void reactivity value will be kept lower than 1 $, or near zero, as compared to 2.4 $ for the PFBR.

Safety enhancements in the CFBR include in-vessel sodium purification, additional passive features in existing shutdown systems such as a stroke-limiting device, a third shutdown system which would be liquid poison-based and improved decay heat removal capability with greater capacity for natural circulation.

Looking ahead to AHWR

Arguably the safest design being developed in India is the thorium-based 300MWe Advanced Heavy Water Reactor (AHWR).

The AHWR uses natural circulation to cool the reactor core at all times. Eliminating major components such as primary coolant pumps and drive motors, and their control and power supply equipment, should make the AHWR more reliable and safer than IPHWRs.

The AHWR has a slightly negative void coefficient of reactivity, passive safety systems, a 8000m3 heat sink in the form of a gravity-driven water pool near the top of the reactor building, direct injection of cooling water by the ECCS inside the fuel cluster, two independent shutdown systems and passive poison injection into the moderator in the event that both shutdown systems are unavailable.

While a reduction in the material inventory of nuclear reactors is certainly welcome, the fact remains that a reactor design will ultimately be only as good as the quality of components that are used to build it. In this context, studies are underway at DAE to determine realistic failure modes under the large magnitude reversing cyclic loads that may be experienced during a severe beyond design basis event such as a major earthquake.

One set of tests and analyses focused on several dozen pipes and elbows. The result is a set of rational design criterion and simplified procedures for integrity assessment against cyclic-tearing and ratcheting-fatigue. These are now being used to refine current design codes for nuclear components. The ultimate goal is to develop additional design provisions that will ensure that basic safety functions are not impaired even, during the most severe earthquakes.

Author information: Saurav Jha, Author and commentator on energy and security, based in New Delhi