Pressurised Heavy Water Reactor (PHWR)

The Pressurised Heavy Water Reactor (PHWR) is a thermal-neutron nuclear reactor that employs heavy water (deuterium oxide, D₂O) as both neutron moderator and primary coolant while burning natural uranium—containing only the ~0.7 percent fissile U-235 found in ore—as fuel. The design lineage descends from the Canadian CANDU (CANada Deuterium Uranium) reactor developed by Atomic Energy of Canada Limited from the late 1950s, the first of which, the Nuclear Power Demonstration unit at Rolton, Ontario, achieved criticality in 1962. India's programme began under a 1963 agreement with Canada that produced the Rajasthan Atomic Power Station Unit 1 (RAPS-1), modelled on the Douglas Point reactor. After Canada terminated cooperation following India's 1974 Pokhran-I test, the Department of Atomic Energy and the Nuclear Power Corporation of India Limited (NPCIL) indigenised and standardised the design, making the PHWR the technological workhorse of Stage 1 of India's three-stage nuclear programme conceived by Homi Bhabha.

The reactor's defining feature is its decoupling of moderator and coolant into separate systems, made possible by the pressure-tube design. Rather than a single large pressure vessel as in a light-water reactor, the PHWR uses a horizontal cylindrical tank called the calandria, which holds the cool, low-pressure heavy-water moderator at near-atmospheric pressure. Penetrating the calandria are hundreds of horizontal pressure tubes, each containing fuel bundles and carrying the separate stream of hot, pressurised heavy-water coolant. Coolant enters at roughly 260 °C and exits near 290–300 °C at pressures around 90–100 bar, then passes through steam generators where it transfers heat to a secondary light-water circuit, raising steam that drives the turbine-generator. Because the two heavy-water circuits are independent, the moderator stays cool and unpressurised.

Two mechanical consequences flow from this architecture. First, the pressure-tube design permits on-power refuelling: fuelling machines latch onto opposite ends of a channel and push fresh fuel bundles in while pushing spent bundles out, so the reactor need not be shut down to refuel. This sustains high capacity factors and, historically, raised proliferation concerns because it eases extraction of plutonium-bearing fuel at optimal burn-up. Second, natural uranium fuel obviates enrichment entirely—an advantage for states lacking enrichment capacity—but heavy water's superior neutron economy is essential, since natural uranium cannot sustain a chain reaction in an ordinary-water moderator that absorbs too many neutrons. The trade-off is a large, costly inventory of heavy water and lower fuel burn-up, generating more spent fuel rich in plutonium-239, which Stage 2 fast-breeder reactors are designed to consume.

India operates the world's largest fleet of PHWRs. NPCIL standardised on 220 MWe units and later scaled to 540 MWe at Tarapur (TAPS-3 and TAPS-4, commissioned 2005–2006) and to the 700 MWe Indian PHWR (IPHWR-700). The first 700 MWe unit, Kakrapar Atomic Power Station Unit 3 in Gujarat, attained criticality in July 2020 and commercial operation in 2023, followed by KAPS-4 and the Rajasthan units RAPS-7 and RAPS-8. In 2017 the Union Cabinet sanctioned a fleet-mode programme of ten 700 MWe PHWRs. Comparable reactors operate abroad as CANDU units in Canada (Bruce, Darlington, Pickering), Argentina (Embalse), Romania (Cernavodă), South Korea (Wolsong) and Pakistan (KANUPP at Karachi).

The PHWR must be distinguished from the Pressurised Water Reactor (PWR), the most common reactor type worldwide, which uses ordinary light water as both moderator and coolant and therefore requires enriched uranium (typically 3–5 percent U-235) inside a single steel pressure vessel with no on-power refuelling. It also differs from the Boiling Water Reactor, where coolant boils directly in the core, and from gas-cooled and light-water-graphite (RBMK) designs. Within the PHWR family, the Indian IPHWR diverges from the original CANDU in containment design, fuelling-machine configuration and the adoption of fully indigenous standards after international cooperation lapsed. The Kudankulam plant in Tamil Nadu, by contrast, uses Russian VVER pressurised light-water reactors and is not a PHWR.

Controversies attach principally to proliferation and safety economics. The unsafeguarded CIRUS research reactor—a heavy-water design—supplied plutonium for India's 1974 test, cementing the association between heavy-water technology and weapons capability and prompting creation of the Nuclear Suppliers Group. Heavy-water reactors also produce tritium, a radioactive hydrogen isotope and a weapons material, raising occupational and environmental management burdens. The 2008 India–US civil nuclear agreement and the subsequent NSG waiver placed designated Indian PHWRs under International Atomic Energy Agency safeguards while leaving the strategic fleet outside, a separation plan central to India's distinctive non-NPT status. Recent debate concerns whether to continue investing in Stage 1 PHWRs or accelerate toward Stage 2 fast breeders, given the protracted commissioning of the Prototype Fast Breeder Reactor at Kalpakkam.

For the working practitioner—particularly the civil-services aspirant or energy-policy analyst—the PHWR is the concrete embodiment of India's strategy of nuclear self-reliance keyed to its limited uranium but vast thorium reserves. Mastery of the design clarifies why Stage 1 exists: PHWRs burn natural uranium and breed the plutonium that fuels Stage 2 breeders, which in turn produce the U-233 from thorium that powers Stage 3. The reactor type explains India's heavy-water plants, its safeguards diplomacy, its export ambitions, and its careful sequencing of an indigenous fuel cycle. Understanding the PHWR is thus prerequisite to analysing India's energy security, climate commitments and strategic autonomy in the nuclear domain.