The ultimate in safety1 January 2003
The design of a 1000MWe heavy water gas-cooled reactor under development in Russia has safety as its driving force. By B I Iljitchev, G V Kiselev, B P Kochurov, M L Ochlopkov, V V Seliverstov and O V Shvedov
Large-scale expansion of nuclear power must be based on reactors that meet two main requirements: "ultimate" safety and highly efficient fuel utilisation. The heavy water gas-cooled reactor concept developed over recent decades in the former USSR fulfils these goals.
"Ultimate" safety may have a variety of safety definitions, often inconsistent. For this project, it means eliminating the possibility of radioactivity releases from the plant that have significant environmental impact, allowing in principle for the possibility of core damage and even core complete melt. Reducing this probability to an acceptable level is also of importance from the viewpoints of financial risk and public acceptance. To achieve this it is necessary to consider the safety impact and the likely challenges to it.
The internal challenges to the envelope are:
• Sudden pressure buildup, due either to a power excursion caused by prompt neutrons or a sudden release of internal energy stored in the primary coolant.
• Gradual pressure increase to a level exceeding design limits.
• Explosive chemical reactions in an emergency.
• Ablation by molten core debris.
To avoid these challenges, the reactor design must preclude:
• Power excursions caused by prompt neutrons.
• The possibility that core debris could be reconfigured to allow criticality.
• Positive power reactivity coefficient at any power level.
• Brittle failure of the pressure vessel.
• Primary system boundary ruptures able to cause sudden pressure buildup.
• Explosive chemical reactions in the case of core damage.
• Molten fuel release to the environment due to ablation.
HWR-1000US reactor design
The inherently safe heavy-water-moderated gas-cooled reactor (HWR-1000US) that fulfils the above requirements has been under development for more than 30 years and has reached the stage of extended conceptual design. The design has been led by ITEP, with the participation of some other former USSR institutions.
Economics and feasibility studies have been performed by VNIPIET and have shown that the reactor design is quite feasible on the basis of existing technology. VNIPIET also concluded that the reactor would have reasonable economic characteristics and its construction and operating costs would not exceed - and may even be lower than - those of the other reactor designs including current LWRs.
The reactor concept has several key features. The entire primary system, including the main gas circulators, steam generators and intermediate heat exchangers, are arranged within a multi-cavity prestressed concrete vessel (PCV) that retains primary coolant pressure. It uses a low temperature heavy water moderator and gaseous coolant, and the fuel is low-enriched.
Coolant pressure is retained within the PCV, in contrast to Atomic Energy of Canada's Candu reactor where coolant is retained by pressure channels. From a safety standpoint, the large-break LOCA is eliminated, as a basic feature of the PCV is the impossibility of brittle failure. In the PCV construction, breaches are restricted to openings of less than 30cm diameter, which precludes sudden release of stored internal energy from the primary system.
The integral arrangement whereby coolant pressure is retained by the PCV has advantages. Since the fuel channels are not bearing the coolant pressure the channel tubes can be thinned considerably. That in turn significantly reduces parasitic neutron capture and increases fuel utilisation.
Using gaseous primary coolant instead of heavy water also has advantages.
Genesis of the reactor design
The HWR-1000US is based on operating experience with the KS-150 reactor Bohunice, in what is now Slovakia (known as A1, this unit operated successfully from 1972 to 1977).
The reactor and the main components of the gas and heavy water loops are located within a PCV, along with a stainless steel liner, core debris catcher and cooling systems, and a system to collect and blow back any gaseous coolant leaks through the PCV.
The PCV and the auxiliary equipment for the gas and heavy water loops are housed within a leaktight steel shell (the primary envelope) with a design overpressure of ~3.5bar. The primary envelope and the reactor rooms are surrounded by yet another steel shell (the secondary envelope) with a design overpressure of ~0.4bar.
The stainless steel reactor vessel, located in the central PCV cavity, houses 361 vertical calandria tubes arranged in a triangular lattice with a pitch of 41cm. Of these, 342 fuel channels each contain 126 metallic fuel rods 0.6cm in diameter, coated in a zirconium alloy and cooled with either carbon dioxide or helium. The remaining 19 calandria tubes are reserved for control devices. The latter have a tube within a tube configuration, with the inner low absorbing tube for regulating and normal load following, and the outer high absorbing tube for reactor shutdown and compensation.
The main core characteristics are as follows: core diameter 8.20m; core height 5.00m; radial reflector thickness 0.60m; that of axial upper and lower reflectors 0.40m; core thermal power 3200MW, including power in fuel 3000MW and in moderator 200MW.
The six steam generators together with gas circulators and the two heavy water moderator intermediate heat exchangers are located in the eight vessel cavities provided on the PCV periphery.
This design concept allows all of our safety goals to be fulfilled.
For all practical purposes, gaseous coolant neither absorbs nor slows down neutrons. Coolant presence or absence in the core therefore has virtually no influence on the system criticality. This means that a loss of coolant accident (LOCA) does not constitute a reactivity accident. This compares to the Candu design, where the relatively large positive coolant void coefficient requires fast-acting active shutdown systems.
Ingress of heavy or light water in the gaseous coolant leads to a reactivity reduction, as neutrons are slowed. This decreases the fast-fission multiplication factor and increases resonance capturing due to the resonance shielding effect reduction.
Accidental withdrawal of all control rods placed in the core during operations adds a relatively small amount of positive reactivity to the system, easily compensated by the negative reactor power coefficient. This is made possible because of the combination of almost no coolant or moderator effects and small fuel effects, as well as by reactivity compensation (burnup and xenon) via an adjustment of boric acid concentration in the moderator. The outcome is a small reactivity holdup in the control rods no greater than about 0.5.
Adopting optimal core geometry with respect to reactivity provides reactivity reduction for channel or core geometry changes.
The danger of core debris reconfiguration to criticality is eliminated by using a fuel with very small fissile content, which is made possible by the low-absorbing heavy water moderator.
Since coolant and moderator reactivity effects are zero for practical purposes, the reactivity effects depend only on the fuel and are consequently negative in all operating regimes.
The properties of the pressure vessel preclude brittle failure. If the pressure buildup rate is slow the excess pressure will be released through cracks in the PCV. The vessel would be destroyed only by a thermal explosion - impossible thanks to the reactor's inherent properties. The non-explosive nature of the chemical processes in an emergency is achieved by the proper selection of reactor materials, which eliminates hydrogen explosion.
The PCV construction characteristics limit the maximum possible vessel breach equivalent diameter to about 30cm.
If the coolant were pressurised water it would undergo phase transformation (water-steam), resulting in a significant rise in containment pressure during a LOCA. The pressure increase must be limited by dousing or by vapour suppression systems. But using gaseous coolant, a small-break LOCA does not lead to a sudden pressure increase and these systems are not required. And since the large-break LOCA is eliminated by using the PCV, in an emergency all the primary coolant can be reliably confined within a relatively small volume envelope whose integrity is not endangered by the relatively low pressure. This is the main safety function of the first intermediate envelope (containing only the PCV). The second envelope, of much larger volume and also capable of reliably confining all primary coolant, extends outward to the reactor containment, forming with the latter an isolation space normally kept under a negative pressure.
Use of a PCV whose integrity is retained under all credible accidents allows systems to be included within the PCV to catch and retain molten fuel, as well as reliable systems for cooling the caught fuel and the pressure vessel itself. This guarantees against the fuel being released to the environment and decreases the extent of gaseous releases during an emergency - including hydrogen formation. With no interaction between the molten debris and the concrete we can eliminate the formation of explosive hydrogen concentrations as well.
In such a way, the envelope pressure would increase slowly, due mainly to fuel decay heat, which is removed by dedicated passive systems within the PCV.
Fuel cycle characteristics
Gaseous coolant, together with a large (10cm) fuel-channel radius, results in a high probability that a fast-fission neutron will cause a new fission in the fuel channel before escaping or being or slowed. This provides a well-moderated neutron spectrum, which is beneficial for effective fissile isotope utilisation, and a high fraction of fissions in fertile isotopes (around 10%), increasing neutron balance and hence fuel utilisation.
To illustrate the HWR-1000US' unique fuel cycle capabilities the main characteristics of the once-through natural uranium cycle are shown in Table 1.
In the closed U-Pu equilibrium cycle we can eliminate the need for mined uranium by using depleted uranium (~2.0kg/tU-235) taken from enrichment plant tailings. Plutonium isotopes, generated in uranium during operation, are mixed with uranium tailings and an appropriate amount of external plutonium (either weapons or commercial). This process is repeated until the feed plutonium isotopes come into an equilibrium composition. The distinct advantage of such a cycle is that it reduces the enormous inventory of enrichment plant tails, and destroys excess plutonium.
An alternative cycle could consist of depleted uranium (~3.5kg/tU-235), and plutonium produced in earlier cycles. When an equilibrium is established the equilibrium fissile plutonium concentration will be ~3.7kg/t. Such a cycle neither produces nor consumes plutonium; the main cycle characteristics are shown in Table 2.
The low U-235 content in the fuel allows us to use a uranium fuel that mixes natural uranium and enrichment plant tailings. With the U-235 content in tails of ~2.0kg/t, the uranium in the fuel is a mixture of ~30% natural and ~70% tailings. For an installed capacity of 1GWe the natural uranium consumption will be 30t/year, along with 70t/year of tailings.
If we use plutonium and tailings, the reactor will destroy up to 250kg/year of plutonium.
Recycling of produced minor actinides is also possible.
In this case the reactor will not produce excess plutonium or minor actinides. The effective burning of minor actinides is due to the relatively high thermal flux ~(0.5-0.7)x1014n/cm2s. Increasing the U-235 content to 3.8-4.0kg/t compensates for the extra absorption.
Tables Table 1: HWR-1000US key physical parameters in the once-through natural uranium cycle Table 2: Main U-Pu equilibrium cycle characteristics