From G-I to G-IV:

RUSSIAN CLASSIFICATION OF NPP GENERATIONS. BRIEF HISTORICAL OVERVIEW

Like in the cases, which being described in the Introduction (see above), classification of NPP, there are large numbers of various classifications ships and vessels[i] powered with Nuclear Power Propulsions exists. For the purpose of consideration of exactly the propulsion complex as a unified system, it would be convenient to use classification according to generations of the reactors, regardless of other peculiarities of the equipment, the intended use and other characteristics of ships. Accepted in the USSR and widely used in Russia, the classification of NPS according to generations of reactors is also convenient in that the transition to a new generation of reactors was simultaneously the transition to new generation of ships, and was accompanied, without an exaggeration, by a “revolutionary breakthrough” in technologies of designing of new engineering, units and equipment, in construction of ships, changing of navigation and weapon.

There are not much principle differences between a naval NPP, floating NEPP, and land based – stationary NEPP created on base of an LWR. All these types of PP operate according to similar thermodynamic cycle and use identical principles of energy transformation and transfers. With relatively smaller dimensions of the core and NR, higher specific parameters are obtained in naval reactors:

• power capacity obtained from a unit of volume of the reactor core
• power capacity taken from a unit of the surface of the reactor core and respectively, from single fuel element
• power capacity obtained from a unit of weight of the fuel
• power capacity related to weight-dimension parameters of the power plant, etc.

The enrichment of nuclear fuel of the core of civilian NEPP with U²³⁵ does not usually exceed 4-6%%. The level of enrichment of U²³⁵ in the core of the various up-to-date naval reactors reaches 90%. In the core for naval LWR (Soviet and Russian manufactured), the enrichment does not usually exceed 40-50% or less, but in US naval reactors utilize fuel enrichment exceed 90% and higher. A high level of enrichment of fuel of the naval core allows to prolong life-time of the reactor between refueling operation, which is referred to as “campaign”, and to make refueling of the core for a number of times less frequently, than it is on reactors of NEPP. However, if high-enriched fuel use, there is another side of the problem, early breach of integrity of fuel elements by using more energy-intensive core associated with swelling of the fuel composition. Often engineering and design work is the search for balance between conflicting components constituting technical characteristics of the system.

The thermal power capacity of naval reactors varies from 2-5 MWt for small auxiliary NPP, up to conditionally standard – average of 190-200 MWt for the reactors of submarines, and 300-350 MWt for the propulsion plants of surface ships and non-military ice-breakers.

We shall consider the basic characteristics of the construction of SGP of various generations designed in the USSR, there principle diagrams[ii] and we shall try to evaluate their advantages and drawbacks, disadvantages. The description of failures and nuclear incidents, and the analysis of the reasons will not be carried out in the present chapter, and this issue will addressed in the appropriate description below.

1.     Naval NPP of the first generation

In 1952, in the USSR, the works began, for the creation of the first nuclear-powered submarine (NPS). For the purpose of solving the formulated problem, it was necessary to solve a huge amount of engineering-design, technological, materials science, organizational-administrative and other problems.

First of all it was necessary to create a unified reactor – the basic component of a naval power propulsion plant. But in addition, for the purpose of provisioning of the operation of reactor, the presence of special systems, mechanisms, various equipment for ensuring operation of the reactor, and the power propulsion plant itself is necessary.

In 1955, in Scientific-research Institute “Gidropress”, the works began, for the creation of a so-called “naval” reactor which subsequently received an index of BM-1. No analogues of such a structure existed in the USSR, therefore, a transposition of the principles of arrangement and design solutions, from uranium-graphite research reactors onto a LWR could not at once result in a success. Thus, for example, in the technical specifications (TS) for designing, creation of a reactor with a horizontal arrangement of channels in the core was proposed; however, due to the decision of academician N.I. Dollezhal, the reactor was rotated to have a vertical position.

During the development an intermediate variant of reactor design was created, which is still working today, this is the reactor of Obninsk NEPP, within the territory of FEI named after academician Leipkunsky. For the first time during creation of new type of reactors the following problems were solved:

• Analysis and optimization of the thermal circuit of NR and searching of basic parameters
• Operating and control system of neutron-physical processes in the core
• Methods of neutron-physical calculation of the cores
• Problems of burning-out of fuel and accumulation of fission fragments of U²³⁵
• Development  thermal-hydraulically model for reactor circuit calculation
• Development of the operating and control system for the reactor and the entire propulsion plant.

As the result of the accomplished of works, a so-called “looped scheme” of reactor of the first generation was created (see Fig. 1). The SGP was remarkable for its fairly good weight-dimension parameters, high parameter of specific power capacity, good operational maneuverability. There was no technology of cladding of the internal surface of the vessel with austenitic steels, and on the first reactor of BM-1 (“military naval first”) type a double vessel was used, consisting of external – strong, and internal corrosion resistant vessel-sheath. The internal sheath was strained by a special screw in the vessel, and on the sealing plane it was squeezed between the cover and the reactor vessel. During operation, the sheath was inflated by the internal pressure and it densely adjoined to the strong pressurized vessel of the reactor, thus protecting it against corrosion. Later, the internal vessel-sheath was replaced with fused plated corrosion-resistant surface. This technology is used till today.

In addition, in several variants, gaskets – press rings between the vessel and the cover of the reactor were tested. Basically, it was concerned with searching of new materials. Thus, on one of the reactors a red-copper (annealed) press ring was replaced with a nickel one. During refueling operations it appeared, that affinity between nickel and steel, and the diffusion of materials into each other were so significant, that during disruption of the cover, the sealing surfaces in the location of disconnection were strongly damaged. Application of nickel gaskets was hereinafter rejected.

After the whole lot of improvements, and the transition from ring fuel elements to pin-shaped ones, ensuring relatively higher reliability of operation of the plant, the reactors of such type were in the service on board the submarines of the USSR Navy till the end of 80′s.

The basic drawbacks of SGP of the first generation, with a loop-back circuit were repeatedly analyzed in the various sources:

• Large extension of pipelines and branching of RCS systems, and respectively, large volume of coolant in the primary circuit
• The presence of pressurized pipelines of a large diameter, connecting the main equipment, and accordingly, the presence of a plenty of welded connections
• The two reasons mentioned above, as a consequence, have generated the following one – a low reliability of the equipment and non-optimal weight-dimension characteristics
• Minimal automation of the process of NPPP control, low reliability of control and protection systems of the reactor and uncertainty of the indications of control-and-measuring instruments. An unreliable monitoring system for the process of fission within the core.

It would be possible to list yet another whole range of drawbacks of NPPP of the first generation; however, one should remember that it was the first experience of creation of fundamentally new engineering systems. Many existing branches of industry and science turned out not to be ready to solve the problems arisen before them. It was impossible to use any experience for solving these technical problems. As a matter of fact, there was a new branch created in Soviet science and industry. The similar works were conducted in several countries in parallel, but in all these countries the research works were strictly classified, and naturally, any international cooperation during that time was out of the question.

For today, all NPS with NPPP of the first generation in Russia have been decommissioned and withdraw from the structure of the Navy, and are now waiting for their turn to utilization. The first NPS К-3 is located at the quay-wall of the shipyard, and the question is being solved, about the sub’s transformation into a museum, like, for example, the first in the world NPS USS-571 “Nautilus” (USA). The reactors similar to those used for the submarines, installed on board the world’s first nuclear ice breaker “Lenin” were replaced in 70′s of the last century, and the ice breaker with the replaced plants has also served the second term already, and waits for utilization.

2.     Naval NPP of the second generation 

At the end of 60′s of the last century, the works began, directed towards designing and construction of NPS of a new generation. Accordingly, for these NPS, a new – the second generation of NPPP was created, not only differing from the previous generation by the design of reactor, but also by the design of all systems of the plant. The works for designing of the reactor were carried out in a special Design Bureau of OKBM (the city of Gorky). The new naval NPP created in the ОКBМ, differed essentially from a plant of the previous generation. The construction of series of NPS of the various projects, equipped with plants of the second generation was stopped only in 1990, but they are being operated till today. And the “OKBM” (Special Design Bureau of Machine-building industry) since that time has been a recognized leader in designing of LWR for the Soviet and Russian Navy.

During the second generation NPP development, the design and operating experience of NPPP of the first generation was taken into account. The new generation of NPPP accounted for the drawbacks of the previous one, and without a doubt, a large step ahead was made in the development of engineering.

After an accident with the first generation NPS (К-19), a concept of “design failure” was installed. As a maximum design failure, the leakage of the coolant in the primary circuit was considered therefore the basic problem was the minimization of the dimensions of the primary circuit. A so-called modular reactor plant was created, in which, on the covers of the 4 chapters of the SG, circulating pumps of the RCS[iii] were installed (see Fig. 2). Thus, with a failure of one SG or RCP, the power capacity of the plant was decreased by about 25%. The SG was allocated around the pressurized vessel of the NR, attached to the strong branch pipes manufactured according to “pipe in pipe” construction. The drawback of the scheme was that irrespective of whether there was a pump out of operation, or whether one section of SG had a leakage, it was necessary to disconnect the whole block RCP-SG at once. Later on, the scheme was improved and the RCP became to be attached to separate branch pipes, however, the problem of superfluous redundancy still was not solved for that generation.

During that time, no strict requirements were placed on the systems of localization of the failures in the RCS, and no possibilities were provided, of the NPPP heat removing, under the conditions of complete “full power loss” accident on the ship.

Doubtless advantage of the new arrangement solutions was the fact that the number of pipelines of large diameter was drastically reduced. All pipelines of the primary circuit were placed in uninhabited premises, under radiation shielding. That has allowed to considerably reduce the total dimensions of SGP, and naturally, improved the weight-dimensions parameters of NPPP.

NOTE:    On the other hand, the increased overall dimensions of the reactor unit required a more developed tank of metal – water protection and accordingly, the overall dimensions of the strong hull of NPS have grown drastically. For the sake of justice, it would be necessary to mention that it has resulted in a benefit for the experts in missile engineering. However starting from the second generation of NPS the dimensions of strong hulls amounted to about 10 meters in a diameter, and if not consider the arrangement of the weapons, that was exactly a dimension which can be the dimension of the containment for SNPPP or NPPP. 

Simultaneously, the operating and control systems (OCS), sensors, Instruments and control operating devices of the plant were improved substantially. The amount of the motor-operating valves, shutters, etc. was increased. That allows us to apply fully automatic algorithms for output PP in the accident. Although it is recognized that some redundant automation craze has not brought any substantial benefit.

The submarines of the second generation were initially designed with the sources of electrical power supply of AC. However, it required introduction into the circuits of special electrical machines, so-called motor-generators, or reversible converters (RC) of DC type to AC type electric power, and vice versa. TG (the basic sources of the electric power) became independent, instead of attaching them to the line of the shaft through special inflatable rubber muff, as it was on the first generation of NPS.

The following can be attributed to the drawbacks of NPPP of the second generation:

• Unsolved, for that moment, the problem of reliability of steam generator (SG) that was a technological problem, rather than a problem of design.
• The absence of passive safety systems, which, in principle, were not addressed during that time.

However strange this could appear, nevertheless, the problems of exploitation of the second generation were partially solved with an introduction of normative documentation regulating the rules of exploitation of naval NPPP. The presented statement specifies on the fact, that it is possible to solve technical issues not only at the expense of design of new, and improvement of old systems, but also at the expense of so-called organizational arrangements, establishing particular restrictions for the technical capabilities and exploitation of any technical systems.

3.     Naval NPP of the third generation 

For the sake of objectivity, it would be necessary to mention, however, that a part of the problems formulated before the designers and builders of NPS of the second generation proved not possible to be solved completely. In the middle of 70′s of the last century, the works began, for creation of NPS of the next, third generation.

At this time, new approaches to the problem of safety were formed in nuclear engineering. NPPP were considered to be the objects of an increased danger, and based on that fact, the new requirements towards the designing and construction of the ships and vessels with NPPP, and directly the power plants themselves, were developed. A separate concept was developed, for the creation of passive safety systems, including the emergency heat-removing system from the reactor core, and localization of failure in the SGP. For NPP of the third generation, the safety systems were designed for a maximum design failure, with an instant, guillotine rupture of the pipeline of coolant in the primary circuit, within the sector of its maximum diameter, in the area of the pressure branch pipe of the RCP.

For the PPP on board the ships of the third generation, an improved reactor design arrangement scheme was applied (see Fig. 3), thus allowing to principally resolve a number of important design problems:

• The scheme allowed obtains the natural circulation (NC) in the primary circuit under high enough levels of the power generation in the reactor core. This has been the determinative factor for the organization of circulation of the coolant and cooling of the core under a complete or partial “electric power supply loss” accident of the NPPP.
• The RCS pipelines were removed from the scheme, with their replacement with short branch pipes, combining the main equipment of the reactor unit into a unified, solid structure. Furthermore, the branch pipes are the elements of the pressurized vessel of the solid reactor unit. These elements are manufactured and delivered to the ship as a single construction of the SGP-block, after assembly, welding and QA under the conditions of the manufacturer of the reactor.
• The NPPP, in the structure, has a so-called passive emergency core cooling system (ECCS), which is automatically switched to operation in case of unavailability of the power supply, and allows carrying out the plant shutdown, without the use of external sources of electric energy – accumulator batteries. It allows to more effectively using the capacities of the battery for the liquidation of consequences of trips and failures, and for subsequent putting of the power plant into operation.
• In addition, serious problems regarding the creation of a new OCS for the plant were solved:
• Transition to new elemental base was carried out, and a new propulsion plant complex operating system (PP COS) was created.
• Algorithms of plant control were improved significantly, thus allowing carrying out its emergency shutdown without an involvement of the personnel.
• A new, essentially improved system of reactor control and protection was created. The pulse start-up equipment (SE) was used, which allowed to monitor the condition of reactor at the levels of power capacity of 10^-8%, including the monitoring in a sub-critical condition.
• The compensating elements, in case of unavailability of the electrical power supply, have a possibility to go down to the bottom supports, by a self-propelled motion, with a high speed. Thus a complete “damping” of chain reaction in the reactor occurs, including the case of subsequent turnover of the ship with locking of compensating elements in the lower position.
• During the improvements and modernization, a multitude of other innovative technical solutions, which increased the reliability and safety of operation, was implemented in the plant and the equipment. A modular configuration of the NPPP and unitization of individual systems have allowed to reduce its overall dimensions, increasing, at the same time, its power capacity, survivability and other exploitation parameters.

In some publications, a point of view is expressed, that the problems of NPPP of the third generation remained as the problems of reliability of the principal equipment, and first of all, the core, the SG, passive emergency core cooling systems (ECCS) and other systems. However, for the sake of justice, it would be necessary to note, that these problems frequently had an exploitation character, and were associated to a higher extent with the modes of use of the plants, and the problems arising in navigational repairs, rather than with the design and technology of construction of NPPP and NPS, especially of the latest, modernized projects.

4.     The fourth, and subsequent generations of naval NPP 

The acquired experience of designing, construction and exploitation of NPP allows making confident conclusions about the safety of the plants with LWR similar to naval ones. It opens vast opportunities for further development and improvement of the plants of such type, at least within the framework of the existing projects and their modernization. The researches within the NC capabilities allowed create an NPPP of the fourth generation. Reactor it self is a mono-block design or so-called “integral or modular” construction, under the name of “mono-block reactor”, or, of cause “small modular reactor” (SMR) (see Fig. 4). The basic advantage of such an arrangement has been the localization of the coolant of the primary circuit within the volume of the vessel of mono-block, at the expense of the allocation of the SG, directly inside the vessel of the reactor, with the absence of branch pipes and strong pipelines of a large diameter.

These new reactors design were created with the considerations of all the modern requirements of nuclear safety, and for today, they supersede all the systems known in the world, in part of passive safety and the reliability of operation at all the levels of generated power. The principle problem which faced the constructors of the integrated power plant – the provision of reliability of SG and protection of their material against radiation, as the sections of the SG are located in a direct proximity to the core, and they are not separated from the later by a thick-walled elements of the vessel of the reactor block, which essentially lowers the influence of neutrons- and hard gamma-radiation.

The scale model of power plant of a similar type, MASLWR, of the design of INEEL (INL today), for civil applications, is started to be subjected to testing in the USA, after 2003 but did not finalized yet. The similar researches are also under way in Russia, since the 90′s of the last century.

It might be possible, that the presented direction of development of LWR reactors has exhausted, and a further development of modern naval NPP should be associated, first of all, with a transition to principally new types of reactors, with other types of coolant in the primary circuit, with the parameters of RCS, or other thermal circuits and schemes. However, for example, despite the existing plans of developments of perspective reactors, the maximum number of researches continues to be carried out exactly in the direction of improvement, optimization, and further examination of namely the LWR type reactor. Furthermore, according to the forecasts of the scientists, in the USA, exactly the reactors of LWR type will dominate in the market within the foreseeable future, and even not because of the fact, that there have been a lot many of that kind that have been built already, and of many different projects, but because of their high reliability and technological effectiveness.

[i] Since the largest number of NPPP was constructed and operated exactly in the Naval Fleets of the various countries, further in the text the term “vessel’s” will be finally omitted, and the term “ship’s” will be used.

[ii] In the presented diagrams of SGP systems, no SGP systems showed which operate together with RCS of the power plant.

[iii] In fact, a few variants of layout solutions of reactor power plants, differing by the number of their SG, RCP, and the places of their location, could be attributed to the second generation.