The reactor with the concrete well equipment, pressurizer system and emergency core cooling system (ECCS) is installed inside the hermetic guard vessel (GV), designed for 0.5 MPa internal pressure, Fig. 3. A part of the ECCS and the reactor water clean-up systems are arranged in the compartments, joined to the GV by pipelines.

The reactor, the refuelling, water pool, the emergency feedwater storage tanks and other equipment is installed inside the containment.

MA CHANGWEN, BO JINHAI,

JA HAIJUN, GAO ZUYING

Institute of Nuclear Energy and Technology,

Tsinghua University,

Beijing, China

Abstract

A Series of tests has been done for the three cases ( breaking pipe opened in the gas plenum, near the liquid level and submerged in the water) in the Test Heating Reactor.

Experiments show that the three cases of LOCA ( Loss of Coolant Accident ) have different patterns. In the case of a breaking pipe connected into the gas plenum the quantity of lost water is indepondent of the diameter of the breaking pipe. In the case of a breaking pipe located near the liquid level, the quantity of lost water depends on the location of the pipe. In the case of breaking pipe submerged in the water, all water above the break will be discharged.

The patterns of the discharge for the three cases are given in the paper.

Key words: Loss of coolant; Nuclear heat reactor

1. Introduction

District heating reactors must be built near the heating load ( heat grid ) . That means they must be built near cities. So the safety requirements for DHR are very strict.

For integral reactors protection against loss of coolant is one of the most important considerations.

On the pressure vessel of a integral reactor, there is no penetration pipe with big size, LOCA is possible only through pipes with small sizes. The possible ways of LOCA are as follows:

Ф Water discharge through the safety valve.

CD Water discharge caused by a pipe breaking ( The pipe is opening in the steam plenum ) .

® Water discharge caused by a breaking pipe (the pipe is opening under the water level.

For checking the design safety of the 5MW Test Heating Reactor ( NHR-5 ) and investigating the characteristics of LOCA during the discharging process, a series of tests have been done on the model thermohydraulic loop in INET.

The test model loop was built on the base of similarity theory for the NHR-5.

The loop consisted of two circuits. The first circuit included two heating test sections (to model fuel assemblies ) , a rising pipe (to model the chimney) , steam-water separator, steam condenser, cooler, down pipe, valves ( to model resistances ) and measurement apparatus.

The Condenser and cooler are cooled by the water of the secondary circuit. Secondary circuit water transfer heat to the environmental air by an aircooler.

In order to study the effects of drain position on the drain process, three positions of the drain orifices have been selected. The first drain orifice is located at the top of the loop. It is used to model the case of “ safety valve opened and can’t be reset” . The second orifice is located in the steam plenum of the separator. It is used to model the case of a breaking pipe opened in the steam plenum. In the NHR-5, most pipes penetrating the pressure vessel are connected to the steam plenum. The third drain orifice is located under the water level. It is a pipe, opening under the water level at about 3m. It is used to model the case of a boron-injection pipe breaking.

The diameter of the heating rods, heating length, height of the chimney, resistance coefficient at the inlet of the heating section, total resistance of the loop, density of the heating power and parameters are the same as those in the 5MW Test Heating Reactor, in order to keep the geometric and hydraulic similarity conditions.

In the water discharge test through the safety valve, according to the geometric similarity, the diameter of the drain orifice should be selected at 4. 3mm. To study the relationship between the orifice

diameter and quantity of drain water, besides 4.3, orifice diameters 6.0 and 8.5 m m are also selected in the tests.

During the test, the pressure of the steam plenum, water temperatures at the inlet and outlet of the heating section, drain water flow rate, and surface temperature of the heating elements are measured. Pressure and pressure defferences are measured through transducer type 1151. Drain water quantity is measured by weighting the condensate and measuring the water level.

The base of reactor type choice is whether it will meet the above-mentioned design objectives, technology reality and feasibility and users requirements’ or not

Pressurized water reactors (PWR), boiling water reactors (BWR) and high temperature gas-cooled reactors (HTGR) can all be used as the co-generation plant Comparing with the other types of reactors, for PWR there is a lot of experience in experiment, research, design, manufacture, installation and operation in China.

The steam supply parameters are also the base of reactor type choice. According to the steam parameters which a few large Chinese technological process users need at present, the majority is steam in the middle or in low pressure. From an application of view, PWR and BWR can all meet users’ requirements. The primary circuit pressure of a BWR is usually 6.86MPa. Its core outlet temperature is 285 *c, while usually the core outlet temperature of a PWR can reach 310-325 which can produce medium pressure steam in the third circuit The PWR has a larger application range than the BWR.

In summary, a PWR should be chosen as the nuclear co-generation plant in China, and the steam supply capacity is usually no more than lOOOt/h, so the single-reactor thermal power will not be more than 600MW. A double-reactor in a plant would be better suited to our conditions in which heat and electricity are supplied at the same time and how much the electricity will be produced is based on how much the steam is consumed.

11 «residual heat removal pump 12.s&fety lnjootlon ршпр 1 3«spray pump 14.refueling water tank 15«high pressure turbine l6.1ow proesure turbine

17.moisture separator and reheat 10.condenser

19. hydrogen Ion oxchangor

20. ulxod bod

Я..uliod hod 32.ілпк. lip „„tor for third circuit M. tn boron reevelo svatom

A—A •

I

D direction vie*

J. ГіііІіі |>uoip

і Z. reactor core 1 reactor vesseJ

4. heat eicbiarer

5. CKOH ‘

6. second side taler inlet

7. second side tiler outlet

Inherent Safety Concepts

After Three Mille Island-2 and Chernobyl-4 accidents within the nuclear reactor society in the world there are groups who are developing ideas of a new generation of reactors which is characterized by a much higher standard of safety incorporating forgiving reactor system and inherent safety design. This set of inherent characteristics by definition shall tolerate any mistake occured during operation. It can mean that in an accident condition the inherent characteristics could allow enough time for the operator to correct the mistakes, could respond to neutralize the hazard, and could be left safely without any human interference.

At present it has been recognized that inherent safety characteristics are the dominant advantage of High Temperature Gaz Cooled Reactors which incorporates multicoated ceramic particle fuel, inert helium coolant, high thermal capacity and good convective heat transfer whereas, PIUS design concept are manifesting the solution for light water reactor versions.

It is generaly accepted among the nuclear reactor designers that the technology of "next generation" reactors must provide a guarantee that the core degradation accident and consequences risk of serious radiation releases will not occur. A core degradation accident can be defined as one where there is widespread break of the cladding and large scale release of fission product inventory to the coolant. The upper end of this accident is of course a complete core melt.

Many ways can be undertaken to prevent a core degradation accident. The following two conditions must be fulfilled :

• keep the core submerged in water at all time to maintain integrity of fuel

• make sure that heat production does not exceed cooling capability of water

• This two conditions could be achieved by placing the core within the neutron poisonned coolant pool. The last will submerges the core when needed by passive nature. Core cooling is effected by evaporation this water.

In the other point of view, in order to eliminate one the most important initiating event of the core degradation accident, i. e. loss of primary coolant accident, the integration of Steam Generators and pressurizer in the reactor pressure vessel has been considered. In such reactor design, due to the absence of large diameter external piping associated to primary system, no large break LOCA has to be handled by the safety system.

Moreover, the simplification design is also considered in some integral type reactors by eliminating control rods. The power output controlling is replaced by adjusting the flow rate and the concentration of boric acid solution. In this case, no reactivity initiated accident should be considered any more.

This new design, which is a complete departure from current Water Cooled Reactor plants, exhibits a remarkable inherent safety characteristics for all accident sequences. However, it creates a new set of complex machinery and operational problems of its own.

Table 2 shows the features of four integral reactor design concepts.

The experimental investigations using MM models were preceded by numerical ones mainly aimed at the selection of geometric and operating regime parameters for the MM and reactor unit as well as the prediction of their behaviour under accident conditions. In the process of calculations for normal operating conditions, the limiting MM power value was obtained and the dynamics of natural circulation development at varying power, etc. were determined.

Calculations were also performed for typical accidents such as the disrupture of the pipeline between the MM and the pressurize, the MM vessel failure and others. In view of the complexity of hydrodynamical and heat transfer processes under such condition some problems should be experimentally verified. These questions will be touched upon later when discussing experimental data in Section 4 and now we shall consider some calculational results for beyond-design accidents in more detail.

In a beyond-design accident with postulated instantaneous MM uncovery, under the reactor emergency shutdown conditions, the maximum fuel temperature at a fuel assembly power of 1070 kW was 1070 C. The fuel melting takes place when the pipeline connecting the MM with the pressurize breaks just near the MM vessle with the simultaneous failure of the reactor emergency protection or when the MM vessel breaks simultaneously with the emmergency protection system failure and the disrupture of the pipeline between the MM and pressurizer. In both cases, the fuel melts only in one MM subjected to the failure.

In beyond-design accidents accompanied by the failure of secondary circuit heat removal systems, the absorber cooling system ensures that the maximum fuel element cladding (1200 C) was not exceeded. If this system fails also, the element temperature of 1200 C is riched for ~ 12 hours because of the large heat capacity of the reactor. During this time it is necessary to take adequate measures to monitor the accident process. Table 2 gives an idea on the radiation exposure as a consequence of different normal and accident conditions.

Research and development activities on major component design are concentrated in the steam generators. The heat transfer features of the tubes have been tested at CAREM operational conditions in a thermohydraulic loop. A full scale model of one steam generator is scheduled to be tested next year.

Dimensions of RPVs are dependent on reactor size. The largest integral reactor pressure vessels that are currently considered (SPWR, SIR, VPBER-600) have the following dimensions:

Diameter 6.5 — 7.2 m,

Height 20 — 25 m,

Wall thickness 265 — 280 mm

(cylindrical part).

These dimensions are comparable with the largest pressurized water reactor (PWR) pressure vessels (diameters, wall thicknesses) and boiling water reactor (BWR) pressure vessels (heights, diameters). Some designers have made enquiries with potential manufacturers and they received positive answers on the possibilities of manufacturing these large pressure vessels. Existing manufacturing technologies in the following areas can be used in the manufacture of integral RPVs

• materials

• forging of semi-products

• welding and cladding

• machining

• inspection and testing

Guard vessels/containment can be manufactured at existing manufacturing facilities. No specific problems have been identified concerning the manufacturing of these large components also The feasibility of RPV transport may be an important issue to be taken into account in site selection Access by water can solve most transport problems

4 3 Primary circulation

Natural circulation of primary coolant is an inherent feature of the integral reactor arrangement due to its simple configuration and low hydraulic resistance of the primary circuit Reactor coolant natural circulation has reliability, simplicity and safety advantages These advantages override economic considerations at lower unit powers As the power level increases, economic considerations become more important and hence forced circulation may be preferred However, under special conditions, for example in marine reactors, forced coolant circulation is utilised even in units with a low rated power

For heat only reactors of any power level, natural circulation seems to be the most preferred solution due to a lower core power density and high reliability requirements typical of this kind of reactor

Cost considerations and technological limitations in RPV manufacture appear to limit the use of natural convection cooling At present integral reactors with reactor coolant natural circulation are limited in their power level to 1000 MWt

D. J. LEE, J. I. KIM. K. K. KIM,

M. H. CHANG, K. S. MOON

Korea Atomic Energy Research Institute,

Taejon, Republic of Korea

Abstract

An integral reactor concept for nuclear cogeneration plant is being developed at KAERI as an attempt to expand the peaceful utilization of well established commercial nuclear technology and related industrial infrastructure such as desalination technology in Korea. Advanced technologies such as intrinsic and passive safety features are implemented in establishing the design concepts to enhance the safety and performance. Research and development including laboratory-scale tests are concurrently underway to evaluate the characteristics of various passive safety concepts and provide the proper technical data for the conceptual design. This paper describes the preliminary safety and design concepts of the advanced integral reactor. Salient features of the design are hexagonal core geometry, once-through helical steam generator, self-pressurizer, and seismic resistant fine control CEDMs, passive residual heat removal system, steam injector driven passive containment cooling system.

1.0 INTRODUCTION

The drought experienced due to the climatic anomalies and the worsening level of pollution have reduced inland water resources significantly for a number of years. A nuclear co-generation plant which can be used for sea water desalination as well as electricity generation can provide a solution in some coastal countries such as Korea and middle east nations. In this regard, Korea Atomic Energy Research Institute (KAERI) has undertaken a study for the development of advanced integral reactor for the application to these purposes as an attempt to expand the peaceful utilization of nuclear energy.

Most of power reactors that are currently in operation and under development have loop type configurations which enable large-scale power output and thus provide economical power generation. On the other hand, integral reactors receive a wide and strong attention due to its inherent characteristics of enhancing the reactor safety’ and performance through the removal of pipes connecting major primary components. Small and medium reactors with integral configurations of major primary components are actively being developed in many countries. The design concepts of those reactors vary with the purposes of application.

KAERI has been putting efforts to research and develop new and elemental technologies for the implementation into the advanced reactors. In parallel with those efforts, an advanced integral PWR with implementation of those technologies as well as passive safety features is under conceptual development.

The reactor power of 300 MWt is considered as suitable size for energy supply to the industrial complexes, remotely located islands, and especially isolated area. The reactor core is conceptually designed with no soluble boron and hexagonal fuel assemblies to enhance the opertional flexibility and to improve the fuel utilization. The reactor safety system primarily functions in a passive manner when required.

This paper describes the conceptual design features of the advanced integral reactor under development at KAERI, and also important R&D subjects concurrently in progress in order to prove and confirm the technical feasibility of the design concepts.

A. N. ACHKASOV, G. I. GRECHKO, O. G. GLADKOV, V. L. PAVLOV, V. N. PEPA, V. A. SHISHKIN Research and Development Institute of Power Engineering, Moscow, Russian Federation

Abstract

The paper describes the conceptual design of an integrated advanced safety nuclear reactor unit for a low capacity floating NPP designed for power supply in areas which are remote with difficult access.

The paper describes the major structural and lay-out components of the steam generator and reactor units with main technical characteristics.

Conceptual design of a reactor facility with enhanced safety for a low capacity floating and environment friendly NPP has been developed in Research and Development Institute of Power Engineering to provide electricity supply to areas which are remote with difficult access.

The most advanced technical solutions well mastered during recent years in designing of navy nuclear power facilities have been used. —

The following technical solutions provide high safety level of the reactor facility:

• application of well mastered technology of water-cooled reactors with developed inherent safety features;

• location of the total primary circuit equipment into one vessel of an integrated nuclear steam supply system (NSSS);

r • provision of defence — in-depth barriers to prevent ionizing radiation and radioactive fission products release into environment, realization of technical measures to protect confining barriers and maintain their effectiveness;

• application of safety systems, mainly based on passive operation principle;

• independence from external power sources.

The Development of the reactor facility (RF) for floating NPP was carried out in compliance with current Russian rules and requirements to provide safety of stationary and floating nuclear power plants and ship nuclear power facilities and in accordance with the modem notion of prospect enhanced safety NPPs, elaborated so far by the world community. It was considered that RF met safety requirements, if its radiation effect on personnel, population and environment in normal operation and design-basis accidents did not lead to excess of specified dose rates, and limits this effect in beyond design-basis accidents. Technical and organizational measures were assumed to ensure safety with any design-basis initial event with superposition of one failure independent of the initial event of any of the following safety system components: active or passive component having mechanical — movable parts or one personnel error independent of the initial event. Besides one failure independent of the initial event, a nondetectable failure of components, not monitored during operation, which influences accident propagation, and results in violation of safe operation limits is taken into account.

• Fig.1. Reactor facility flow diagram:

1 — core; 2 — steam generator; 3 — pressurizer; 4 — primary circuit, electric circulation pump; 5 1

steam-generating system vessel; 6 — iron-water shielding tank; 7 — bubbling tank; 8 — high pressure gas cylinder; 9 — emergency flooding cylinder; 10 — emergency cooldown tank; 11 — heat-echanger-condenser; 12 — safeguard housing; 13 — containment.

1 — steam-generating system No.1; 2 — steam-generating system No.2; 3 — isolating valves for steam and secondary circuit feedwater; 4 — emergency cooldown tank; 5 —

emergency flooding cylinder; 6 — strong leaktight partition.

Structurally the RF is divided into two separate and independently operating steam generating facilities with identical principal flow charts and equipment (see Figs. 1-3).

The RF employs water-moderated, water-cooled reactors with inherent safety and control features due to negative power and temperature reactivity coefficients. The core physical characteristics are so selected that the above coefficients be negative in the entire range of temperatures during core life. This eliminates spontaneous reactor power excursion in normal startup and heatup and stabilizes operation in steady-state conditions and transients.

The adopted core structure rules out the possibility of forming local critical masses both in normal and emergency modes. Formation of secondary critical mass in hypothetical accidents, resulting in partial or full core melting is also excluded.

All equipment of the primary circuit (the core with reactivity compensation components, steam generator, electric circulation pumps, pressurizer) except gas cylinders are located in cylindric vessel of the integrated nuclear steam supply system (NSSS) (see Fig. 4).

Fig.^ Steam-generating system:

1 — vessel; 2 — ring-type cover; 3 — core; 4 — steam generator; 5 — pressurizer; 6 — intermediate capacity; 7 — electric circulation pump; 8 — control and protection

members drives.

The pipeline between gas cylinders with pressurizer and make p system pipelines is connected to the NSSS cover and ends at this point.

Such NSSS structure allows:

• to reduce the number and extent of external primary circuit lines down to minimum and reduce probability of its depressurization;

• to provide high level of natural circulation for the primary circuit coolant;

• to increase water volume above the core and to improve its cooling conditions in a hypothetical accident related to primary circuit depressurization and water cut-off from the NSSS.

In order to decrease the leak and mitigate accident consequences in case of pipeline rupture which connects pressurizer with receiving cylinders, a constricting device is installed on it. Check valves are installed on makeup pipelines (in places of its attachement to NSSS cover).

RF employs defence-in-depth principle, based on barrier system to prevent ionizing radiation and radioactive substances release into environment supported by technical procedures to protect these barriers and maintain their efficiency.

In accordance with this the RF safety systems provide for:

• reactor emergency trip and keeping it subcritical;

• emergency heat removal from the core;

• confining of radioactive products within specified boundaries.

Flow chart and design of these systems are based on their passive action as much as possible. In order to ensure reactor emergency trip and keeping it subcritical the following systems have been designed:

• main emergency protection;

• additional emergency protection;

• liquid absorber injection system.

The main and additional emergency protections have different activation principles. They are activated automatically by CPS and complex system for technical means control (CSTMC) signals or by power cut-off.

Liquid absorber is injected by remote control.

Design of all compensating group (CG) activation devices is based on insertion of shim rods into the core using dump springs or the rods weight during power cut-off from actuators and CPS; the design also ensures keeping shim rods in inserted position in case of the unit turning-over.

The shim rods ensure reactor tripping and keeping it subcritical in all operational modes in case of one (any) CG failure (stuck in the extreme position) by means of CGs remaining in operation.

Additional emergency protection is activated automatically by a signal from CPS in response to failure of two or more CGs.

Besides activation by CPS signals the additional emergency protection becomes operational without any external signal under increase of primary circuit coolant temperature up to specified ultimate value. This is achieved by the additional emergency protection design.

Liquid absorber is injected into integrated NSSS only when the reactor has not been transferred to subcritical state due to failures of the main and additional emergency protections, but such situation is hardly probable.

Availability of the listed means for reactor emergency protection and keeping it subcritical can provide safety reactor trip not only in case of design-basis accidents, but also in hypothetical accidents, which are not related to core or NSSS damage.

In order to avoid spontaneous chain reaction during scheduled maintainance or repair works, CG actuator design envisages clutches, which are installed below "cold" startup position. Electromagnetic clutches are opened only by signals, which allow the core start-up.

‘ To provide heat removal from the core in different emergency situations the RF has the following safety systems:

• emergency cooling system;

• emergency core flooding and cooling system.

RF can be cooled down by feedwater pumping into the steam generator by the steam-turbine means in both operational and emergency modes. In case of primary circuit depressurization the makeup system is also involved in heat removal from the core.

In emergency situations, which are not related to steam turbine failure, the RF is usually cooled by feedwater pumped into the steam generator by steam turbine. In case of the steam turbine failure and during power cut-off the integrated NSSS is automatically disconnected from the steam turbine (from steam and feedwater supply) and emergency cooling system (ECS) starts operating. Heat removal from primary coolant is provided for both operating and non-operating primary circuit electric circulation pumps (due to natural coolant circulation). In case of RF power supply loss the disconnection of NSSS from steam turbine and ECS activation are ensured by isolating and cut-off valves, which perform this function in "normal" position.

Emergency cooling system, consisting of 4 independent sections, can provide NSSS cooling during 24 hours in case of failure of any 2 ECS sections even with the plant blackout. Heat removal from primary coolant and its transfer to water in ECS tanks results from coolant natural circulation in this system. With electric supply availiable the water in ECS is cooled by pumping it through heat exchangers, which are cooled by outboard water. In case of NPP power blackout heat from ECS heat exchangers is removed due to heating followed by evaporation of water from ECS tanks.

Core makeup and core emergency flooding systems provide removal of residual heat from the core and eliminate its drying and further melting in case of primary circuit depressurization and coolant circulation failure. Each of these systems has redundances with respect to equipment and water supply channels to NSSS. The makeup system operates automatically by pressure drop in primary circuit signal from CSTMS and activation of the emergency core flooding system does not require any operator actions or CSTMC activation and occures due to membrane rupture under primary circuit pressure decrease down to specified level.

Taking into account that the makeup system besides RF normal operation assurance is supposed to provide safety, three independent NSSS water supply channels with high pressure pumps in each channel are envisaged in the makeup system. Water for primary circuit makeup is taken from condensate-feedwater system of steam turbine or from water storage tanks, and after water level in bubbling tank reaches specified level the makeup is performed in closed cycle, i. e. water for makeup is taken from bubbling tank, cooled in heat exchangers and delivered into NSSS; the coolant, flowing from NSSS again comes into the bubbling tank.

The emergency core flooding system is based on passive operation principle and comprises two independent channels. Water is supplied into NSSS through these channels due to gas pressure from pressure cylinders.

In case of primary circuit depressurization the feedwater is delivered into steam generator from steam turbine in parallel with operation of makeup and emergency core floodibg systems. This ensures additional removal of heat, released in the core and increases time needed for the core drying in hypothetical accidents due to natural convection of gas and steam in NSSS and partial steam condensation on steam generator surface. Steam turbine failure to supply water into the steam generator activates the ECS.

The availability of cooled iron-water shield around NSSS vessel and favourable conditions for natural circulation of steam-gas medium inside NSSS thereby reduce probability of core melting in hypothetical accidents and eliminate possibility of NSSS vessel melting down in such accidents.

In case of an accident, all actions to localize it are automatic. Control systems (CPS and CSTMC) have three-channel scheme and the control and emergency signals are generated using major principle (2 out of 3) that ensures sufficient operational reliability.

The following safety barriers are envisaged to reduce radioactivity release into environment in design — basis and hypothetical accidents:

• corrosion resistant matrix of fuel elements;

• claddings of fuel elements;

• leaktight primary circuit;

• safeguard vessel;

• containment.

Small number of the equipment, its lifetime characteristics and high automation degree ensure RF operation during one year without maintenance within the containment. This allows to adopt additional technical measures to enhance RF safety. Namely, ventilation system equipment located outside the containment is isolated with isolation valves designed for maximum emergency pressure in safeguard vessel, and is made operational only when it is necessary to make repairs inside the containment (gas is discharged from containment into environment via special filters under control). Conditioning system operates in closed cycle, its total equipment is located in safeguard vessel. Such technical solutions eliminate radionuclides release from containment in normal operation of floating NPP.

Leaktightness of the containment and safeguard vessel and no inside repairs allow to maintain decreased oxygen content (11-13 %) in equipment rooms, thus enhancing their fire safety.

The RF design solutions eliminate a possibility of safeguard vessel and primary circuit damage in case of earthquakes, aircraft crash or other external impacts.

Due to high safety level (probability of severe beyond design-basis accidents, which can result in serious core damage or melting and following radionuclides release into environment does not exceed 2х10я5-8 яОрег reactor per year) the RF can be recommended to be used for low capacity floating NPP, and its unique mass-dimension characteristics enable to construct a plant with such a draught (estimated as 2.6 m), which will make it possible to ship it along Nothern rivers to regions which are far away from the coast. 1

The steam generator is an once-through vertical cassette surface-type heat exchanger consisting of straight-tube steam generating elements, where steam with the required parameters is generated due to the heat from primary circuit coolant. The flow of media in the steam generator is counter-current: primary coolant flows downwards between tubes inside block-section shrouds, secondary medium (water-steam) flows upwards inside the steam generating elements.

The design of the steam generator is based on know-how decisions which include:

1) chemical composition of tube system materials;

2) design of steam generating elements;

3) method of spacing of the steam generating elements;

4) design decisions for assuring SG hydrodynamic stability;

5) the method of assembling the steam generating elements into the block-section;

6) the device for fixing and sealing the SG block-section into the reactor vessel internals;

7) the technique for obtaining a strong, leak-tight joint of titanium alloy and steel.

The SG consists of 12 identical block-sections 1 (see Fig. l), located uniformly

in the annular space between the reactor vessel and core barrel.

Each block-section is individually isolated by valves for feed water and steam.

The block-section consists of steam generating modules (cassettes) 2, header 3, feedwater tubes 4, steam tubes 5, shroud 6, nozzle with the seal 7, strong and leak — tight joints 8 and 9.

Steam generating elements are assembled into steam generating modules 2. Module groups are united into independent subsections, supplied with feed water individually.

The block-section header 3 is intended to organize feed water supply, heat removal and block-section fixing in the reactor vessel. 18 holes for welding of feedwater tubes are located over the header centre and 18 holes for the attachment of steam tubes are located over the periphery.

Feed tubes 4 are in the compartment 10, and intended for the supply of feed water from the header 3 to subsections.

Steam tubes 5 serve for steam removal from the subsection to header 3.

The shroud 6 embraces steam generating elements of the block-section and serves for:

1) the organization of primary coolant flow;

2) forming of specified geometry dimensions of the block-section;

3) tightening of the steam generating elements to exclude vibration.

The shroud is fixed to the block-section header.

The nozzle with sealing 7 is located in the lower part of block-section shroud and intended for prevention of coolant leakage bypassing the steam generating elements and

1. Block-section

2. Module

3. Header

4. Feedwater tube

5. Steam tube

6. Shroud

7. Nozzle with the seal

8,9. Strong and leak-tight joints

10.

Compartment

for restraint of the lower block-section part from latteral displacements. The seal design assures longitudinal temperature displacement of the block-section.

Strong and leak-tight joints 8 and 9 ensure the connection of steam generator items, made of titanium alloys to the stainless steel items and are located on each feedwater tube 4 and steam tube 5.

The block-section is installed into the reactor vessel and headers 3 are welded to the vessel from outside.

The steam generator design permits inspection by non-destructive methods during fabrication, including 100% radiographic inspection of all welded joints subjected to primary and secondary pressures.

Unification of items and assembly components by block-section, modules and steam generating elements, application in steam generator design of a minimum quantity of items of standard sizes and assembly of components of different types and dimensions allows organization of parallel process flows for automated manufacturing. This assures high quality of a steam generator and reduces the fabrication cycle.

The steam generator operates as follows. Primary coolant from the pressure chamber enters the shrouds of block-section 1, flows downward between steam generating elements transferring heat to the secondary medium. In the lower part of the shroud the coolant leaves the block-section through the nozzle with a seal 7 and goes further to the core inlet.

Feedwater enters feed tubes 4, then it enters modules 2 and is distributed between steam generating elements. Then water goes upward through the steam generating elements and is converted to superheated steam. Steam from the modules passes through steam tubes 5 and holes in the block-section headers and is supplied to steam chambers and removed from the steam generator.

The steam generator design gives the possibility of diagnostic control in operation and during sheduled shut downs of the reactor by the methods adopted in Russia. It allows estimation of the real state of the steam generator and to ensure its operational safety.

The steam generator design is maintainable and allows to leakage wear in and to isolate any non-leaktight module in case of intercircuit leakage. If one module is isolated, the heat exchange surface is reduced by 1/216 part.

The modular structure of the steam generator allows performance of complex representative testing of steam generator including the confirmation of lifetime characteristics in the rigs of relatively low power.

In the steam generator project design decisions are realized which have resulted in limitations of cross-sections, through which primary coolant leaks in case of some structural failures, including the rupture of a feed tube sheet or of the nozzle in the vessel.

Examination of various SG element failures has shown that the maximum equivalent diameter of the leak is between 5-40 mm.