The Reactor Vessel is of cylindrical shape with hemispherical heads.

The construction material is low-alloy carbon steel, internally lined with austenitic stainless steel.

The main openings of the Reactor Vessel are the water/steam nozzles and the two connections to the Pressurizer.

Core

The reactor core consists of 69 typical (17 X 17) PWR fuel assemblies with a reduced length to limit pressure losses.

Steam Generator Unit (SGU)

The SG features an annular tube bundle with helicoidal tubing.

The steam is generated tube-side. The feed water piping is connected to feed water headers, located symmetrically inside the reactor vessel within a calm zone, provided each with two tubeplates laid out vertically. The tubes depart circumferentially from the tubeplates.

A similar arrangement is provided at the top for the two steam headers connections.

The vertical arrangement of the tubeplates aims at preventing crud deposition at the tube-to-tubeplate connections, where the corrosion is likely to occur.

The higher outer rather than inner tube pressure, a reversed situation with respect to a conventional SGU, reduces the risk of flaw growth in the tubes.

Primary Circulation Pumps

The two Primary Pumps of the variable speed, glandless, wet winding type (like the pumps manufactured by Hayward Tyler Fluid Dynamics) are fully enclosed within the Reactor Vessel. The pump motor is cooled by the water of the Intermediate Plenum.

Above Core Structure (ACS)

The ACS, shaped like a flat-bottom cylindrical glass, provides the support for the core instrumentation and forms the inner wall of the annular riser of the primary water. The ACS is open at the top. The water within it is part of the intermediate plenum and this helps to limit the primary water inventory in the reactor module to a minimum. The ACS is flanged to and suspended from the top of the Inner Vessel for easy removal to allow standard fuel handling.

Pressurizer

The Pressurizer is of a slim cylindrical shape with hemispherical heads.

The pressure control function is carried out in the upper part, which is externally insulated to limit heat losses from the steam and hot water plena.

The remaining bottom part contains a cold water plenum, hydraulically connected to the upper hot water plenum by means of a number of pipes.

The function of the pipes is to enhance mixing of the hot water with the cold water, in case of water flow towards the reactor vessel during transients.

One of the possible trends in further improvement of the reactor of the proposed design is to use a micromodule with boiling water in the primary circuit. Fig.5 illustrates the experimental results of an experiment in which the micromodule is filled with water to a certain level (V= 90, 80 and 65% of the total free micromodule volume) and cut off the pressurizer. Thus, the pressure balance under heating-up conditions is due to the gas (steam-gas) volume being in the upper part of the micromodule. It is evident from the figure that a micromodule power of 1070 kW can be reached with V0= 80% at a pressure of 5.3 MPa. With filling V0 = 65 %, such a power is reached at a pressure of З. ЗМРа; the coolant flow rate of the primary circuit with boiling water in both cases being higher than under the non-boilin conditions. It is necessary, however, to note that flow rat fluctuations were observed in a restricted range of powers (e. g. with a amplitude of 1.5 % at V 80 % within the range of power from 250 to 350 kW and 15 % with V = 65 % at a power up to 150 kW.

Principal semiproducts for pressure vessels are forged rings and plates. Plates are subsequently hot-stamped to the form of spherical or elliptical dishes for bottoms and covers.

Rings for integral RPVs will be about 4m long. 6-7m of outer diameter and 250 — 300 mm thick. flange-rings will be shorter with the wall thicknes about 500 mm — There is not such need of complicated thick-walled nozzle rings as at PVR pressure vessel. Plates for dishes will have to have diametres about 10m. It is not excluded to make the part by welding.

Existing procedures of heat — treatment of forgings for quality Cquenchlng. tempering) give high certainty in reaching homogenous structure and mechanical properties even in very thick valIs.

Velding

Vessels are welded from individual rings and dishes by circumferential welds. Automated submerged-arc welding to a narrow gap is generally applied. A promissing method could be electron beam technique which is applicable to large material thicknesses and allows the performance of high-quality welds in short time spans.

The simultaneous heating by gas flame or by electricity is applied during the welding.

The number of circumferential velds in an integral RPV will be 4-8.

The necessary welding equipment consists of a welding automat, rotating positioners of an appropriate loading capacity and of a heating device (electric heaters or gas torch). Monitoring and registration of welding parametres is required.

Cladding

Cladding is performed by automated submerged-arc method by using strip or wire (multiple wire) electrodes: plasma technique is a pespective method.

Following i-equlrements have to be fulfilled when cladding1 •

— the cladded surface must be clean. vithout impurities and sulphur Inclusions which is assured by piercing ingots by the mandrel which removes the central part of the ingot containing segregates, or by using hollow ingots.

— proper laying of individual beads and their mutual overlapping,

— an optimized heat input which is important for the depth and properties of the heat-affected zone under the cladding

All the measures are aimed at eliminating the formation of underclad cracks.

The cladding equipment consists of positioners (rotating ones for cylindrlc rotating ones for spherical eliptical and heating device.

The philosophy of containment for integral reactors is the same as that for loop type designs The containment has the following functions

• Protection from the effects of internal events, especially, retention of radioactivity released from the core

• Protection from external impacts

Single and duel containment designs are possible The low discharge rate from LOCA in integral reactors, results in pressure suppression systems being very attractive The guard vessel concept where the size of the leak tight containment is minimized to a shell which fits closely around the RPV has been developed for integral reactors It provides simplicity of design and has positive benefits in accident management

The guard vessel is usually made of steel but could also be made of pre — strressed concrete However, the emphasis on improved safety characteristics of integral reactors has led to a preference for steel, rather than steel lined concrete as it allows for ultimate heat removal by heat transfer through the steel All containment designs share the general objective of plant size reduction

The guard vessel concept can be applied to integral reactors, but not to loop type reactors since the whole of the primary circuit including CVCS is compact and can be contained in a vessel of reasonable size The advantages are

• Very effective containment of radioactive species both in normal operation and in upset conditions

• Possibility of reduced specification in terms of pressure and volume of the containment

• The additional protection the guard vessel provides, appears in some conditions to allow construction of nuclear plants closer to centers of population

For marine reactors, a water filled containment giving pressure suppression and elimination of the additional weight of a shield has been adopted for some designs.

The requirements for the CEDM are fine position control, lubrication by primary coolant, easy access to electrical parts, and high seismic resistivity. Soluble boron free reactivity control and load following operation over its full power range require a fine positioning capability of the control rod. The investigation of the failure frequency of operating CEDMs has revealed that the major source of trouble was in the electrical components requiring an easy access for maintenance. The long protruded part for the extension shaft stroke outside the reactor pressure vessel leads to high level of seismic excitation and reduces the margin to design stress limit. The current magnetic jack type CEDM used in the Korean Standard PWR is considered to be inadequate to meet the fine control requirements because it has only a step wise positioning. In addition the adoption of a self-pressurizer which occupies the upper plenum of the reactor vessel introduces difficulties in lubricating the moving part with the primary coolant since the latch mechanism would be located in the steam-gas region of the self pressurizer. Therefore, new concept of CEDM was proposed. The proposed design consists of position encoder, brushless DC servo motor, lift magnet coil, rare earth permanent magnet rotor, driving tube and split ball nut assembly. The rotor, driving tube and ball nut assembly are all connected into a single piece and lodged within the pressure housing which forms the pressure boundary. The encoder, DC servo motor, and lift magnetic coil are installed outside the pressure housing for easy maintenance. The use of a brushless DC servo motor with rare earth permanent magnet rotor allows a maintenance free operation of the motor. The fine control capability of the

CEDM is assured by the use of ball nut — lead screw mechanism and its lubrication with primary coolant is provided by placing this part below the water steam interface surface in the pressurizer. The ball nut assembly is of three pieces split type. Lift magnet located below the DC servo motor engages the ball nut to the lead screw by lifting the driving tube and the rotation of the rotor induces linear motion of the control element assembly up and down. The lead screw is part of extension shaft at the bottom of which a control element assembly is attached. When the scram signal is issued, the current supply to the lift magnet coil is cut off and the split ball nut releases the lead screw while dropping down by gravity and spring forces.

The large amount of cold borated water in the PCRV may be used for condensation of the vapor during accidents with primary coolant boiling due to overheating or depressurization by leakage. For this purpose, the following design modifications were performed:

-the riser was locked by a cover;

-the outlets of by-pass pipes were located several meters below the level of cold borated water;

-the cross-section of by-pass pipes was increased up to 1 m2.

If depressurization or overheating the coolant occurs, these innovations cause the steam-water mixture to pass through the layer of cold borated water.

There are three types of probable leakage:

— destruction of the boron control pipe (050mm) outside the reactor vessel;

— gas pressurizer leakage due to destruction of PCRV cover design or double

(inside and outside the vessel) destruction of steam generator section leg pipe (0200mm);

— double (inside and outside the vessel) destruction of the cooling system

pipe (050mm) under the cold borated water level.

Therefore, large primary leakage is a leakage from the gas pressurizer only. If this leakage occurs, the reactor design will keep the coolant inside PCRV. The steam-water mixture will be squeezed through by-pass pipes into the volume of cold borated water and condensed there.

In the case of very large leakage, when the depressurization speed is very high, there is a probability of squeezing the steam-water mixture through the density lock too (according to a hydraulic resistance of the each channel). During this process full reverse flow in the core is possible. It may cause overheat of the cladding. However an extended cross-section of by­pass pipes allow to pass all the bulk of boiling coolant during several seconds.

Large primary leakage. The scenario of this accident is the following: — depressurizing of the cold borated water and degasing;

-propagation of the depressurization wave into the primary coolant via the by-pass pipes, penetrations in the riser cover (from the top) and the density lock (from the bottom);

-boiling of the primary coolant;

-raising of the primary coolant level up to the by-pass pipes’ position due to coolant density reduction;

-probable stopping of the primary circulation;

-squeezing the steam-water mixture through the by-pass pipes and density lock (according to a hydraulic resistance of the each channel) into the volume of cold borated water and it’s condensation;

-reactor shut-down due to negative fuel and coolant temperature reactivity feedback;

-restoration of the primary circulation;

-compensation of coolant losses by cold borated water;

-pressure reduction down to atmospheric;

-reactor’s heat removal by steam generators.

Small primary leakage. The most probable reason for small primary leakage is a external destruction of the boron control pipe (050mm), which opens into the primary coolant. There is slow depressurization of coolant during this process. The primary coolant level in the riser comes up to the by-pass pipes location, going into the cold borated water volume and condensing there. This process accelerates the depressurization and decreases the coolant outflow. Coolant losses are compensating by cold borated water flow through the density lock. It results in shutdown of the reactor. When the coolant pressure decreases to atmoshperic, the coolant outflow will be stopped. The residual amount of water is enough for the heat removal process.

The probability of leakage via pipes of the borated water cooling system is very negligible: it can be caused only by simultaneous breakage of the pipe inside- and outside the vessel. Breakage in a single place will be detected: there is a checking of pressure in the cooling system. Large losses of cold borated water are excluded by the design: the borated water coolers are located in the upper part of the volume. Therefore, loss of borated water will be limited by the volume above the damage position. When the level of borated water reduces below the rupture location, outflow of borated water will be stopped. The accident will be continued with the same scenario, described for the large leakage. The rate of depressurization will be slower. There is no stoppage of the primary circulation during this accident.

It is apparent that the containment and safety systems may be considerably simplified due to innovation in question.

Decomissioning. Another IRIS feature is a thick (about 3m) borated water layer between the core and PCRV — walls. It allows decrease in PCRY activation significantly. It makes possible decrease of the residual PCRV radio-activity up to an environmentally acceptable level and to simplify reactor decomissioning.

Radioactive waste storage. The large scale PCRV may be used as a long-time storage facility for bumt-up fuel assemblies. Hence, there is no radioactive waste with high activity outside the reactor vessel.

Siting near population sentres. The high level of safety (core melt probability about 108 per reactor*year for PIUS) makes possible siting of this reactor near population centres.

A beyond the design basis loss-of-coolant accident (LOCA) for the floating nuclear power plant (FNPP) with two water-cooled integral ABV reactors (Fig. 1) has been considered. Primary coolant natural circulation is ensured in the ABV reactor. The multi-sectional once-through steam generator is built inside the reactor pressure vessel. A noncondensible driven pressurizer is connected to the reactor by a pipeline. The design characteristics of the ABV reactor plant are presented in Table 1.

The emergency core cooling system of the FNPP includes two active (high and low pressure) subsystems designed to make up the primary coolant losses even at the guillotine break of any primary pipeline. Besides the ECCS, the active residual heat removal (RHR) system with emergency feedwater pumps (EFP) or the passive RHR system with pressurized accumulators can be used to provide emergency heat removal through the steam generator.

All primary system pipelines, including the pressurizer surge line, have 15 mm dia. orifices to reduce the leak flow rate during a LOCA.

The FNPP accident localization system includes the following leak-tight compartments (Fig. 2):

• Reactor containment shell.

• Pressure suppression pool (PSP) compartment.

• Auxiliary equipment compartment (not shown on Fig. 2).

Safety membranes are installed:

(1) in the connection between the containment shell and the pressure suppression pool compartment,

(2) in the connection between PSP and auxiliary equipment compartments (not shown on Fig.2), so that all plant compartments are separated at normal operation. The rupture of the first safety membrane occurs when the containment shell absolute pressure exceeds 0.2 MPa. After that, the mixture of water, steam, and noncondensible gases starts to flow through the gap around the metal-water shielding tank (Fig. 2) to the pressure suppression pool. In this way, a major portion of the steam mixture condenses on the outer surface of shielding tank and the remaining steam condenses in the pressure suppression pool. Cooled noncondensible gases accumulate in the free volume of the PSP compartment. The rupture of the second safety membrane occurs when the absolute pressure inside this compartment exceeds 0.3 MPa.

The above-mentioned metal-water shielding tank surrounds the reactor and contains around 26 tons of water. Under normal operation this water is cooled by special heat exchangers.

The following specific features of the ABV reactor plant design can be highlighted as the factors of high importance to the LOCA progression:

— elimination of large break LOCA thanks to exclusion of large-diameter primary pipelines;

— employment of the outflow restrictors to decrease the primary coolant loss during a LOCA;

— upper location of the penetrations of the primary pipelines through the reactor pressure vessel;

— large water inventory above the core;

— favourable conditions for the emergency heat removal through the steam generator thanks to the primary system integral design;

— nitrogen presence in the primary system, including nitrogen dissolved in the primary coolant during reactor plant operation in transient regimes. The analysis of this factor is one of the main objectives of the study.

Accident Scenario

The postulated LOCA scenario is based on the following assumptions and events:

1. The accident is initiated by the double-ended break of the pressurizer surge line near the reactor pressure vessel.

2. Both trains of the ECCS fail to operate.

3. Additional failure of the shielding tank cooling system is supposed.

4. Full power operation of the reactor plant is assumed before the start of the accident.

5. Reactor scram occurs when the reactor pressure reduces to 12.5 MPa.

6. The EFP starts to operate after reactor shutdown and provides around 15 m3/h of feedwater at a temperature of 40 C.

7. The total amount of nitrogen dissolved in the primary water inside the reactor vessel is assumed to be 9.4 kg. The assessment of initial nitrogen concentration in the reactor coolant was based on the Henry law.

8. No operator action is performed to mitigate the accident.

Reactor plants of ATETS-200 type are a group (ATETS-80, ATETS-150, ATETS-200) of plants of the same kind, developed on the base of an integral reactor with natural coolant circulation, they are autonomous sources of electric energy and heat.

Compactness of the integral reactor, simplicity of the primary circuit and use of a highly efficient steam generator allows natural circulation in all conditions and excludes the use of pumps when providing electric power of up to 200 MW. The ATETS-200 reactor vessel dimensions are not greater, than the dimensions of AST-500 reactor vessel, mastered by the industry.

The investigations confirm the possibility of increasing power up to 250-280 MW.

The plant is notable for the wide variety of passive channels for residual heat removal:

— to each of two heat exchange loops (Fig. l) a channel is connected, which provides

ATETS-200 reactor plant flow diagram

Fig-1

ERHR channel
on the Reactor Upper head

Water inventory in ERHR tank — 150 m3

Cooldown diration (grace period) — 24 hrs

The channel self actuates at the reactor pressure increase up to 21 ‘H MPa

Fig. 2

residual heat removal through SG with natural circulation with heat removal to water tanks, from where water is evaporated to atmosphere;

— independent passive channel of heat removal (Fig.2) is located on the reactor. With its help primary circuit heat is transferred through the wall of the condenser-heat exchanger by natural circulation to a water tank and then it is removed to atmosphere.

Self-actuation of ERHRS channels with emergency protection actuation and, if necessary, of actuations of location system with use of self-actuation devices is provided.

At the upper part of RPV there is a space filled with the mixture of N2 and steam providing a surface in the primary circuit where liquid and vapor can be maintained in equilibrium under saturated condition for pressure control purpose during steady state operations and during transients The pressure of the pnrnarv svstem is equal to the N2 partial pressure plus the saturated steam pressure corresponding to the core outlet temperature A hydraulic instability way take place in some conditions m steam-water two phase flow under lower pressure and in natural circulating loop A N7 partial pressure of 0 6Mpa is chosen to maintain sub­cooling in the core outlet with large margin in order to avoid two-phase flow even m the hot channel dunng normal operation and transient The volume of upper space is large enough to prevent the safety valve from opening during most of the transients, such as loss of off-site power In the present design the safety valve is only opened in case of ATWS mduced by loss of mam heat sink The operational experiences gamed from NHR-5 shows that the N2 as a blanket gas does not cause a problem with water chemistry The concentrate of nitrate and nitnte m the coolant is around 5 ppb

For nuclear ships, it is essential from the economical point of view to shorten the time of the maintenance and refueling works. In operation of nuclear ships, the maintenance and refueling facilities as supporting systems are very important to make these works shortly, simply and safety. From this standpoint, the design study of a one-piece removal method is being carried out. This method is that the CV with its internals is removed for the maintenance and refueling, and is replaced to another one whose maintenance

and refueling have already been completed. Figure 4 shows the concept of the one-piece removal of the CV with its internals. It is thought that this method is promising because the integral type marine reactor is relatively small and light. The merits of this system are: (a) To shorten the days required for the maintenance and refueling. Ships are required to stay in dock about 3 weeks for these works, (b) To carry out the maintenance and refueling in a large space of land facility with highly safe, (c) To reuse the reactor after the ship’s life, (d) To make the decommissioning of the ship easy.