2.1 Concept of reactor control

Two characteristics of CAREM reactor core design will be discussed:

— strong negative temperature coefficient

— no soluble boron for bumup compensation

The stron negative temperature coefficient enhances the self-controlling features of the PWR: the reactor is practically self-controlled and need for control rod movement is minimal. In order to keep a strong negative temperature coefficient during the

/1/ Chapter 6 oflAEA’s TECDOC on "Status of Small and Medium Reactors" In print.

whole operational cycle, it is necessary to do without soluble boron for bumup compensation. Burnup reactivity compensation is obtained with burnable poisons, i. e. gadolinium oxide dispersed in the uranium oxide fuel. Nonetheless, soluble boron is used to compensate cold-hot reactivity difference (the strong negative temperature coefficient means that a large difference in reactivity must be compensated between cold and hot states). Soluble boron is also used as a back-up of the safety shutdown rods: if the protection system of the plant orders to shut reactor down, and safety rods fail to do so, the reactor is shutdown by boron injection.

The effect the two core design features have on burnup, tend to cancel each other. A strong negative temperature coefficient means that fuel arrangement is not optimum from the reactivity point of view, but this charecteristic is compensated by the fact that power density in the core is lower than normal for PWRs. As a consequence, fuel temperature is lower, and the reactivity at hot state is higher.

The rated power of current integral reactor designs is limited to 700 MWe due to manufacturing limitations in the size of the RPV. The maximum diameter of the RPV is limited to 7 m, based on present technological capabilities. The steam generators have to be of a special compact design with a high power density to enable their location inside the RPV.

In-service inspection(ISI), maintenance and replacement of equipment and components is recognized to be more difficult as a result of the compactness of the

integral reactor designs and therefore, these aspects have received special attention in the design stage itself Some unique solutions and special tools have been developed for this purpose

The design characteristics are chosen to enhance nuclear safety and thus enable siting close to population centers. Decommissioning is also facilitated due to the availability of a fairly large RPV which can be used to store all the active components for a few decades in a safe maimer

New concepts of integral designs are being developed in the Republic of Korea and the Russian Federation In other countries, some modifications to existing designs have been undertaken Other design areas receiving active attention include safety — related heat removal systems for integral reactors, design of compact steam generators and decommissioning aspects

From operational point of view, integral reactor designs do not differ in principle from loop type reactors The principal advantages of integral reactors over current generation loop type reactors include the following:

• Enhanced safety level due to the location of primary coolant circuit within the RPV, in particular, a reduction in the probability of accidents accompanied by core damage

• Use of natural convection principle in the design of the primary coolant circuit not only provides a passive system for emergency decay heat removal, but also permits design of natural circulation reactors operating at rated output

• Reduction of neutron fluence on the reactor vessel to a negligible level enhances RPV life substantially

• Significant increase in shop-fabrication of the reactor systems reduces the volume of assembly work at site, and improves conditions for implementation of quality control procedures.

• Stringent requirements of leak tightness for the outer containment shell are relaxed as a closely fitted steel containment vessel called guard vessel functions as the first containment barrier A reinforced concrete shell would be adequate as an outer containment for protection from external effects Consequently, requirements of special safety systems for primary coolant inventory control and decay heat removal are also substantially simplified

• Potential reduction in construction time improves the economics of integral reactors

• Simplification of decommissioning work enables an earlier reuse of the site

Aspects currently receiving greater attention from designers of integral reactors include the following’

• Significant increase in overall dimensions and weight of the RPV resulting in the need to employ special handling and transport facilities during construction/assembly work

• Restriction of maximum reactor power capacity to 700 MWe as a consequence of limitation in allowable overall dimensions due to constraints in production capability of the industry

• Special design characteristics of the steam generators such as compactness and high power density which have a critical impact on the RPV dimensions

• Designs to facilitate comprehensive planned maintenance for trouble-free operation and replacement of major components.

• Use of nitrogen gas for pressurization

The CAREM Project is under the development of the Detail Engineering needed to start the construction A technical revision was made by the owner of the project (CNEA) in 1992 and a cost estimation following TRF 269 was finished in 1993

During the present year the main tasks were devoted to work on those subjects with lack of experience and were research and development were needed as RPV Containment, Internals, Thermal hydraulics of different Systems (RPV, Secondary, Primary, Containment etc) lay-out This in depth studies encourage the Project Personnel to study different solutions for the containment and RPV From the left to the nght in the next picture we have the previous design and the one in which studies are done at present The “old” design has a containment made totally with steel and the RPV with a conical shape The design was originally aimed to have a modular containment which could be transported in big pieces in order to reduce the on site works However, the solution suggested for the fixation of the containment was a difficult one from the point of view of civil engineers Because of the reduced containment size it was necessary to have the conical shape (small space for the RPV head) The latest design made from concrete m its lower part allow to have bigger available spaces so the RPV is of the straight type This solution increases the maintainability of the NPP, mainly the SG are more available for repairs, one of the weak points in the “old’ The more available space allow us to change the SG feeding pipes from a downwards type to a horizontal which permits an easier manufacturing solution Up to today the only drawback of the latest design is the corresponding to the drive mechanisms support which now is positioned in a bigger cylinder

In the following sections a short description of the activities in process except for those related to the SG in the CAREM Project are mentioned In relation with the SG, a Special Program is been conducted That Program has started with the construction of a mini SG to be tested in a high pressure loop Studies are underway to define the whole scope of data that should be obtained from a 1 1 model The studies include the qualification process and the search of a facility for testing the SG

2. CAPCN (Circuito de Alta Presion у Conveccion Natural): High Pressure Loop.

2.1. Description and Similarities with CAREM-25

The High Pressure Natural Convection Loop is part of the Thermal Hydraulics Laboratory Designed, Constructed and operated by INVAP for the CNEA Its purposes are to verify the thermal hydraulic engineering of CAREM NPP mainly on two subjects dynamic response and critical heat flux These are accomplished by the validation of the calculation procedures and codes for the rig working in states corresponding (by similarity criteria) to the operating states of CAREM reactor

CAPCN resembles CAREM in the primary loop, while the secondary loop is designed just to produce adequate boundary conditions for the heat exchanger Operational parameters are reproduced approximately for intensive magnitudes (Pressures, temperatures, void fractions heat flux etc) and scaled for extensive magnitudes (flow, heating power, size, etc) Height was kept in a I I scale

CAPCN was constructed according ASME for the following parameters I SO bar, 340°C for the primary and 60 bar 340°C for the secondary (pump exception T<240°C) The level difference between center points of core and steam generator is ~ 5 7 m Primary loop may operate in saturated regime (self — pressurized), or subcooled (dome pressure increased with nitrogen injection), with heating power up to 300 kW and different hydraulic resistance

The nuclear core is reproduced by electric heaters operated by a feedback control loop of dome pressure (primary self pressurized) For the dynamic tests the heaters are 1 2 meters long

The heaters bundle that will be used for the CHF tests differs from the one used for the dynamic ones, in order to have a configuration that allows to reach heat fluxes high enough to be sure to obtain departure from nucleate boiling for the complete range of pressures, flows and subcoolings. The heated length is 400 mm. The seven rods bundle has two of the rods which allow an overpower of 20% These rods have six thermocouples each in order to ensure the measurement of the exact location of CHF The rest of the rods only have two thermocouples that guarantee CHF detection if it takes place on One of these rods. AJ1 the thermocouples are located near the upper end of the heated zone because the axial power distribution is uniform in the CHF heaters bundle

The group of states foreseen for these tests include the following range of main parameters’

Natural circulation flow may be regulated by a valve m the cold leg and a by pass to the bottom of riser. A gamma densitometer is available for void fraction measurements The heat exchanger (SG) has two coils, once through, secondary inside For the CHF tests the steam generator is useful only as a cold source, so the secondary loop operating parameters are not relevant as long as they can be controlled

The secondary loop pressures and cold leg temperatures are controlled through feedback loops operating valves. The pump allows the regulation the flow The condenser is an air cooled type with flow control

Both loops allow automatic control and can be pressurized by nitrogen injection

The present configuration of CAPCN allows the study of a stationary state similar to the CAREM conditions of

pressure, specific flow and enthalpy

The combination of primary and secondary states is limited only by conditions attainable with the heat transfer capacity of the heat exchanger As a consequence this loop will permit to validate the calculation tools used on the CAREM Project (RETRAN and ESCAREM) in those conditions.

The inclusion of a SG (IS tubes) with a design similar to the CAREM will allow to have a full 1-D thermalhydraulics analogy, allowing the extrapolation of results directly to the CAREM-2S

The thermal-hydraulic design of CAREM reactor core has been performed using a version of 3-D, two fluid model THERM ІГ code In order to take into account the strong coupling of the thermal-hydraulics and neutronics of the core, THERMIT was linked with an improved version of CITATION code (developed in FNVAP, and called CITVAP) This coupled model allows to obtain a 3-D map of power and thermal-hydraulic parameters at any stage of the bumup cycle.

The thermal limits were calculated using the 1986 AECL-UO Critical Heat Flux Lookup Table, validated till now with all the available measurements in the operational range in order to ensure a 9S/9S reliability/confidence in the thermal limit. The CAPCN CHF test results will be used to improve these calculations by increasing the experimental data in the operating range and fuel element geometry of the CAREM core.

The two Pipe Ducts between Pressurizer and Reactor Vessel connect hydraulically the top and the bottom of the respective cold water plena in order to create a common cold water plenum.

The choice of two connection levels makes natural circulation possible in case of temperature difference between cold plena. If the normal decay heat removal route (i. e. the active steam/water system) is lost, the

uninsulated wall portion of the Pressurizer would thus help removing by conduction the decay heat towards the Plant Pool.

Conveyed water to and from each vessel, belonging to a common cold water plenum, does not significantly contribute to the thermal loadings on the pressure boundary during transients.

Air Coolers

Two finned-tube Air Coolers are arranged in loops in natural circulation.

Each Air Cooler is rated 1 MWth at 30 °С ambient air and 95 °С pool water inlet temperature.

The onset of natural circulation occurs every time the pool water temperature becomes higher than the ambient air temperature.

Operation of the air coolers would prevent, for unlimited time, the pool water from boiling, in case of long-term loss of the operational decay heat removal system.

The technology of the air coolers in natural circulation is derived from the design and operating experience of ANSALDO in the field of the LMFBRs.

one of the conditions of the reactor reliability and safety is the favourable water-chemical conditions for the construction materials. The traditional measures of water purification and the introduction of correcting agents in the MM reactor being the "blind" part of the loop are practically impossible.

The study of the water-chemical conditions was performed using MM models under in-pile conditions. All in all 5 micromodules were tested in the reactor of the First Nuclear Power Station. In total, the time of operation during testing amounted to about 150 000 days and nights. The duration of the operation of four micromodules amounted to about 10 years for each of them. The main results of experiments are as follows:

— while filling the gas space of the pressurizer with nitrogen, the ammonia synthesis takes place in the water of the primary circuit and the pH value is set up at a level of 9-11;

— the content of corrosion products in stainless steel is on the average — 0.15 mg/kg for iron, for nickel and chrome it was 20 mg/kg;

— the amount of chlorides does not exceed 0 05 mg/kg. The amount of gasses amounts to 30-150 H. cm5 /kg;

After 5 years of testing, one of the micromodules was withdrawn out of the reactor to assess the corrosion condition. The visual inspection showed that the fuel elements and the MM vessel are in a satisfactory state. No mechanical damages of fuel elements and spacing grids were detected and no visible depositions were present on the fuel element surfaces. After examination, this MM was placed again into the reactor, where it operated for about 5 years more.

In addition to the above mentioned studies, the work was performed on a number of directions of supporting the RKM-150 reactor design. They included comparative experiments on enhancing the absorbing materials control rods system. The best results were obtained with usinig Dl/2 Ті 05 which was adopted in the technical design of the reactor

Also, the problems of the fuel element cladding tightness control system were studied. It was found that such a control, which does not require sampling the coolant can be implemented using gamma-ray detectors being moved between the micromodules in the above-reactor space, when the reactor is shut down.

Conclusion

1. The concept is proposed to enhance the reactor safety at the cost of subdividing the primary circuit into small parts. This is achieved by means of micromodules incorporating a fuel assembly, a heat exchanger and a natural circulation of coolant.

2. Wide experimental and computational investigations confirmed the design characteristics of such a reactor (RKM-150) regarding:

The thermohydraulics involving the temperature regimes; flow rate and natural circulation stability; thermal core safety under the natural circulation; the water-chemical condition;

the manoeuvrability in var.-Ing the power during load variation; and the correspondence to the nuclear and radiative safety requirements.

Each weld and cladding have to undergo the stress relief annealing by a prescribed procedure: heating-up — dwell — cooling-down, so as to remove residual stresses in the material which attain up to the yield stress of the materia 1.Each section welded from two or more parts must be annealed: it must be kept in view that each annealing impairs material properties.

For the annealing. a furnace must be at disposal which enables placing in even a whole pressure vessel Dimensions of integral RPVs require very deep furnaces (more than 20m). (E. g. the annealing furnace at SKODA Reactor Shop has dimensions 7.5×7.5xl5ra which makes possible to anneal parts up to 13m long: pressure vessels VVER 440 and VVER 1000 are І1.8 and 10.8 long, respectively) In the last resort, a local annealing can be applіed.

Main machining operations are as follows­- turning of cylindrical surfaces of rings and welding edges,

— turning of spherical and/or eliptical surfaces of bottoms and covers including welding edges,

— boring holes and cutting threads in vessel flanges for flange bo1ts.

— boring holes in vessel covers for control-rod-drіve nozzles, instrumentation nozzles and flange bolts,

— machining of main nozzles and emergency-cooling nozzles including welding edges.

— machine or manual grindinq of outer and inner surfaces of vessels and covers, etc

A quality assurance programme must be followed during all phases of the manufacturing process a part of which is a programme of inspections and tests.

Main groups of Inspections and tests:

— material properties (chemical composition, structure,

mechanical properties, transition temperatuure curve etc.)

— geometry ( dimensions, shape, surface quality)

— integrity (surface integrity, volumetric integrity,

leak-tightness, strength).

Current techniques, apparatusses and devices which have been used at the manufacture of classic PVR and BVR pressure vessels are applicable for integral RPVs.

E. g.,« radiography of thick welds is performed by linear accelerators in special shielded cells. A heavy piece is transported to such a cell by a special truck with rotating positioners which have to have an appropriate loading capacity. At Skoda, the linear accelerator NEPTUN 10 MeV is used which is capable to test thicknesses up to 450 mm. A trolley rail-wagon with rotating positioners of 400 t loading capacity attends the testing cel 1.

5.1. Decay heat removal

All designs use the in-vessel steam generators or dedicated in-vessel heat exchangers for decay heat removal(DHR). Steam generators require valve movements to use them for this purpose. In all cases there is an external heat exchanger to transfer the heat to the atmosphere or to water tanks. This arrangement permits the use of a passive system based on natural circulation. Use of an independent heat exchanger in a dedicated circuit is claimed to give added reliability at the expense of extra cost. Since the primary coolant feed and bleed pipes as well as the CVCS pipes are small in diameter, the possibility of a large LOCA is eliminated. If the water level drops below the steam generator/heat exchanger level or drops to a level where natural circulation within the vessel is prevented, heat continues to be removed through steam/water being released through the break and cool water returned to the vessel by various systems or by make up from the inventory maintenance system. In the low pressure NHR, such a system is unnecessary and is not provided since heat can be removed by condensation on the surface of the heat exchanger.

The safety concept is taking advantage from the intrinsic safety characteristics of integral reactor and pursuing passive safety principles common to most small and medium reactors. The fundamental safety characteristics are:

— Low ratio of power density to heat capacity resulting in a slow rise of fuel element temperature under accident conditions;

— A substantially negative moderator temperature coefficient resulting from no soluble boron usage generates beneficial effects on self-stabilization and limitation of reactor power;

— Integral Reactor Vessel eliminates large primary coolant pipes and thus large break loss of coolant accidents;

— Large Passive pressurizer significantly reduces pressure increase for decreased heat removal events;

— Large volume of primary coolant provides more thermal inertia and makes plant more forgiving;

— No RCP seals eliminates potential for seal failures, a concern during station blackout;

— The use of passive safety systems leads to directly to simplification in design since it eliminates the need for multiple redundant safety systems with their redundant safety grade power supplies.

V. A. SILIN, A. M. ANTONOV.

I. V. Kurchatov Institute,

Moscow

A. M. AFROV, M. P. NIKITENKO, A. V. BUHTOYAROV Experimental Design Organization Hydropress,

Podlosk

Russian Federation Abstract

Pressure increase in the primary circuit to above the critical value gives the possibility of constucting the V-500 SKDI (500 MWe) lightwater integral reactor with natural circulation of the coolant in a vessel with a diameter less then 5 m. The proposed reactor has a high safety level, simple operability, its specific capital cost and fuel expenditure being lower than a conventional PWR. The development of the V-500 SKDI reactor is carried out taking into consideration verified technical decisions of current NPPs on the basis of Russian LWR technology.

1. Introduction

Nowadays together with improvement in the conventional NPP designs some organizations are developing reactors characterized by fewer potential accidents initiators and higher level of inherent safety. Integral PWRs may be considered as an example of such new generation reactors. These reactors can satisfy the safety principles formulated in [1], especially in the case of coolant natural circulation.

However their high specific capital cost may be the reason for their noncompetitiveness with other power plants. It is connected, first of all, with low specific power in such reactors, because of the difficulty of placing the steam generator (SG) large heat transfer surfaces within a steel pressure vessel and providing necessary coolant flow.

Under the constant heat transfer surface of SG and the secondary circuit parameters the amount of heat power of the reactor transferred from the primary circuit to the secondary one can be increased due to the temperature increase of the coolant in the primary circuit. However at subcritical water pressure the possibility of the departure from nuclear boiling hampers the increase of the reactor heat power.

The problems connected with the departure from nuclear boiling are eliminated at supercritical pressure as at supercritical pressure (SCP) the liquid is of one phase through all the range of temperatures. This property also simplifies the provision of the stable coolant circulation.

Increasing the temperature of the coolant at supercritical pressure and corresponding increasing of the temperature difference between the primary and secondary sides gives the possibility of increasing by several times the reactor power.

The SCDAP/RELAP5/MOD3.1 computer code was used to determine both the reactor and containment system behavior in the course of the accident. Several modules of the code were modified for this study to simulate the process of dissolved nitrogen release from the primary water to the vapor phase within the reactor vessel.

The computer simulation results showed that the reactor coolant leak rate through the pipe break decreases rapidly during the initial 30 seconds of the accident. Its value reduces from 16.5 kg per s at the accident beginning to about 3 kg per s, when the reactor end of the pressurizer surge line becomes uncovered and the leak flow changes from liquid to vapour. The subsequent discharge rate of steam follows the primary pressure (Fig. 3).

The heat removal from the primary system through the steam generator has an essential influence on the reactor coolant discharge rate and the rate of reactor pressure reduction. Heat transferred to the steam generator secondary exceeds the energy transferred to the coolant from the core for almost the whole duration of the accident, causing a decrease in the primary system pressure and temperature. However, the process of nitrogen release from the primary water (Fig. 4) leads to a reduction in the steam condensation rate at the primary side of the uncovered steam generator tubes. The break flow reduces at a lower rate as the result of this process. A significant part (about 20 %) of the released nitrogen mass remains in the vapor phase inside the reactor pressure vessel and localizes mostly above the water level inside the steam generator.

The containment shell pressure continues to rise even after the first safety membrane rupture at 16 s and reaches its maximum value just before the second membrane ruptures at approximately 2650 s (Fig. 5). The values of containment atmosphere pressure and temperature do not exceed permissible design limits of 0.6 MPa and 150 C during the accident progression. Their maximum simulated values are 0.34 MPa and 130 C, respectively.

A steady pressure decrease in all containment system compartments begins at ~5000 s of the accident. At the same time, a positive pressure difference between the containment shell and the PSP compartment starts to reduce to zero and becomes negative at -7000 s. This result is explained by decreasing discharge rate of the primary coolant and continuing heat removal from the containment atmosphere to the containment vessel walls and interior equipment, including the metal-water shielding tank.

During the next phase of the accident, a gradual displacement of water from the PSP compartment to the containment shell is observed as the result of increasing pressure difference between these compartments. At ~ 10000 s, the rising water level inside the containment shell covers the pressurizer surge line break located at the elevation of ~3.5 m above the shielding tank bottom level. After this event, the reactor begins to refill with suppression pool water (Fig. 6). The mass of coolant inside the reactor pressure vessel increases up to the value of -6000 kg and stabilizes beginning at -13000 s. During the last 5000 s of the transient, the calculated leak flow rate has a small value. The direction of the flow alters irregularly, so that the coolant mass within the reactor remains almost invariable for the rest of the accident simulation.

The results of the accident simulation indicate that a sufficient inventory of the primary coolant remains within reactor vessel to maintain reliable core cooling. The critical heat flux is not reached in the core during the accident progression.

The further state of the reactor plant will depend on assumptions concerning the degree of leak-tightness of the containment system compartments, RHR system working conditions, and operating staff efforts to activate failed safety systems. The available grace period is long enough to undertake measures for accident control.

Conclusions

In the course of performing the simulations, it was found that:

1. Neither core uncovering nor the critical heat flux in the core was reached without any operator action during the accident progression.

2. The process of nitrogen release from liquid to vapour phase within the reactor pressure vessel led to increased total loss of the primary coolant, but the values of containment shell atmosphere pressure and temperature did not exceed their permissible design limits.

3. Reactor refilling with the water of pressure suppression pool driven by the pressure difference between containment system compartments occurred during the last phase of the accident.