F. M. MITENKOV, V. M. RULEV OKB Mechanical Engineering,

Nizhny Novgorod,

Russian Federation

Abstract

A once-through cassette steam generator for an integral reactor is considered. The steam generator consists of separate cassettes and is arranged in the annulus between the reactor vessel and the core barrel. Specific developments of integral reactor designs show that the technical feasibility of their implementation is determined to a substantial degree by the characteristics of steam generator used. This paper describes a steam generator with improved overall and thermotechnical characteristics compared to existing steam generators. The high specific characteristics of steam generator allow development of an integral reactor for 600 MWe NPP.

1. INTRODUCTION

Integral reactors characteristics depend greatelv on the characteristics and structure of the steam generator.

Power rise of integral reactors is limited due to the low specific power of the existing steam generators, in particular those made of helical tubes.

The necessity to locate steam generator inside the reactor vessel causes some additional requirements on its structure, the most important of which are high compactness of the heat exchange surface and high specific power of the steam generator.

OKBM has developed various types of once-through SGs being operated in marine reactor plants PWRs. These SG types, in particular, include:

1) SG made of helical tubes;

2) Cassette-type SG made of straight tube steam-generating elements.

It has been proven by numerous experiments that the specific design of cassette — type SGs with straight steam-generating elements has better specific characteristics compared to SGs with helical tubes other conditions being equal.

For example in the operated cassette steam generators, developed by OKBM specific power in the effective heat-exchange zone reaches 60 MW/m3.

The design of cassette steam generator for VPBER-600 reactor is considered in this report.

The transport — from the manufacturing plant to the construction site is most probably the main limiting aspect of pressure vessels of integral reactors. Though there were several pressure vessels the manufacture of which was completed on the site, the present interest both of plant owners and manufacturers is to have shop-fabricated pressure vessels.

There are three possible ways of transporting such a large and heavy component: railway, road, ship or a combination of them.

Railway transport Advantages =

— accessibi1 іty: both manufacturing plants and construction sites are usually served by railway.

Disadvantages:

— limitations in the diameter (transporting profile) and in the weight.

— a special railway truck is needed

Advantages:

— accessibi1 іty■ both manufacturing plants and construction sites have an access by road.

— variability^ there is a variety in choosing a most convenient route of transport.

D і sadvantages =

— limitations in diameter (not so strict as at railways: there

is more possibilities in the selection of the route),

— a special truck is needed

Water transport Advantages =

— no limitations by dimensions and weight

— low price

D і sadvantages =

— access ibi 1 і ty-‘ navigable water way is needed both at the manufacturing plant and on the site

Combined transport:

For the final part of the route, both-. railway and water transport are combined with the road one for the transport from the port or railway station to the reactor building.

A combination: railway — ship — road — sship was used at the

transport of the VVER-1000 vessel from Plzen to Belene in Bulgaria.

There are certain safety features and advantages in integral reactors related to decay heat removal due to their power range and compactness. They are, especially, beneficial if they are implemented with passive initiation as well as passive operation.

The benefits of these systems are

• Extension of the time available for operator action

• Possible reduction in redundancy requirements

• Use of intermediate circuit eases maintenance problems in the steam generators and turbine due to the absence of radioactivity

• Cost reduction due to a reduction in the number of system components

An intermediate circuit is necessary for heating applications to eliminate the possibility of radioactive leakage to the end user This protection is particularly effective if the intermediate pressure is higher than the primary pressure

Since a water/water heat exchanger can be more efficient than a steam generator, in terms of specific transfer capacity and m certain design conditions, an intermediate circuit can allow greater power in a given size of RPV or a reduction of operating pressure for the same power The choice is an optimization between space available and the relative costs of a larger internal steam generator compared with an intermediate circuit

Passive system initiators used in integral reactors are

• Rupture discs

• Non return or check valves

• Valves which operate on changes of pressure differential

• Systems constantly in operation in normal as well as accident conditions

The integrated primary system concept eliminates all large primary circuit pipe work, thus intrinsically eliminates large loss of coolant accidents. The largest pipe break in the primary circuit is the break of the connection pipe supplying chemical and volume control system(CVCS). To prevent siphoning off the reactor wrater inventory in the hypothetical event of a CVCS line break, open connections are made between the steam generators and pressurizer. Thus there is no possibility of rapid emptying of the reactor vessel requiring massive and early injection of ECCS water. Since reactor vessel is flooded all the time by the water in the safe guard vessel, there is no need for external emergency core make up. The safe guard vessel is sized to provide a minimum of 72 hours’ heat removal without operator intervention.

2.3.3 Containment Overpressure Protection System

Since the maximum pipe break is small due to the integral nature, the pressurization rate of containment is slow. Energy released to the containment through the break point is removed using the steam injector driven containment spray system to prevent exceeding the containment design pressure. The steam injector is a simple, compact passive pump that is driven by supersonic steam jet condensation. The steam injector can operate even by atmospheric pressure steam. The steam from break point is supplied to the steam injector. The steam injector pumps up the water from a water storage tank on the ground to the spray nozzle located at the top of the containment.

V-500 SKID is an integral PWR in which the core and steam steam generators (SG) are contained within the steel pressure vessel, Fig.2. The

pressurizer is located apart from the pressure vessel. The guard tube block shroud is of 2.8 m in diameter and it separates riser and downcomer parts of the coolant circulation path. The hot coolant moves from the core through the riser and upper shroud windows into the steam generators located in the downcomer. The coolant moves due to the difference in coolant densities in the downcomer and riser. The pressurizer is connected, by two pipelines, to the reactor pressure vessel and the water clean up system. Injection of cold water into the pressurizer improves the pressurizer parameters.

The experimental facility is presented schematically in Fig. 1. Its basic elementa are the next:

• heating element: spiral or with internal heater 1:

• assembly body made from the transparent quartz-glass tube for assembly dre-out process visualization 2;

• bars 3;

• flange 4

• thermocouples J;

• refrigerator 6;

• calibrated measuring glass for hydrogen accumulation 7;

• level gauge (volume) in 8:

• expansion tank for collection of water replaced by steam and hydrogen 9;

• water-provided tank 10.

The collection and treatment of the information is fullfilled by the computer — controlled system. ‘

During the process of the model dry-out a steam-water mixture is directed to the refrigerator where it is condenced, while hydrogen is led to the accumulator. Hydrogen replaces water to the calibrated replacing tank. The pressure difference gauge registers the change of water level in replacing tank vs. time what enables to identify the moment of the steam-zirconium reacton beginning and to measure the amount og the hydrogen produced. The thermocouples are used for measuring the temperatures of the simulator surface and that of steam. For the simulator to be vacuumed and and after that filled with argon serves a special system.

EXPERIMENTS

In the experiments carried out two types of heater were used: one of spiral type and another — as a rod with internal heating.

The spiral type heater was fabricated of stainless steel capillary tube 4 mm in diameter. Inside this heater a small 60 x 5 mm plate zirconium was arranged. The chromel-alumel thermocouple was aranged at the plate centre.

The rod (HPE simulator) was made of the zirconium tube 09.3 x 0.5 mm. It was heated by molybdenum-made electric heater arranged inside which was insulated from the rod wall by pressed magnesium oxide. The total length of the simulator was 550 mm and that of the heated zone was 300 mm. The AC voltage was supplied to the simulator shell ahd heater from the transfomer. The internal part of the simulator was filled with argon.

The technique of the experiments carrying out was as follows. After the facility being filled with distilled deairated water one turned on simultaneously the simulator heating and the system recording in time the temperatures of the simulator surfase and exitingd steam and water level in accumulation tank. Once the electric powder was turned on, the simulator surface reached the saturation temperature and the boiling took place. The beginning of boiling was accompanied by intensive dry-out of the model and consecutive increasing of the rod surface temperature. The front of temperature increase was spread down the rod following the water level in assembly. The experiment was finished when the volume gauge signal reached its extreme value.

The scope and contents of the procedures for maintenance of integral reactors, developed in OKBM meet the requirements of national regulatory documentation. Thereby the following peculiarities of the integral reactor are taken into account:

— presence of the guard vessel; .

— location of steam-generators (heat exchangers) in the reactor vessel.

A complex of special devices for scheduled servicing and, if necessary, for repair and reconditioning work which account for the peculiarities of integral reactor lay-out has been developed and tested in AST-500 reactor conditions.

As for inspection of metal and welded joints the following measures are provided in the design:

— periodic visual inspection with video recording of the part of the reactor vessel visible in the zones between heat exchangers with the help of a periscope and of the whole surface when the heat exchangers (SG) are removed;

— periodic eddy current and ultrasonic inspection of the welding and main metal of the reactor vessel in the core zone:

— periodic outside visual and ultrasonic inspction of the reactor vessel with the help of a rotational device and a universal self-propelled device;

— periodic radiographic inspection of the welds of nozzles and penetrations in the upper part of the reactor;

— periodic inspection of the main metal and welds of the reactor vessel using test sample;

— periodic visual inspection of the state of in-vessel devices on removal from the reactor.

The strength and leak-tightness of the structures is confirmed by:

— periodic hydraulic tests of the reactor and heat exchangers of the primary and secondary circuits (steam generators) ;

— periodic pneumatic tests of the guard vessel.

In RP power operation, constant control monitoring of the reactor and guard vessel leak-tightness is provided by measuring the GV environment parameters (pressure, activity, gas content), also acoustic-emission methods of inspection are used.

Constant monitoring of primary-secondary circuit heat exchanger (SG) leak — tightness at RP power operation is performed by measuring the activity and gas content of the secondary circuit medium.

In the event of heat exchanger (SG) loss of tightness the leaking section is looked for, the leaking part is plugged or (if necessary) the whole section is substituted with the help of special devices.

Analysis and discussion of the decisions made by operating staff experts have shown their acceptability during operation.

Arrangements for monitoring the AST-500 reactor vessel

When the problems of scheduled servicing and repair are considered, the problem of the cleating to operate the plant before the end of its lifetime should also be discussed.

One of the advantages of the integral reactor is that it is simple from the point of view of technology and radiation to remove it from operation. The presence of a thick water layer between the core and reactor provides low radiation levels from the structures. It allows performance of dismantling work in the reactor cavity using standard equipment without using special means of protection and unique mechanical arms, very soon after reactor shut down and core unloading. The main pan of the equipment is low radioactive or even not radioactive and may be dismantled in the same way as at industrial plants. The mass of in-reactor equipment with high radioactivity, which is dismantled by standard means is 2% of the mass of the reactor unit.

In conclusion it should be mentioned, that the results of OKBM work on the reactor plants NDHP, NCP, VPBER show, that an integral lay-out of the reactor gives additional, new possibilities for NPP increase of safety in comparison with loop-type plants. The difficulties of operational servicing, caused by compactness are overcome when highly reliable in-reactor equipment is used. .

In September 1989 the Indonesian government decided to perform a new the NPP’s feasibility study including a comprehensive investigation of the Muria site. The study itself should be carried out by the National Atomic Energy Agency / BATAN under the directives of the Energy Committee (PTE) of the Department of Mines and Energy, and with the cooperation of other institution such as State Electricity Enterprise (PLN).

On August 23, 1991, an agreement was signed in Jakarta between the Indonesian Ministry of Finance and BATAN on behalf of Indonesian government, and the consultancy company NEWJEC Inc. (Japan). This agreement contracts NEWJEC Inc. for a four and a half year period to perform a site selection and evaluation, as well as a comprehensive NPP’s Feasibility Study.

The scope of the feasibility study includes two main components :

• The non-site studies, covering energy demand and supply, energy economy and financing, technology and safety aspects, the fuel cycle and waste management, and general management aspects, among other things.

• Site and environmental studies, covering field investigation and assessment of candidate sites, site qualification and evaluation, and environmental, socio­economic and socio-cultural impacts.

On December 30, 1993, the Feasibility Study Report (FSR), also called Non-site Study Report (FSR), was submitted. A final report including a site and environmental report, and a preliminary safety analysis report will be provided at the end of the four and a half year contract. These documents will provide the information necessary for site permit application, for the design engineering basis and other project and industrial infrastructure preparations.

Safety aspects are of utmost concern in the FSR, which cover and assess not only the proven designs available in the market at present, but also advanced systems expected to enter the market in the near future.

The main results of the FSR are among others Primary Energy Supply Scenario shown on Table 1.

Coal fired plants will dominate the electricity generation system. Nuclear Power Plants will increase in accordance with the demand. The data from this scope of work are used for optimization studies in the development of the Java-Bali electric system. The optimization shows that the first NPP’s operation is feasible in the year 2004. In the year 2019 the share of NPPs will give a contribution of 10% of the electricity supply, am amount equal to about 12,600 MW.

The concept suggested consists in subdividing the primary circuit of a reactor in small volumes with minimal hydraulic connections between each other. This concept is realized by employing devices of special construction, the so-called micromodules (MM). A micromodular channel is a tubular pressure vessel (Fig. l) 100-200 mm in diameter, where a fuel assembly is positioned in its lower part and a heat exchanger (HE) (1-2 circuits) in the upper part. Using flow splitters, the natural circulation of coolant is maintained in the primary circuit of the MM. Each micromodule is provided with inlet and outlet pipelines for secondary circuit coolant and with a pipeline connecting the MM primary circuit with the pressurizer (P). Thus, from the viewpoint of the construction and thermal-hydraulic scheme, the MM is a miniature integral reactor. To provide a prescribed reactor power, the core is composed of the necessary number of micromodules which are inserted in the moderator (e. g., graphite, heavy water, etc.).

Qualitative analysis and quantitative estimations obtained in late 1970’s and early 1980’s revealed that based on the above mentioned concept, the creation of a reactor of enhanced safety is possible at the expense of the following factors:

— The subdivision of the primary circuit in a number of small independent volumes (up to 100 1), resulting in reduction of the scales and accident consequences due to the

depressurization in the primary circuit. In the event of the MM vessel rupture, the heat removal from fuel assemblies is accomblished due to water inflow from the pressurizer and then due to water supplied by feed pumps.

— The absence of circulation lines of large cross-sections in the primary circuit: in the case of the rupture of the small diameter (about 10 mm) pipeline connecting the MM with the pressurizer, the reliable cooling of fuel assemblies will be provided during a long-period of time at the cost of possible steam condensation in the-micromodule heat exchanger.

— The limited velocity and amount of coolant discharge while disrupting the primary circuit, makes the problem of the accident loca. lization within a plant easier.

— The compactness of the MM primar. y circuit creates ideal conditions for the natural circulation (NC) in the circuit in each flow regime.

— In the case of channel reactors, it is easier to solve the problem of organising independent (autonomic) cooling system for control rods in the MM-based reactors. This system can operate as an additional channel for heat removal in case of failure of major heat removal systems.

— In view of the fact that the MM power and correspondingly, its cross-sectional dimentions are dozens and hundreds times lower than the reactor power, there exists a real potentiality to conduct experiments on different operation conditions, accident conditions included, using electrically heated modules of the full scale hight.

The above given considerations stimulated the development of a reactor based on the proposed concept.

2.1 Pressure vessel

The pressure vessel, without head, steam generators and internal components, weights 100 tons.

The internal arrangement of the vessel is dominated by the in-vessel disposition of steam generators. Hydraulic control rod drives are also included inside the vessel. For refueling, control rod drives must be dismantled. Steam generators can be replaced during a longer than normal maintenance outage.

The reactor is self pressurized, so no prressurizer is needed. A steam volume, in the upper part of the vessel, accommodates volume changes during operation.

The following "in-vessel" variables must be monitored during operation:

— Water level

— Pressure

— Core inlet and outlet temperatures

— Neutron flux

— Control rod position

— Coolant flow

In a natural convection reactor, coolant flow measurement is not straightforward. Different methods are being evaluated; correlation of flow with pressure differences across steam generators is the most promising one.

The first four variables can be measured without difficulties specific to this type of reactor. Control rod drive position is more difficult to measure as a consequence of drive design. Ultrasonic methods are being studied for solving this problem.

Control drive Control dr. ve rod structure Vote/- level Steer» generotor SG water inlet SG steam outlet Absorbing element Fuel element Core

Core support structure