The approach for nuclear reactor technology and safety is developed in two steps. The first step is the next 50 years where the system will be dominated by the present proven system notably LWR’s and to a lesser extent PHWR’s with some evolutionary improvements in technology and safety of the system. The main reasons for this is : the present NPP’s are quite competitive with respect to coal plant, they are cleaner than fossil plants and they have good operational and safety performance.

The evolutionary improvements in the safety and technology must be performed to reach better performance, better economics, better protection for the investors and wider acceptance by the public.

The so-called "advanced reactor", such as ABWR or AP-600, offer improved safety systems features by simplification and introduction of passive safety features. In other point of view, the improvement of man-machine interface in the system is also a good point for these advanced reactors in order to resolve the complexity of the machine and make the operator’s life easier.

In longer term future, despite efforts in conservation and efficiency programs in energy production and use, the tendency of energy demand will keep increasing well into the long term future. This is due to increasing of living standard and the population growth especially in developing countries.

Nuclear energy can have a very important role in energy supply system, if it succeeds in its development and rises to challenges it is facing. If not, the role of nuclear energy will be replaced by coal or gaz.

The challenges that should be resolved are :

• possible weapons proliferation and diversion of nuclear materials

• sabotage against nuclear facilities

• safety concerns

• radioactive wastes and environmental issues

• standardization of future plants

From the challenges mentioned above, the possible requirements for next generation of NPP’s would be in essence based on revolutionary concepts which are among others :

• design NPP’s that provides inherent resistance to sabotage and man-induced events through physical, security, and institutional design and arrangement

• concerns over safety should be eliminated through non catastrophe reactor and fuel design (no core melt accident)

• standardization of future NPP’s is of great development in order to reduce and keep the development cost at reasonable level, on the other hand this will improve the economy of the NPP’s

In this context, the integral reactor concepts provide a remarkable inherent safety features.

In the 1980’s, the Research and Development I nstitute of Power Engineering (RDIPE) together with the Institute of Physics and Power Engineering (IPPE) has developed a number of reactor designs (with a capacity from 20 to 300 MW) for the Nuclear District Heat Plant. Fig.2 shows the schematic diagram of a reactor, and Table 1 incorporates the main parameters of the RKM-150 unit, for which the technical design and extensive calculational and experimental investigations were performed. Alongside with the concept consisting in sectioning the primary circuit into small volumes, the MM — based reactor safety is guaranteed by a set of other factors:

— Self-protecting features of the reactor: negative power, temperature and steam reactivity coefficient: the stability of the power generation field; — high heat capacity of reactor; — natural circulation of water in the primary circuit.

— The availability of safety systems: — reactor shutdown by introducing absorbers under the action of the force of gravity; — the micromodule replenishment with water during the vessel rupture from the pressurizer (passive) and then by pumps; the circulation of the secondary circuit with water from hydroaccumulators (passive) and then by pumps; heat removal from the MM by the evaporation of water from tanks (passive) via the water cooling tower (active) or to the network circulation loop (active).

— By additional technical solutions: moderate parameters (the heat fluxes of fuel elements, pressure, temperature); independent cooling system for absorbing rods; a good thermal contact between the MM vessel and graphite; pressure relation in circulating loops Pj >P2<P3 , etc.

The containment is of the pressure suppression type. This type of containment is appropriate for integrated reactors, because reactors of this type don’t need large containment volumes for accommodating primary circuit equipment. The height and volume of the containment are determined by the requirements posed by refueling operations and not by the need to accommodate operating equipment.

Containment wall acts as heat exchanger for removing decay heat to the ultimate heat sink, which is the atmosphere: after a

LOCA, pressure in the containment rises up, till it reaches design pressure, more than three days after the accident. At this pressure, heat transfer through the containment wall to the atmosphere is enough for preventing further pressure increases.

No specific manufacturing, transport or erection problems related to this component have been identified.

It is not necessary to access the containment during reactor operation: maintenance tasks can be carried out during refueling, as all equipment in the containment is static (no movable parts, apart of a few safety system valves)

3 1 Steam generator design

In steam generator design for integral reactors, the primary objective is to develop a compact steam generator to enable locating them inside the reactor pressure vessel, eliminating the possibility of a large loss of coolant accident (LOCA) and making efficient use of the space available This implies a high power density in the tube bundle

The main candidates for tube material are titanium alloys and Inconel, mainly 690 and incaloy 800 The choice would be related to national experience Titanium alloys have the following advantages

• Low coefficient of linear elongation, about forty percent lower than stainless steel

• Less sensitivity to thermal loads, as a consequence of a low module of elasticity

Inconel has the advantage of larger thermal conductivity Experience with its use has been mainly in the present generation loop type reactors

Incoloy 800 exhibits excellent properties for heat transfer and corrosion resistance and the allowable stress is larger than stainless steel Experience with Incoloy 800 as a tube material is quite large especially in France and Germany

It could be said that all three materials show good mechanical properties and have demonstrated good corrosion resistance as a tube material for steam generators The basic arrangement in integral reactors is to locate the SG within the RPV, in the annular region above the core level Consideration, however, has to be given to the distance between the lower part of the SG and the core, to prevent secondary water activation and radiation damage to the component

With regard to tube support, straight tube SG allows a simple support systems, using simple spacers, since flow is parallel to the tube, while m helical steam generators the flow crosses the tube Tube support in the latter case is more complicated Corrosion and build-up at tube supports with the secondary working fluid on the vessel side are more pronounced Primary coolant outside the tubes has the following advantages

• a high resistance to stress corrosion cracking,

• safety advantages in case of tube failure and

• a reduced risk of crud accumulation at the tube plate connection

In all cases, a provision for the possibility of in-service inspection of the complete bundle is recommended. The design must allow tube plugging, component removal and replacement. Steam generators that are not once-through type are more suitable for load following due to the water inventory but this option does not appear to be followed in any of the Member States.

Some reactor designs prefer secondary water boiling inside the tubes to reduce the reactivity effect in the core in case of a steam line break. Some designs place the secondary outside the tubes to reduce hydraulic losses, especially if the design uses natural circulation.

Hydraulic stability of parallel tubes is one of the most important design aspects to be considered. Experience shows that instabilities can be controlled through careful design. One of the features currently adopted is the introduction of orifices at the tube inlet to increase the secondary side pressure losses in the liquid phase. To avoid operational problems, the chemistry of the secondary coolant must be maintained at a high quality level for this type of SGs, to avoid crud deposition. Even in the case of some crud deposition, adequate experience exists regarding washing it away. RPV penetration for feed water and steam outlets can be optimized in number according to the needs of diameter limitations and to the specific design requirements.

7.1. Drive Mechanisms Instrumentation

The position of the piston in the Drive Mechanism gives the position of the neutron absorber in the core. The length of the piston is 300 mm and the length of the cylindrical cover is 1850 mm. The method used to measure the piston position is called Magnetically Variable Inductance.

The piston made from magnetic Stainless Steel is moving inside the cylinder covered with an electrical coil with high concentration of rings in one end and decreasing to the other. All other pieces of the Drive Mechanism are non-magnetic. A fine measurement of the inductance gives a measurement of the piston position and consequently of absorber position. The resolution measured was better than 1%.

The test were made at room temperature and non immersed in water and the inductance was measured. Interference and very low frequencies are a challenging situation and due to that, lab instruments of high complexity and cost were used. The design of a specifically designed instrument to measure inductance was finished and a prototype to operate at CAREM conditions will be tested in the “warm” and “hot” test mentioned in Section 6. of this work.

7.2. Reactor Protection System (RPS)

Description

The RPS is based in solid state intelligent processing units and hard-wired multiplexed voting and protective logic units. It has four redundant, independent channels, whose main features are:

• High reliability and availability as a result of design criteria and technology

• Fault-tolerance with on line auto-verification routines and auto-announcing capability

• Compactness and robustness

• High simplicity

Current developments

The current developments are originated from the Safety Requirement Specifications of the RPS.

Trip Unit:

The trip unit performs the data acquisition of the safety variables and compares them against the Safety System Settings to initiate the protective actions in case of anticipated operational occurrences or accident conditions. The development of the Trip Unit is subdivided in the following stages:

• Software Requirements Specifications & prototype

• Software design, code and implementation

• Hardware Requirements Specifications & prototype

• Hardware design and implementation

• Integration Requirements Specifications

• Integration Hardware/Software

• Validation

• Installation & Commissioning

• Operation & Maintenance

At present, the status of the development is at design stage in both hardware and software

Voting and Protective Logic Unit

The voting and protective logic unit performs the voting of

the redundant safety trip signals coming from the Trip Units in a logic arrangement of 2 out of 4 and then, according to the logic relations of the trip signals and initiation criteria, triggers the protective actions.

The development of the Trip Unit is subdivided in the following stages:

• Hardware Requirements Specifications & prototype

• Hardware Requirements Specifications

• Hardware design and implementation

Description

The highly automated digital Supervision & Control System, has an architecture of 5-level hierarchy with distributed processing and modem control technology It is conformed by different types of processing units

• Supervision Units (SU)

• Information Units (IU)

• Control Units (CU)

• Field Units (FU)

The Supervision & Control System is totally independent of the RPS High system reliability and availability are achieved by the use of redundancy and fault-tolerance in communications and processing unit

Operator interface is based on digital visual display units for safety, alarms, logics, processes and documentation presentation in the reactor main control room and others supervision and control centers Modem technologies as touch-screens, track-balls, custom keyboards, etc are used

Current Developments

The Supervision & Control System includes the development of a Control Operating System that implements all the low level functions as

Real-Time Data Base Communication System Control Functions Svstcm Management

This Control Operating System acts as a software platform on which the Supervision & Control System application is built on 1 he development process is divided in the following phases

Software Requirements Specifications & Prototype Software Design Specifications

Software Coding, Implementation & Integration Validation & Verification

Instalation <S_ Operation

The Ward Si Mcllor Methodology is applied in every phase of the development process At present, the status of the development is at Coding phase

8. Fuel Elements

The activities in this subject are being carried out by CNEA itself At present the detailed engineering for the CAREM 25 Fuel Elements and absorbers are under execution

Development of equipment for components and FE manufacturing.

The following tasks have already been earned out

• Development and construction of equipment for caps welding by TIG method

• Development and construction of dies for stamping and cutting elastic spacers components

• Development and construction of FE assembly and final control boards

• Construction of different manufactunng and metrological control devices for FE manufacturing

• Prototype of elastic spacer for the FE

• Dummy FE to define handling tools

Current developments

The following tests are under definition stages

• Elastic spacers mechanical and stress tests

• Fuel element seismic behaviour test

• Thermalhydrauhc behaviour in a low pressure loop

• Thermalhydrauhc behaviour in a high pressure loop

References

/1/ Dcnimns M. Метопа desenptiva del CAPCN, INVAP 0758 5302 2IASS 315 10 (1994)

/2/ Carnca, Pablo, Analisis de Analogies entre el reactor CAREM-25 у el CAPCN, INVAP, 0758-8700-2TAJN-003-10 (1994) /3/ Masnera. Nestor, Ensayos Dinamicos previstos para el CAPCN, INVAP 0758 8680 31AIN 012 1 0 (1994)

/4/ Camca, P & Balina J, Plan de Ensayos de FCC, Parte I, caractenz instalacion, INVAP 0758-8720-31 AIN-001-1A (1994)

/5/ Carrica P &. Balina J. Plan de Ensayos de FCC, Parte II, Ensayos Preliminares, INVAP 0758-8720-3IAIN-002-1AO(1994)

/6/ Carnca. P & Balina J, Plan de Ensayos de FCC, Parte III, Ensayos Finales, INVAP 0758-8720-31 AlN-003-10 (1994)

ПІ RA-8 Preliminary Safety Report,

/9/CONDOR 1 3, Villanno Eduardo INVAP (1995)

/10/ Strawbidgc and Barry Cnticality Calculations for Uniform Water Moderated Lattices, NSE, 23, 58 (1965)

/11/ Raslog. Muligroup Methods in Thermal Reactors Lattice Calculation, Lecture, Bogota, Colombia

/12/ Macdcr and Wydlcr, International Comparison Calculations for a BWR Lattice with Adjacent Gadolinium Pins, EIR-Bericht 532, NEACRP-L-271 (1984)

/13/ Arkuszcwsky, MCNP Analysis of the Nine-Cell LWR Gadolinium Benchmark, PSI-Beritch 13 (Aug 1988)

/14/ Szatmary, Experimental Investigation of the Physical Properties of WWER-Type Uranium Water Lattices, Final Report of TIC. Vol I, Akademiai Kiado, Budapest (1985)

At the beginning of the development of the ISIS concept (about seven years ago) it seemed reasonable to foreseen a moderate increase of the cost of the fossil fuels in the near future which would have improved the economic competitiveness of nuclear energy. Today, instead, two facts worsen this competitiveness:

— the fossil fuels price has remained low and stable,

— the efficiency of the modern electric energy generating conventional power plants is continuously increasing.

The importance of the second fact is such that it will drastically affect the energy market, in particular it will impact the market of nuclear energy.

In the past, the efficiency of electricity production of the nuclear power plants was similar to that of the conventional power plants. Under that condition it was profitable to generate electricity by the large-size nudear power plants that dominate the nudear panorama.

Today, the effidency of the modern Combined Cyde Turbo-Gas (OCTG) Power Plants has exceeded 50% and in the near future (before the year 2000) will reach and perhaps trespass 60%, while the one of the nudear water reactors stagnates at about 33%.

This new fact will have two main consequences:

— if the cost of fossil fuels remains stable, the cost of heat will remain substantially stable, but the cost of electricity will be reduced;

— the ratio electridty/heat production will increase up to the optimum dictated by the modern fossil-fired со-generative power plants.

Qualitatively it can be affirmed that today an effiaent use of energy favours the fossil fuels for electricity and of the nudear fuel for heat production, because of the lower electric effidency of the nudear power plants.

Quantitatively, a preliminary economic evaluation carried out comparing 60% effiaent CCTG, со-generative CCTG, conventional boilers and nudear power plants, has shown that nudear power plants could recover part of their lost economic attractiveness only if exploited as

со-generating or as thermal power plants.

The со-generative use appears attractive from 3000 hours/уг. and the thermal use from 5000 hours/yr. upwards (increase of the value of the plant in the order of more then 50 %)

The increase of the value of the plant is more important for со — generative reactors and can largely exceed 100% for specific site conditions where heat can be used during the most part of the year.

A trivial condition for any interest of a prospective utility in a со-generative nuclear plant is that an adequate со-generating reactor exists. For this the reactor designers, besides providing the reactor with convincing characteristics of radiological safety, have to overcome the unfavourable scale-effect of downsizing, because the thermal power needed is mostly in the order of hundreds of megawatts against the thousands of megawatts available from the today large nuclear reactor conceived for electricity generation only.

In the view of a reactor designer, the smaller reactor can be competitive, in spite of downsizing, provided that:

— the number of systems of the larger plants is strongly reduced,

— the specific mass of steel of the NSS it is not significantly increased (scale effect neutralised),

— the specific operation & maintenance costs become not excessive.

The со-generating version of the ISIS reactor is being designed to cope with these requirements.

The technical features illustrated in this article and the results of preliminary analyses for an use of ISIS as со-generating reactor can be summarised as follows:

— no active safety system is necessary to assure safety. All active safety systems can be eliminated.

— no adverse scale effect on the specific mass of steel of the ISIS NSSS with respect to the larger modern PWRs. This is possible also thanks to the milder operating conditions of a reactor designed for co-generation (e. g, lower operating pressure ).

Ongoing studies explore furthermore the possibility of reducing operating & maintenance costs, taking profit of the predicted simple operation of ISIS, of the reduced number of systems, of the modular approach that makes possible to share facilities, such as fuel handling and component handling equipment, between the reactor modules installed in the same reactor building.

3. CONCLUSION

The ISIS is an innovative Nuclear Power Plant under development in ANSALDO. It is based on original ideas derived by ANSALDO experience on proven LWR and LMR technologies.

The main features of ISIS are as follows:

— Outstanding passive safe behaviour of the Reactor, which means core shut down and cooling functions ensured in all accident conditions and no release of primary coolant outride the Reactor Building.

— Compact reactor layout, associated with modular fabrication and erection, made possible by the integrated design of the primary circuit.

— Reactor concept flexible for combined generation of heat and electricity made possible by the modular solution and the low sensitivity to downsizing.

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left BS. ANK

Number of independent subsections

Number of steam generating elements

Heat exchange surface, m2

Compactness of heat exchange surface, m2/m3

Thermal power density in the zone of

effective heat exchange, MW/m3

Mass in dry state, t

V. S. KUUL. A. V. PICHUGIN, O. B. SAMOILOV

OKB Mechanical Engineering,

Nizhny Novgorod,

Russian Federation

Abstract

World nuclear power rends constantly to decrease the permitted ultimate dose of NPP personnel and population during normal operation and accidents and NPP decommissioning.

1. ISSUES AND HOW TO DECIDE THEM WHEN DECOMMISSIONING NPP

The conditions and radiological safety of reactor decommissioning project are determined by:

1) decommissioning concept ("immediate" or "delayed" dismounting).

2) structure and equipment radioactivity;

3) radioactive waste amount (activity, mass, volume);

4) technology of removable and non-removable equipment;

5) doses of personnel and population.

It is known, when decommissioning NPP with BWR or PWR, that radioactive waste forms an activity of more than 10 MCi. The total mass of waste at NPP decommissioning is several hundreds thousands of tons, and about 1-2% of the waste has high and medium radioactivity, which should be disposed off. The potential danger of radioactivity requires protective measures, eliminating release of radioactive materials into the environment.

Irradiation doses of the personnel involved in NPP decommissioning range from several hundreds to tens of thousands of man. rem depending on the decommissioning strategy adopted.

One of the main problems of NPP decommissioning is handling of hot large equipment. NPP reactors accumulate substantial radioactivity (up to 10 5 Li), which causes high radiation levels from 1 to 10 Sv/hr

Dismounting and disassembly of vessel and structures requires development of special remotely controlled equipment, heavy protective flasks and, hence, large costs or prolonged plant preservation that reduces irradiation doses, potential releases and environment contamination.

But, from the point of view of economic, quick rehabilitation of the site is not evidently advantageous as compared with immediate dismantling.

A safety depressurization system is provided on some designs where it is simple and economic to do so Blow down into pressure suppression tanks is employed in SPWR and SIR

The integral water-cooled reactor with the pressurizer inside the pressure vessel gives a more direct and efficient connection between the pressure vessel and the emergency relief valves than in present generation reactors where the pressurizer and relief valves are separated by piping

VPBER-600 has an additional possibility of a partial depressurization which operates when the gravity boron injection tanks are discharged by connection to the pressurizer steam space

Depressurization is also achieved through the heat exchangers or steam generators used for decay heat removal There are other depressurization systems such as spray which are under operator control

To investigate the possibility of passive initiation of the isolation valve located inside Safe Guard Vessel, an experimental study is being performed.

Heat Pipe Application for PRHRS

A separate type heat pipe is experimentally studied for PRHRS. A computer code is developed to model the thermo-hydraulic behavior of the heat pipe.

Helically Coiled Tube Once Through Steam Generator

The thermal hydraulic design and performance analysis computer code, ONCESG, for a once through steam generator is developed. An experimental study is being performed to generate the heat transfer correlation and pressure drop correlation of the helically coiled tube once through steam generator.

Critical Heat Flux Test

An experimental test facility is being constructed to study critical heat flux and pressure drop correlation for the tight latticed hexagonal fuel assembly.

Steam Injector Application for PCCS

An experimental study is being conducted on a steam injector driven passive containment cooling system. A computer code is developed to model the thermo-hydraulic behavior of the steam injector.

An experimental investigation is being conducted for material selection and performance test for the wet thermal insulation system.