An integral reactor with forced coolant circulation at emergency power level operation and natural circulation for residual heat removal is used in the design of the VPBER-600 reactor plant for the power unit of a new generation NPP of 640 MW (el) power. ‘

Forced coolant circulation is provided with the help of six leak-tight circulation electric pumps, located on the bottom of the reactor vessel.

In the design of some equipment and systems the decisions have been mode which have been verified by long experience of operation of the existing nuclear power plants.

Calculation analysis of the wide range of accidents, performed on the base of both deterministic and probabilistic approaches demonstrated the high safety of the plant. Safety in the course of three days is provided by passive means without power supply from outside nor personnel intervention.

Form the point of view of a deterministic approach for severe core damage, multiple failures of safety systems elements and systems as a whole are necessary. The probability of severe damage to the core, evaluated deliberately conservatively is < 10’8 per reactor/year.

Nevertheless the search for design decisions for optimization of the reactor design and improvement of characteristics including safety provision for severe accident-accidents with postulated melting of the core continues.

As a result circulation pumps were moved from the bottom to the cylindrical part of the reactor vessel, reactor internal heat exchangers of the system for emergency heat removal were excluded, engineering decisions for the limitation of the consequences of severe accident were proposed and the possibility of corium confinement in the reactor vessel or guard vessel was shown.

Moving the circulation pumps to the cylindrical part of the vessel simplifies the operational servicing of the reactor, excludes the possibility of coolant leakage below the core and improves the conditions for corium confinement in the core and in the reactor vessel and for creation of an in-reactor corium catcher.

The schematic diagram of the system of severe accident localization, presented in Fig.3, includes:

— heat exchangers-condensers, of the system of purification and boric reactivity compensation, total power 10 MW are located in the guard vessel;

— two safety complexes (DN 100), each consisting of a membrane-rupture device and a safety valve in series;

— two temperature-actuated devices for the reactor pressure relief;

— bubbler of approximately 300 m3 volume;

— receivers of 600-700 m3 volume.

The bubbler and receivers have the same strength as the guard vessel. In a severe accident with core melting when the temperature in the reactor reaches 500-700°C, safe

SYSTEM OF SEVERE ACCIDENTS LOCALIZTION

Complex System of VPBER-600 Thermo-physics and Safety Investigations
Build-In steam-gas pressorizer
Investigation* of dynamics at normal and emergency conditions
Investigations of Static dtrtnbuUont of temperature* and gas
О as transfer processes investigations
Passive system of heat removal
Investigations of the efficiency of dry fleam generator In steam and steam-gas mixture
Investigation of tesmal efficiency of heat-exchanger-condenser in ateam-gaj medium
Power determinations, trying-out of start up conditions in water
Loss of tightness accidents
Investigation of termo — hydndic processes in the reactor system-guard vessel	Determination of the efficiency of sbtems/ influencing upon the accident run (system of passive heat removal, hydraulic accumulator	InveshgiiKms of icihi hydraulics of primary «.псин at tuiural circulation and loss of circulation
Hydrodynamics of the core Inlet chamber
Hydrodynamics investigation and development of live structure for rates eguabzation at core inkt
Determination of temperatures variations at partially disconnected steam generators
Complex Investigation of the KP at normal and emergency conditions
Aerodynamic model
Severe accidents with the core failure
Cooling of the reactor Investigation In accordance Cooling of the core vessel and guard vessel with complex program on severe accident* in steam gas medium llc. il removal
Fig.4

temperature devices of the reactor are opened and steam-gas mixture is discharged from the reactor to the guard vessel, untilpressure is equalized in the guard vessel and the reactor. When pressure in the guard vessel is 5.0 MPa a membrane-rupture device on the guard vessel is broken, a safety valve is opened and steam-gas mixture is discharged from the guard vessel to the bubbler. Gases liberated in core melting are pressed out to receivers, this passively solves the problem of provision for hydrogen safety.

Heat removal from the reactor vessel when corium is confined in the reactor or from the guard vessel false bottom when corium is confined in the guard vessel if it leaves the reactor vessel is performed with the help of heat exchangers-condensers in the guard vessel of the purification system. —

In normal operation heat exchangers in the guard vessel of the purification system are disconnected from heat exchangers unit by the pipeline for water supply and are connected remotely by the operator in the event of severe accident.

There are 96 fuel assemblies m the active core The expected region discharge bum-up is 30000 Mwd/T One fourth of the fuel assemblies are scheduled to be refueled after each 5 winter seasons operation Therefore, the total discharged spent fuel assemblies at the end of lifetime of NHR-200 are only about the amount of another two cores of fuel assemblies It makes possible to storage the whole spent fuel inside the RPV There are m total 200 cells divided mto two layers for spent fuel assembly storage surround the active core A stand-by small pit for defective fuel assemblies is reserved m the reactor building This arrangement makes refueling and spent fuel storage equipment and facilities quite simple In many countnes the concept of spent fuel storage inside RPV is under developing to mcrease the bum-up of fuel For NHR — 200 this arrangement also could mcrease the bum-up about 15%

The accident analysis shows that the design features of NHR-200 provides an excellent safety characteristics not only with a low probability of initiators, but also with a unserious consequences of accidents. The results of LOCA analysis of NHR-200 are listed in Table 2[2],

5. Conclusion

In light of the requirement on safety and economy a NHR should be designed with inherent safe features instead of complex safety systems. An integrated arrangement is one of the key characteristics The experiences gained from NHR-5 four winter seasons operation and the practice of NHR-200 design show that an integrated arrangement with self-pressurized performance, natural circulation and in-vessel spent fuel storage is realistic and has many advantages.

Reference

[1] . Zhang Dafang, Dong Duo etc. " 5 Mw Nuclear Heaung Reactor", Internal Report. (1994. 3).

[2] . LiJincai, Gao Zuymg etc. " Safety Analyses for NHR-200", Internal Report. (1994. 3).

&EXT PAGE(S)
left BLANK

Whether or not nuclear ship has economic benefit is a fundamental matter to judge the social incentive to realize the practical use of nuclear ships. To evaluate the economics of nuclear ships, the construction and operation costs of nuclear and conventional ships have been studied and compared as parameters of the ship’s speed, containers’ capacity in terms of TEU (Twenty — feet Equivalent Unit), etc. Figure 5 shows the comparison of RFR (Required Freight Rate, $/TEU : Operation cost to transport one container) between two container ships of 6,000 TEU and 30 knots (one uses reactors equipped with two MRXs of total power 348MWth and the other a diesel engine), in case of Asia — North America route, commissioned for 20 years from the year 2015. The quantity of container transport through this route is the largest among the the world’s three biggest sea routes (Routes of Asia-North America, Europe — North America and Asia-Europe). The crude oil price is assumed to be $36/bbl. averaged within the service period. This figure shows, (a) The capital cost of the nuclear ship is about 2 times larger than that of the diesel ship, (b) On the other hand, the fuel cost of the nuclear ship is about 1/2 of that of the

Ship Speed (knots)

Fig.6 RFR as a function of ship speed

diesel ship, (c) The environmental cost of the diesel ship accounts for 22% of the RFR. The cost study of nuclear and conventional ships shows that the situation becomes more favorable for nuclear ships as increasing of the speed and the load. Figure 6 indicates the RFR as a function of ship speed. Over 28 knots, the nuclear ship holds an economically dominant position over the diesel ships, because the diesel ships must have higher environmental costs.

3. R&D work

To realize nuclear ships with commercial use, it is necessary to obtain an economical, safe and reliable reactor system. In addition, the supplementary items such as the international agreement on safety, the preparation of the maintenance yard, etc., should be solved. R&D programs on nuclear ships have been proposed and being performed to solve technical subjects so that the nuclear ships will be put into commercial use in the future.

(1) Experimental study on thermal-hydraulics

To study the hydrothermal behavior in the water-filled CV, especially steam condensation characteristics, a small scale test facility (volume ratio : about 1/300 of MRX) has been fabricated and fundamental experiments on the following behavior are in progressC5>, (a) thermal and hydraulic responses in both the RPV and the water-filled CV under LOCAs, (b) evaluation of mechanical loads generated by LOCAs, and (c) capability of natural circulation and decay heat removal. Furthermore, to confirm the function of the safety features such as an integral PWR with a water filled CV and passive safety systems, a large scale synthetic test facility is planned. The thermal power of the facility is 5MW (1/20 of the MRX), however the height is same as that of the MRX because it is most important to simulate accurately the natural circulation condition. Experiments on board are planned to obtain the behavior under the ship inclination and oscillation.

(2) Development of components

The components of in-vessel type CRDMs such as the motor, the latch magnet, etc., have been developed <6>, and the function and reliability tests using the full mock-up CRDMs are planned. The water-proofed design is being performed for the components and the thermal insulator placed in the water- filled CV.

(3) Automatic control system

It is important to reduce the number of reactor operators from the view of the ship’s economy. From this standpoint and to enhance the ship’s safety, highly automated control systems have been studied which will be adopted and will cover the whole operations during normal, abnormal and accident conditions. This system consists of control and diagnostic systems as shown in Fig. 7. The control systems generate control signals for control equipments, for example, control rods, pressure control valves, flow control valves, etc., in accordance with the reference signals (demand signal) and signals of each measured parameter. If a difference exists between the signals of the reference and the parameter, a control signal is generated based on the operational procedure and changes the parameter to the demand conditions. The diagnostic systems are provided to monitor the malfunction of the systems and

the plant operating conditions. For the operational procedures, the optimization is made based on the operator’s knowledge and the learning AI.

(4) Development of nuclear ship simulator

The nuclear ship simulator NESSY (Nuclear Ship Engineering Simulation System) has been developed in JAERI and used for the simulation of the nuclear ship "Mutsu" so farC7:>. It can simulate both the behaviors of the reactor systems and the ship motions. Mutual interactions can be analyzed for the situations such as changes of the reactor power, the steam generator water level, etc., due to the ship motions caused by waves or the navigator’s maneuvering.

The accuracy of the system has been verified with the operation data of the "Mutsu" and it is proved that this system is a very useful tool for developing advanced marine reactors<e>. Modifications of the models and the parameters are being made for the MRX reactors since 1995.

4. Conclusion

An advanced marine reactor, MRX, has been designed to be more compact and lightweight with enhanced safety. The engineered safety is accomplished through a simplified system which is suitable in particular for a marine reactor, since it must be operated by limited number of crews. The LOCA analysis shows that the core flooding is maintained passively even taking into account the ship inclination. The one-piece removal method is proposed to keep the maintenance and refueling works short and safe. This method also makes the ship’s decommissioning easy and enables us to reuse the reactor after the ship’s life. The economic evaluation shows that for container ships of 6000 TEU traveling over 28 knots, the nuclear ships will hold an economically dominant position after 20 years from the present time, because the diesel ships must have higher environmental costs. In addition to the design study, extensive R&D activities are being performed. These can contribute largely to the realization of the nuclear ship in commercial use in the future. Considering that the MRXs are small size reactors with highly safe capabilities and transferable ones, they have a wide variety of uses in the energy supply system.

Reference

(1) A. Yamaji and K. Sako : Shielding Design to Obtain Compact Marine Reactor, J. Nucl. Sci. Technol., Vol. 31, No. 6, 1994.

(2) K. Sako, et al. : Advanced Marine Reactor MRX, Int. Conf. on Design and Safety of Advanced Nuclear Power Plants, Oct. 1992, Tokyo, Japan.

(3) T. Hoshi, et al. : R&D Status of an Integral Type Small Reactor in JAERI, ICONE-3, Apr. 1995, Kyoto, Japan.

(4) A. Yamaji, et al. : Core Design and Safety System of Advanced Marine Reactor MRX, ibid.

(5) T. Kusunoki, et al. : Steam Condensation Behavior of High Pressure Water’s Blowdown Directory into Water in Containment under LOCA, ibid.

(6) Y. Ishizaka, et al. : Development of A Built-in Type Control Rod Drive Mechanism (CRDM) for Advanced Marine Reactor MRX, Int. Conf. on Design and Safety of Advanced Nuclear Power Plants, Oct. 1992, Tokyo, Japan.

(7) T. Kusunoki, et al. : Development NESSY (Nuclear Ship Engineering Simulation System) and Its Application to Dynamic Analysis, ibid.

(8) M. Ochiai, et al. : Present Status of Nuclear Ship Engineering Simulation System, The 1994 SCS Simulation Conference, Apr. 1994, San Diego, U. S.A.

The plant is controlled by a microprocessor based distributed system with a redundant communication network. The critical functions of the system required for the plant to operate, are also duplicated.

The protection system, also based in microprocessors, uses a minimum of programmable actions. The software is written directly in machine language (assembler), with a strong effort made to keep it extremely simple. Apart from these small pieces of software, the protection system is "hard".

These features are similar to most recent designs of non-integrated PWRs: they are not a particular characteristic of integrated reactors but a new trend in nuclear power plant design.

&

/

The CAREM reactor uses hydraulic control rod drives. This type of rods are wholly contained in the pressure vessel, carrying the concept of integration one step further.

A sketch of its functioning is included. Rods are kept in position by water flow. A positive flow pulse causes the rods to climb one step, while a negative flow pulse causes the rods to go down one step. If water flow is suddenly interrupted, the rods, in a fail safe action, are dropped into the core. The water, after passing through the drives, is dumped into the primary system, inside the pressure vessel. The main advantage of this type of drive is that it can fit in the volume available inside the pressure vessel of an integrated reactor. The use of conventional electromagnetic control rod drives would require a higher containment.

The protection system is able to trigger a scram using a built in valve, that causes the rods to fall by gravity.

However, the

drives are designed in a fail safe way: rods will fall automatically in the event of a plant black-out or in a loss of coolant accident, due to the interruption of the water flow through the drives, without requiring a trip from the protection system.

Design and development work for the Argentinean project called CAREM is continuing The project is ten years old and the initial design power level was 15 MWe The current work is on a 25 MWe reactor, with thought being given to a 100 MWe reactor. Engineering for the 25 MWe plant is scheduled for completion by the end of 1996. Financial, political and siting decisions are expected to be made in 1996, which if favorable, will lead to construction in 1997. Experimental work is under way in a high pressure loop to study critical heat flux (CHF) and dynamic response, and to make a comprehensive study of the reactor physics of the core in the RA-8 critical facility Development and testing is being carried out on the core internals, control rod drives & position indicators and the reactor protection system, especially the trip system

1.2. China

Loss-of-coolant experiments are being carried out for the Chinese nuclear heating reactor (NHR200). A test loop has been in operation since 1989, and allows tests on the three possible positions for a break to occur, (i) pipes on the upper plenum, (ii) steam generator pipe break and (iii) boron injection pipe below the vessel water level. Experiments have shown good agreement with the results of calculations on the influence of the break position on RPV water level, on discharge quality and hence on the depressurization rate. Depressurization is quite slow (thousands of seconds), due to the small size of pipe connections.

China is carrying out a study on the choice of a reactor system for a co­generation plant. The study is based on a 2×450 MWt plant and hinges on the configuration of the intermediate loop A comparison is being made between two alternatives; (i) steam generator in the RPV supplying steam to the turbine and an external steam water heat exchanger, and (ii) a system with an in-reactor high pressure water heater which provides hot water to generate steam in an inverted U-tube steam generator The steam passes to the turbine and back-pressure steam provides the heating load.

The second alternative allows a lower primary pressure and hence a saving in cost The possibility of radioactive leakage into the tertiary circuit is reduced due to the high reliability of water-water heat exchangers and by having a secondary pressure higher than the primary pressure. The overall efficiency loss from having a 10 MPa primary is offset by the higher efficiency of the second heat exchanger which is a steam generator rather than a low efficiency steam/steam heat exchanger

З 3 Indonesia

Indonesia is presently giving serious consideration to the introduction of nuclear power Strategic planning in Indonesia also envisages the utilization of an integral reactor design for the supply of heat and electricity to many of its islands. The perspective plan is to Install 12,000 MWe of nuclear power by 2019, of which the majority will be based on large reactors of existing designs. There is, however, a potential market for a 30 MWe design, suitable for several small islands in Indonesia resulting from the high cost of transportation of fossil fuel

3 4 Italy

The reactor continuing to be developed in Italy is named Inherently Safe Immersed System (ISIS). This reactor has components which are passive at category "B" in the IAEA definition which means that they need no valve movements or logic circuits for system initiation. The reactor uses some of the concepts used in the "PIUS" design but has a very small heated primary inventory with density locks to give access to highly pressurized cold water in the event of a reactor malfunction. Data on safety analysis shows the very high level of safety that can be achieved The problems of deployment are now economic rather than safety related The system could be competitive in a co-generation mode, perhaps with a pressure reduction to decrease the mass of steel needed. Further work on tackling the economic competitiveness aspects is under way.

The Program is divided in two main subjects

Dynamic Tests/3/

In order to reduce the range of experiments it was defined that the studies will be limited to the regimes, states and perturbations foreseen for the CAREM 25. The studies will be limited to those parameters included in the modelling by computer codes.

Several conditions were established:

• The primary subcooled state is relevant only during the start up of the CAREM

• The existence of secondary subcooled and saturated states will depend on the method adopted for the start up. The situation can be avoided if there is steam available."

• Nitrogen in the dome will be limited up to 30 Kg/cm2.

• Possible perturbations in the Primary Side are produced by neutronic changes and power extraction form the secondary side. Parameters of interest are:thermal balances; water flow; thermal transference coefficients in SG, dome, and void fractions in the riser; Primary and secondary pressure and temperature; dome volume; hydraulic drag; neutron kinetics.

This stage of experiments is at present underway and will be finished by march of next year.

CHF Tests /4/, /5/ and /6/

It is known that to perform reliable thermalhydraulic calculations it is necessary to perform CHF experiments for the start up and nominal power states. In this case the CAPCN will be used with some modifications in the control system in order to obtain stable values for the pressure, water flow, water quality, with a positive power ramp. Part of the experimental work to be started next year will be devoted to determine limits and operational conditions for the CHF experiments. A detailed study was already done with the corresponding Program. For the CHF experiments themselves it was necessary to make carefiill studies because it was not possible to install a section of the same dimensions as for CAREM Fuel Element. Variables kept as in CAREM are rod diameter, pitch and ratio total section/water flow section. Preliminary experiments for temperature oscillations due to slug flow patterns will be made before starting the CHF experiments.

During normal operation the hot/cold water interface level in the Lower Density Lock is maintained constant by varying the speed of the Primary Pumps.

Any rising of the interface level is counteracted by an increase of the pump motor speed. Any lowering of the interface level is counteracted by slowing down the pump speed.

Dynamic Analysis of ISIS Control System is in progress. Preliminary results, not yet published, confirm that the reactor power is controlled by the concentration of boron in the primary water and by the intrinsic negative feedback of the core.

Yu. D. BARANAEV, Yu. I. OREKHOV,

Yu. A. SERGEEV, I. M. SHVEDENKO Institute of Physics and Power Engineering,

Obninsk, Kuluga Region

Yu. P. FADEEV

OKB Mechanical Engineering,

Nizhny Novgorod

V. M. VOROBYEV CDB Baltudoproject,

St. Petersburg

Russian Federation

Abstract

Several independent active and passive safety systems are employed in the design of integral WCR in order to provide together with the reactor inherent safety features realization of the emergency heat removal function:

— active primary coolant clean-up system transmitting heat to the tertiary and then to the forth circuit,

— passive emergency heat removal through steam generator to the atmosphere,

— reactor cavity flooding system providing heat transfer from the reactor vessel to the reactor metal-water shielding tank,

— high and low pressure make-up system using water from the special storage tank. ‘

Operation of the systems under LOCAs and accidents with normal heat removal systems failure is considered in the paper. Special emphasis is done to the description of systems characteristics, systems interaction and modes of operation which are influenced by the reactor integral design.

Emergency heat removal (EHR) is one of the major safety functions that should be provided by the design of reactor safety systems.

Under the design of the ABV-6 reactor designated for the floating NPP "Volnolom" the following specific features of the reactor were taken into account:

— Integral reactor design (Fig. 1)

— Low reactor capacity (38 MWt)

— High reactor heat accumulation capability (~1.8 s per °K)

— Natural convection in the reactor primary circuit

— High margin in the strength characteristics of the primary circuit equipment (tolerable

pressure of 31 MPa is ~2 times higher than the nominal design pressure)

— Employment of metal-water shielding tank containing 26 m^ of water

— High thermal conductivity of reactor fuel made of uranium-aluminum alloy.

These features of the ABV-6 reactor strongly influence the solution of the EHR

issues.

1. The integral arrangement of the primary circuit results in exclusion of big primary pipelines as well as a larger primary coolant inventory in the reactor pressure vessel (RPV). These result in long period before it is necessary to start coolant supply to the RPV under LOCAs, since reactor core uncovery in the case of reactor make-up system failure would take place only several hours after the beginning of the accident. Besides, there is a possibility for the EHR through steam-generator in integral reactor design and this also enlarges the grace period.

2. The relative portion of heat dispersed to the surroundings is higher at small reactor capacity. This allows decrease in the time required for the EHR systems to operate.

3. The high reactor heat capacity allows reduction of the EHR system capacity compared with the residual heat rate. The excess of extracted heat is accumulated in the reactor primary coolant and structures in the first stage of the accident without unacceptable reactor temperature and pressure rise.

4. Natural convection ensures reliable heat removal from the reactor core with sufficient departure from nucleate boiling margin during transient and accident conditions.

5. Large margins in the strength of the primary circuit together with high heat capacity of the reactor ensure tightness of the primary circuit under accidents with failure of all EHR trains for a long period of time. For some beyond the design basis sequences this time is unlimited.

6. The metal-water shielding tank is an effective heat sink. The amount of water in the tank is sufficient to provide the EHR for 3 days.

The ABV-6 reactor plant is equipped with the following systems that can be used for the EHR:

— Active two trains system supplying water to the steam generator (SG) from the feed water storage tank (48 m^ ) by the emergency feed water pumps with a flow rate up to 15 m^/h each (Fig.2).

— Passive two trains system supplying water to the steam generator from pressurized storage tanks containing 4 m^ of water (Fig.3).

— Active primary coolant purification system with pump and heat exchanger designed to remove 1370 kW (3.3% of reactor rated power) of heat directly from the primary water (Fig-4).

— Active two trains emergency core cooling system (ECCS) supplying water to the reactor vessel by 3 high-pressure (head 17 MPa, flow rate 1.2 nP/h each) and 2 low-pressure (head 3.5 MPa, flow rate 20 m^/h each) pumps (Fig.5).

ABV-6 REACTOR SAFETY CHARACTERISTICS UNDER BDBA

CT

vO

— Passive reactor cavity flooding system with a pressurized water tank up to 7 MPa containing 1.2 m3 of water (Fig. 6).

Safety analysis for all initial events considered has demonstrated a high level of reactor safety system effectiveness and sufficiency under design and beyond the design basis sequences. Some results of the safety analysis are shown in the Table.

It is reasonable to add that according to the PSA, the frequency of black-out with simultaneous failure of two passive EHR system trains does not exceed 10"10 per reactor — year. Cumulative frequency of core melt was assessed as 2.5*1 O’7 per reactor-year.

The above information allows the conclusion that the integral design concept provides broad range of opportunities to enhance reactor safety with reliance on clear and simple design solutions. This concept looks rather promising at least for small water-cooled reactors.

The hydraulic pressure test is an important testing operation which is required by all relevant codes and standards. The test gives the final confirmation of the integrity and strength of the pressure vessel. The arrangement of of a large integral RPV about 20 m heigh will be similar to one of a BVR RPV having similar outer dimensions. The test pressure the value of which is different in individual national codes is in the range of 1.1 — 1.6 of the operational pressure, will be comparable with one of PVR RPVs, i. e. 17.3 — 25 MPa.

Check assemhlу

Good experience has been made at Skoda with application of the final check assembly of reactor internals. which was recommended by the original Russian technical project. This operation consisting in assembly of all reactor internals at the manufacturing shop made possible to find an optimal position of individual parts and remove all problems in the shop. As a result. assemblies of reactors and Installation if internals into pressure vessels on sites vere smooth and fast.

There are more components in integral reactors: in addition to classic internals there are steam-generating elements, water-stream guiding structures etc. A specific feature is a long distance between control rod drives and the core where control rods are placed. For such a complex lay-out, the check assembly in the shop could be a good confirmation of mutual fitting of individual components.

Large and heavy parts of integral RPVs require appropriate shop transporting and handling means.

Principal handling an transporting operations include’­- transport of semiproduts to the shop.

— transport of workpieces including the whole RPV between individual workplaces and machines.

— tilting of workpieces from the horizontal position to vertical one and back; finally, the whole pressure vessel have to be erected to the vertical position for the pressure test and back.

— rotating workpieces during fabrication (welding, cladding) and і пресtions (ultrasonic tests, radiography),

— loading workpieces to the annealing furnace, radiographic cells etc. .

— installation of components during the check assembly.

— final expedition of the RPV from the shop (loadincj on a truck or ship).

The necessary transporting and handling means are rotating and/or tilting positioners. trucks. wagons and cranes of corresponding loading capacity. Cranes have to have an appropriate lifting travel and — for a precise setting

— a microdrive.

There is a need to improve the availability of data and the method for dealing with the effects of nitrogen originating in gas pressurizers, on heat transfer in emergency conditions.

5.1.1. Severe accident in an integral reactor

There is a significant increase in safety of the integral reactor in comparison with the loop type reactor, due to a reduction in accident initiators and the use of passive safety features and inherent characteristics.

Integral reactor accident sequences have not been fully analyzed. It would be expedient to carry out such analyses since the results could influence reactor design. It is also necessary to carry out probabilistic safety analysis.

The reliability of available codes (RELAP/SCDAP/MELCOR) should be evaluated for severe accident analysis in integral reactors and the need for modification and validation determined. International cooperation could play a key role in code validation for integral reactors.