In the Russian Federation, a series of integral reactor designs are being actively pursued by the Experimental Machine Building Design Bureau (OKBM).

The reactor called Atom Thermal Electric Plant (ATEC 200) comes in sizes from 80 to 250 MWe. They have natural circulation systems for residual heat removal, positioned in the upper head. They are intended for use in remote locations and are designed to be sited below ground level.

The "AST 500" is a natural circulation district heating reactor which is ready for operation but due to public acceptance problems, the project has been suspended for the present. The safety of its design and operation have been reviewed by an IAEA OSART team and found to be adequate.

The Passive Safety Integral Reactor Plant (VPBER 600) is a 640 MWe reactor designed earlier A number of significant changes have been made to its design, including moving the pumps from the bottom of the vessel to a position above the core, and addition of a core catcher. A comprehensive in-service test and inspection programme has been set up and equipment design carried out.

Integral Reactor for a Floating NPP (NIKA 120) is designed by the Research and Development Institute for Power Engineering (RDIPE) as a floating power unit for use in northern Russia and is under development. The plant consists of two reactors, each of 42 MWt. The reactor vessels are enclosed in a safeguard vessel which is immersed in a bubbling tank. The fuel is 21% U-235 as UO2 in a zirconium matrix.

Control rods are designed to stay inserted even if the floating unit is inverted in the sea.

Development work is also being carried out on a small unit called the Autonomous Co-generation NPP (UNITHERM), with a power of 30 MWt or 6 MWe for use in difficult-to-reach regions. Emphasis is on maintenance-free operation between the annual visits of the maintenance team. There is an intermediate heat exchanger to isolate the final steam supply including that to the turbine, from any risk of radioactive contamination. Such an intermediate circuit is necessary for a district heating system to act as a pressure barrier Maintenance of the steam generator and turbine can be carried out using conventional methods as the steam is free from radioactive contamination. The penalty is an increased generation cost due to loss of thermal efficiency and the cost of an extra plant system.

Development work is in progress with regard to emergency heat removal systems on the "ABV-6" reactor system, planned for installation in the floating nuclear power plant, "Volonolom". This is a 38 MWt plant using a uranium-aluminum(U-Al) alloy fuel in a natural circulation reactor with nitrogen pressurization There are five

emergency heat removal paths some of which are dedicated systems and others are shared systems(e. g use of the coolant purification system heat exchanger for emergency heat removal). There is a passive system using the flow of stored water to the normal steam generators and discharging the resulting steam to the atmosphere The ABV-6 containment system embodies two rupture discs The first releases steam from the reactor containment shell to the pressure suppression pool, and the second releases pressure in the suppression pool compartment to the auxiliary building The reactor and containment systems were modeled using the computer code RELAP5 MOD-3, with modification of some code modules to allow for release of dissolved nitrogen A severe accident scenario with a double ended break of the pressurizer surge line, failure of the emergency core cooling system (ECCS), failure of the shielding tank coolers and no operator action, was modeled. The results of the analysis have conclusively shown the very high level of safety achievable by this design.

For many years, OKBM has been conducting design and development work on highly efficient cassette steam generators. These cassettes are produced on an automated assembly line and have been successfully used in many reactors They are also specified for VPBER 600 design, where 216 independent sub-sections are assembled into two different shapes of boxes to fill the annular space available The design is based on straight tube steam generators with secondary flow inside the tubes to ensure that they are always in compression There is considerable practical experience with these cassette steam generators in WER plants (500,000 hr) and they are backed by a large programme of developmental tests

OKBM is also working on radiological aspects of decommissioning integral reactors The main advantage of integral reactors results from the large water filled space between the core and the pressure vessel This results m a reduction in activation

level of the RPV by a factor of up to 10^ compared to WER reactors, with a corresponding reduction in activity in the adjacent concrete structures The overall effect on occupational dose during decommissioning, involving breaking up and removing all active plant components, is a reduction by a factor of ten

Design work on a new concept called Integral Reactor with Inherent Safety (IRIS) based on the "PIUS" reactor was started under the leadership of IPPE It is a 600 MWe reactor with natural circulation The steam generator is within the main vessel and the upper density lock is replaced by a by-pass tube which descends below the water level in the outer tank, which in turn, is below the water level in the reactor vessel during normal operation An increase in power leads to vapour formation and flow of primary water/steam into the outer tank Borated water enters the primary through the lower density lock and shuts the reactor down The design is still in the conceptual stage

Design work at the Kurchatov Institute is continuing on the use of super critical water as the primary fluid in the "V-500 SKDI" reactor With subcritical water, power increase is limited by the possibility of departure from nucleate boiling (DNB) This possibility is eliminated in super critical conditions, since the fluid remains in a single phase through all the temperature range Furthermore, the enthalpy of super critical water approaches that of steam, giving improved heat transfer in the steam generator Density variation with temperature provides an excellent negative reactivity

feedback for reactor stability The water density effect is strong enough to allow compensation for reactivity changes with bum up, without control rod movements and only a small change in primary temperature Some modem fossil-fueled stations operate with supercritical water in the boiler at 26-28 MPa and 560-580 K, giving a basis of practical experience for the use of these conditions Safety analyses have confirmed the high level of safety achievable by this concept, even in a very severe transient caused by multiple failures

3.1 Description

The RA-8 critical assembly has been designed and constructed as an experimental facility to measure neutronic parameters of the CAREM NPP, under contract and supervision of the National Atomic Energy Commission (CNEA) of Argentina. It may be used, with relatively minor changes, as a facility to perform experiments for other light water reactors. It provides a reactor shielding block and reactor tanks that can be adapted to hold custom designed reactor cores.

The RA-8 critical facility is located in the PILCA IV Sector of the PILCANIYEU TECHNOLOGICAL COMPLEX, in the Province of Rio Negro, Argentina, at approximately 30 km East of San Carlos de Bariloche. It occupies the main hall of a building shared with the Laboratory for Thermalhydraulic Tests (LET), described in Section 2 of this report, and other special facilities for CAREM Project Geometry and location of core shielding inside the main hall are such that radiation dose levels are acceptable in adjacent rooms, for all operational conditions.

General Characteristics of the RA-8, are:

Low operating power, which makes cooling systems unnecessary.

Extinction systems. •

Rapid insertion of control rods Dumping of the moderator

Regulation and safety rods. There are at present 13 mechanisms to drive control rods in and out of the core. The Control System allows the definition and use of some of the control rods as Regulation Rods, and some as Safety Rods. The number of rods assigned to each function depends on the specific core being tested.

Argentine Regulatory Authority (ENREN) imposes the following requirements for the design of critical facilities:

• Negative reactivity introduced by control rods must be higher than 50% of the critical assembly reactivity excess.

• Core reactivity with control rods must be negative and higher than 3000 pcm.

• Core must remain subcritical in at least 500 pcm after extraction of the control rod of maximum negative reactivity

• Reactivity worth of control rods defined as regulation rods must be such that their insertion makes the core remain subcritic in at least 500 pcm.

• Movement of any control rod mechanism must not produce a reactivity insertion higher than 20 pcm/sec.

Operating modes. There are two possible ways of operation:

Operation by critical height. (Reactivity is determined by the moderator level surrounding the core)

Operation with control rods.( Reactivity is regulated by the amount of absorbing material introduced in the core)

Water System: water level is controlled by the RA-8 Control System. During operation, water fills simultaneously two

concentric and connected tanks. The inner tank is designed to hold the core, its structural components, and nuclear instrumentation. The water in the outer tank serves the purposes of shielding and reflector. Filling of the tanks is performed in two stages: a first stage of fast pumping, followed by a second stage of slow pumping to approach operating level. Safety Logic takes into account the position of safety and control rods to allow pumping of the moderator into the reactor tanks. Tanks are emptied by the opening of two butterfly valves, centred in the inner tank, which dumps water into the hold-up cistern, below the reactor block. It lakes no more than 4 seconds to empty the inner tank.

The water system also has provisions to add boron and to clean, drain and recirculate water. Water temperature can be varied in up to approximately 75 °С. The hold-up cistern has its own water recirculation system.

Data acquisition is done simultaneously and independently by two means: the “hard logic” Instrumentation and Control system needed to operate the facility, and the Control and Data Acquisition System, a microprocessor based system, by means of which the reactor operator will be informed of reactor and experiment related parameters.

At present the facility is being completed and the cold initial start-up is programmed for the end of the present year (1995). The core elements (Fuel Rods and absorbers) for the RA-8 are in manufacturing process and expected to be finished in the first half of next year (1996). The experimental program is foreseen to last for about one year and a half.

3.2. Experimental Program

The cores to be used for CAREM related experiments are made from fuel rods with the same radial geometry of the ones for CAREM, but shorters with a length of 80 cm. The pitch of the core was studied by calculation and is in principle the same as for the CAREM. The core calculations are made with a Diffusion Code (C1TVAP), so the size of the core has to be enough to have good results. A central homogeneous zone is needed to study perturbations as: rods loaded with different concentrations of burnable poisons, absorbing rods, guide tubes, structural materials, etc. The maximum core reactivity covering all the experiments is around 7500 pcm. To meet the requirements given by the ENREN and with the experiments, plate type absorbers made from bare Ag-ln-Cd are used. The distribution of the absorber elements is given in the following figure. The central absorber is not shown.

The studies will be conducted over different cores using two enrichments (E= 1.8 % and E= 3.4 %), some of the defined cores arc:

A. One region of E= 1.8 %.

В Two regions: inner of E= 3.4%, outer of E= 1.8%.

C. One region of E= 1.8% , perturbed with non fuel rods (guide tubes, control rods, burnable poisons) homogeneously distributed in the core.

D Two regions, the inner with E= 1.8% and perturbed with non fuel rods, the outer region with E= 3 .4%.

E. Two region, the inner with fuel rods resembling CAREM fuel elements, the outer with the needed fuel rods to reach enough reactivity to perform experiments with different configurations for the CAREM FE.

A detailed experimental program defines the experiments to be conducted with each core type. Some of the already defined measurements are:

Influence of different boron concentrations at different temperatures. Extensive properties as : critical height, critical buckling and reflector saving. Intensive properties as disadvantage factors, fission ratio (U235 & U 238), epithermal to thermal fissions ratio, epithermal to thermal absorption ratio, average spectra in fuel and moderator. Fluxes and spectra in non fuel rods (control rod guides); similar in macrocells (assembly of non fuel rods with neighbouring fuel rods). Power distribution. Mutual influence in CAREM Control Rods positioning Reactivity changes for different Boron concentrations, different temperatures and different void fractions. Determination of control rods reactivity and fuel rods with different concentrations of burnable poisons.

NOTA5 .03

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i) CoMrdod toto’ de ogujeros diametro 9 5 • 3500 ( tres mil

dumtcnlos ) ‘_>otгg ФЬЪО *

3. Benchmarking

The calculation line in INVAP for the CAREM Project and the validations already done are presented.

Nuclear Data

ESIN library originated from the WIMS (1976) updated with data for Ag, In, Cd and Gd from ENDF/B-4 and Nb from WIMCAL-88 (COREA). The WIMS has been in used by INVAP with validated good results in Plate Type FE (for RR) of 20% enrichment and fresh cores. The Gd isotopes were tested using the CONDOR in two numerical benchmarks with good results; however the dispersion of results given by participants are quite wide.

( ell Code C ONDOR 1.3 /9/

A preliminary validation with 91 typical PWR cells (fresh) was done /10/ and /11/, with a difference of400±800 pem Two more benchmarks /12/ and /13/ with mini FE of PWR using burnable poisons were done with satisfactory results Recently more comparisons (49 cases) were performed with results obtained for cores similar to that of the CARJEM /14/ and /15/

Core Code CITVAP

The code was validated for plate type fuel elements of 20% enrichment with good results For 90% enriched FE the results are not as good

The calculations to validate the line are underway in order to reduce the experimental works with the RA-8

Up to this stage of CAREM Project, several design aspects in the internals are recognised that need expen mental verifications The aim of these experiments is to verify the behaviour under normal and abnormal conditions and to define the manufacturing and assembling allowances as well as handling procedures and auxiliary tools Following is a general description of the arrays under construction and the foreseen experiments for each of them

A dummy of a sector of the core containing the following items:

Core support, three FE, upper structures with control rod guides The experiments will be done with water at room temperature The aim is to make fine adjustments in the design and manufacturing and the influences of the different variables in the behaviour of the assembly Also to verify the design of couplings and auxiliary tools This stage will be started by the end of the

A 1:1 in length of a Sector of the Control Rod Drive Structure (for one Control Rod) with the Connecting Rod attached to the dummy core mentioned above and a Drive Mechanism.

The experiments will be earned out in air and in water at room temperature The objectives are to obtain the manufactunng and operational allowances

Examples of the expenments to be conducted dunng this stage are definitions related with alignment, clearances in linear bearings Dynamic Analysis to determine natural frequencies and mode shapes and responses of the system under vanous external excitations

“Warm “ Experiment. T= 80 °С, atmospheric pressure. Characterisation of the mechanism and the driving water circuit at different temperatures. Study of abnormal situations: increase in drag forces; pump failure; Primary level influence; SCRAM valve failure; uncontrolled water flow and temperature; two phases water injection; suspended particles influence; air bubbles influences; drainage blockage.

“Hot” Experiment. A simple loop is under design to reach CAREM nominal operational values in normal and abnormal conditions. The objective are the characterisation of the mechanisms, durability tests, and behaviour of systems under abnormal conditions: breakage of feeding pipes; LOCA, behaviour under relief valves actuation.

The present international trend in the nuclear industry focuses on the simplification of the nuclear plants and on the reduction of the construction time. The reduced size of the most attractive modular reactors is dictated by the design target to remove the decay heat directly through the wall of the reactor vessel itself, thereby drastically reducing the number of safety-related systems.

The selected unit power of the ISIS Reactor Module (200 MWe) is consistent with this design target.

Layout studies of the ISIS power plant are in progress in ANSALDO, to optimize component arrangement and reduce erection time of the Reactor Modules and of the Balance of the Plant.

The compact reactor layout is made possible by the integrated design of the primary circuit.

When developing the SG for the VPBER-600 reactor special attention was paid to design decisions, which ensure:

1) high reliability requirements;

2) operation and repair safety;

3) possibility for inspection during fabrication, including 100% non-destructive testing of materials and welds;

4) minimum weight and dimensions, which are especially important for SGs of integral reactors;

5) maximum unification of SG units;

6) organization of parallel process flows during fabrication;

7) maximum automation of fabrication process;

8) block-section-by-block-section assembling and replacement of the SG after lifetime exhaustion, that allows a decrease in the joint diameter in reactor vessel;

9) the SG on the secondary side is made of some independent block-sections, that allows isolation of non-leaktight block-sections for both feed water and steam;

10) diagnostics during operation;

11) isolation of any non-leaktight subsection during repair;

12) possibility of using structural units that show good performance in operated SGs.;

13) guaranteed approval of SG performance by representative testing of full-scale cassettes in test facilities.

A principal requirement laid on the VVER reactors which were manufactured in the Czech Republic under the Russian licence was the one of transportability of all components by rail. The requirement was based on conditions of countries in which the units were constructed, i. e. the choice of construction sites and their availability and localization of manufacturing factories out of the reach of water-transport. Skoda Nuclear Machinery Plzeft is one of three plants in which VVER reactors were produced (besides Izhora Plant near St. Petersbourgh and ATOMMASH at Volgodonsk, both in Russia). 21 VVER-440 reactors and 3 VVER-1000 were produced at Skoda. A special railway truck KRUPP was used for the transport of large components including pressure vessels — In several. cases the railway transport was combined with road and water one. E. g.. pressure vessels for NPPs Nord in Germany and Zharnowiec in Poland, were transported by rail to the river port on the Danube near Bratislava and then by the ship to the Black Sea and around the whole Europe to ports and sites in Germany and Poland.

Way of the transport.

Railway to Bratislava, river-boat to Paks Ra і 1way

Railway to Bratislava, ship to Greifswalrl

Railway Ra і 1way

Railway to Bratislava, ship to Gdansk

Railway to Jihlava, road-truck to Bratisla­va. river-boat to Belene Railway

CONCLUSIONS

1. Pressure vessels of integral reactor of medium output have outer dimensions which are comparable with ones of BVRs. Vails of them are more thick and their weights exceed to date heaviest RPV.

2-No substantial differences are expected between existing manufacturing technologies of reactor pressure vessels of PVRs and BVRs and one of future integral reactors. Most significant there will be the size factor. Integral RPVs will be preferrably completely manufactured at shops.

2- Inspection and test techniques and devices for integral RPVs will correspond to ones currently used at PVR and BVR technology. The shop check assembly of reactor internals should be considered as convenient preparation for the on-site assembly.

3. The transport of large pressure vessels is the limiting factor of deployment of integral reactors. Integral units of a medium output should be constructed on sites in the reach of water transport. For the local generation of electricity and heat, small modular units transportable by railway or by road seem to be more prospective.

4. The introduction of integral reactors would require additional investments to productional base. mainly in increasing the loading capacity of manufacturing equipment: welding positioners, supports of machine tools, means of the shop transport etc. The scope of such investments will undoubtedly depend on the number of units required by customers. An appropriate ad-hoc solution would be certainly found for the manufacture of small number of RPVs.

EMSLAND CHOOZ В

KONVOl N4

1. World Nuclear Reactors Survey 1994, Nuclear Engineering Int.

2. Working Material: The Role of the Agency in Advanced Reactor

System Development. IAEA-TC-792.Vienna. 1992

3. Working Material-‘ Review of Advanced WWER Designs, IAEA-CT0764. Vienna. 1992

4. Working Materials for the TCM on Integral Reactors. Obninsk, the Russian Federation. 1994.

Integral water-cooled reactors make efficient use of the primary coolant inventory to prevent core damage under emergency conditions Under LOCA conditions, the steam generators of integral reactors assist decay heat removal for a longer time than in present generation designs since they remain covered by water, cool the primary fluid and reduce the loss of coolant vapor

The following are the basic systems for minimizing coolant inventory loss

• Use of a guard vessel or containment which fits closely around the lower part of the reactor vessel In LOCA this space rapidly fills with coolant ejected from the vessel and the core remains covered In some designs, e g SPWR, this space is connected to the outer water filled pressure suppression environment which gives an adequate supply of water to keep the core covered

• Provision of water by passive means from external tanks The feed may be by gravity at low pressure, gravity from tanks pressurized by automatic connection to the pressunzer in appropriate accidents, or by passive pumping devices such as steam injector pumps

The system using a guard vessel has the advantage of providing for a cheap and simple plant with no need for extra supplies of emergency coolant

The system using gravity feed tanks assisted by steam pressurization has the advantage of operation over the full pressure range The disadvantages are the limited volume of the tanks and the need for isolation valves to isolate pipe breaks outside the containment Besides, steam injectors are not yet proven for the duty required and are undergoing further development

To evaluate the characteristics of various passive safety concepts and provide the proper technical data for the conceptual design of the advanced integral reactor, the following R&D activities are being performed.

Hexagonal Semi-Tight Lattice Fuel Assembly

Numerical Analysis Technology is being development to analyze the hexagonal fuel assembly core with tight lattice. Some thermal-hydraulic experiments will be performed to understand the phenomena in the semi tight hexagonal lattice. The suitable thermal-hydraulic analytic model will be developed, especially thermal-hydraulic correlations, which are vital for the semi-tight hexagonal geometry. The developed model is incorporated into computer codes for the design and safety analysis.

No Boron Core Concept

The use of no soluble boron in the core design causes to utilize large amount of burnable absorbers to properly hold down the excess reactivity at the beginning of cycle and to install considerable number of control rods for the reactor control and operation. The optimization in the number of burnable absorbers and control rods is required with respect to the reactivity compensation with fuel bumup and reactor control through the cycle, and this study in conjuntion with the core design with hexagonal fuel assemblies are thus investigated in this R&D subject.

Natural Circulation Phenomena for Integral Reactor

To investigate thermo-hydraulic characteristics of primary circuit in natural circulation operation mode, an experimental test loop is being designed. An computer code is being developed to model the thermo-hydraulic behavior of the primary circuit.

The possibilities of contemporary reactor pressure vessel (RPV) fabrication techniques were taken into account while choosing the reactor capacity. Due to the extreme weight of RPV ingots and the technique of their fabrication, the external RPV diameter should be less than 5 m. The RPV’s height is 23500 mm, its external diameter is 4780 mm and its wall thickness is 330 mm. The electrical power of the reactor was found to be 515 MW.

The RPV is made of 15Х2МФА-А steel and the RPV flanges are made of 25ХЗМФА-А steel. The upper part of the RPV has 6 nozzles of 350 mm diameter for the steam outlet, 6 nozzles of 200 mm diameter for the feedwater inlet and other nozzles are of 150 mm diameter.

1.1. Core

The core has 121 shroudless fuel assemblies, being on a spacing of 226 mm 85 fuel assemblies have 18 control absorbing rod clusters and 36 fuel assemblies have burnable poison rods. Fuel assembly has 252 fuel rods arranged on a triangular lattice with 13.5 mm pitch. The fuel rod design is based on the WER-1000 fuel rod design. Stainless steel is expected to be used as the fuel cladding material. The fuel height is 4200 mm.

The results of the V-500 SKDI core neutron calculations are listed in the

table 1.

Main characteristics of V-500 SKDI core

Name, size ______________________________________________________________ Value

1. Core sizes (m)

— equivalent diameter 2.610

— fuel length in cold state 4.200

2. Fuel rod cladding material SS

3. Mean fuel burnup (MW day/kg U) 40

4. Mean volume power density in the core (MW/m^) 68.2

5. Makeup fuel enrichment (%) 3.5

6. Assembly number in the core 121

7. Assembly number with absorber rod 85

8. Mean fertile coefficient 0.78

The steam generator (SG) is a once-through vertical heat exchanging apparatus arranged in the annular space between the RPV and guard tube block shroud. The SG consists of 18 modules which are joined into 6 sections. Each of the sections has an individual steam header and feedwater header, inserted through the RPV nozzles. The SG modules consist of titanium alloy tubes of 10.8 m in length, 12 mm in outer diameter and 1.3 mm in wall thickness, surrounded by a stainless-steel shroud. The guard tube shroud is freely installed on the core support barrel and after removing it from the RPV there is an opportunity to repair or to remove the SG section.

The main technical parameters of V-500 SKDI are listed in Table 2.

Main characteristics of B-500SKDI

The experiments have shown that the noticeable hydrogen production due to the steam-zirconium reaction took place at the temperature about 650 °С. It was observed both visually as hydrogen bubbles in the accumulator and by record of the water level changes in the tank 9. In the teats with spiral heater where the temperature didn’t exceed 800 °С, it was marked also that the amount of hydrogen produced diminished with every consecutive turning on of the heater. It apparently may be explained by the

0 50 100 150 200 250 300 350

t, сек

Fig. 2. Change of the water volume in accumulator tank and simulator surface temperature vs. time.

— — TestN. l: —- TestN. 2.

influence of the zirconium oxide layer at the rod surface on the steam access to the clean ziconium surface and, finally on the hydrogen production. At the test with the rod simulator, however, at the temperature of 1000 °С there was a local break-off in oxide film. It may be resulted from both by the rod distorsion and by the hydrodynamic exerting of the steam-droplet mixture on its surface.

In Fig. 2 the results of two tests earned out at the identical initial conditions with the rod IIPE simulator are shown. The electric power in both tests was 1.5 kW. Fig.2 illustrates evidently the dynamics of the assembly dryout and the hydrogen discharge during steam-zirconium reaction. The curves V(i) and T(i) are correlated with each other qwite well. So, the signal from the level gauge being initially constant begins to increase with the begiiming of the steam-zirconium reaction. It’s the moment when the rod temperature reaches its maximum value. At the meam temperature of 1000 °С the amount of the hydrogen produced during 170 s was as much as 143 cm3 in the test No. 1 and 110 an3 in the test No.2. The considerable rod deformation over the length was marked.

In [3] the measuring of the amount of hydrogen produced in the steam-zirconium reaction was carried out at the temperature of 1000 °С. The measuring technique was in determination the mass increase of small zirconium specimens in steam atmosphere during the definite time and consecutive calculation the amount of hydrogen. The mean value of the hydrogen volume produced by 1 cm3 of the specimen surface per 1 s was about 0.0136 cm3. In our tests this value was as much as 0.0125 cm3. The agreement is quite satisfactory. It may be consider as a confirmation of the reliability of the dilatometric method used for the measuring the amount of hydrogen produced.

CONCLUSONS

Experimentally it has been shown that the dilatometric method of the amount of hydrogen produced during steam-zirconium reaction is a perspective one for carefull investigation of this reaction kynetics.

The visualization of the process of the one-rod assembly dry-out has been carried out.

REFERENCES

1. Kovalevitch О. M., Budaev M. A. The Problem of Hydrogen Production at APP Accidents. Afomnaja technikaza rubezhom, 1952, N. 12, c. 25.

2. Fujishiro, T. Hydrogen Generation during Cladding/Coolant Interactions under Reactivity Initiated Accident Conditions. Report Presented at the International Conference NURETH-4, Karlsruhe, 1992. p. 23.

3. Levin A. Ya., Izrailevsky L. B., Morozov A. M. Investigation of Interaction of Zirconium with Steam at 1000 °С. Thermohydraulic Processes on APP Facilities, Proceedings ofVTI, Moscow, Eneigoatomizdat, 1986.

LIU KUKUI, LI MANCHANG, TANG CHUANBAO Nuclear Power Institute of China,

Chengdu, Sichuan Province,

China

Abstract

In China, heat supply consumes more than 70 percent of the primary energy resource, which makes for heavy traffic and transportation and produces a lot of polluting materials such as NOz, SOz and C02 because of use of the fossil fuel. The utilization of nuclear power into the heat and electricity co-generation plant contributes to the global environmental protection.

The basic concept of the nuclear system is an integral type reactor with three circuits. The primary circuit equipment is enclosed in and linked up directly with reactor vessel. The third circuit produces steam for heat and electricity supply. This paper presents basic requirements, reactor type choice, design characteristics, economy for a nuclear co-generation plant and its future application.

The choice of the main parameters and the main technological process is the key problem of the nuclear plant design. To make this paper clearer, take for example a double-reactor plant of 450 x 2MW thermal power. There are two sorts of main technological processes. One is a water-water-steam process. Another is water-steam-steam process. Compared the two sorts, the design which adopted the water-water-steam technological process has much more advantage. The system is simplified, the operation reliability is increased, the primary pressure reduces a lot, the temperature difference between the secondary and the third circuits becomes larger, so the size and capacity of the main components will be smaller, the scale and the cost of the building will be cut down.’ In this design, the secondary circuit pressure is the highest among that of the three circuits. So the primary circuit radioactivity can not leak into the third circuit in case of accidents.

keywords

small-sized nuclear co-generation, integral, self-pressurized, forced circulation, water-water-steam technological process

1 Introduction

With the development of human society, more and more heat supply is needed. The heat supply consumes more than 70 percent primary energy resource in China, while the electric power supply consumes only about 20 percent of the resource. It not only uses up about more than 4 x 10s tons coal every year, but also burns a great quantity of oil, which makes for heavy traffic and transportation , and causes environmental pollution and unnecessary resource waste. It is a good idea to construct a batch of nuclear co-generation plants near to the cities where the steam and heat supply is centralized in the future. This is a new way to save energy, relax traffic and transportation and reduce environment pollution.

The development of nuclear co-generation plant is a possile of prospective economical way, but how to ensure that the small-sized plant clean, safe and cheap is the first problem to be considered by the designers. Based on previous design experience, the plant design objectives are put forward as follows:

(1) Low basic capital investment The specific cost of the plant should be less than 750 $/KW;

(2) Construction period: A single-reactor plant needs about 5 years and a double-reactor plant needs about 6 years;

(3) Operational reliability: The availability is up to about 85%-90% and the load factor is more than 80%;

(4) No large pipe break accidents (all the plant pipe diameters are less than 100mm) and no LOCA in the primary circuit The reactor melt probability is less than 10’7/reactor-year. The reactor can be cooled using its own resources and the storage battery sources when normal onsite and offsite electric power is lost The plant can be built in a usely-populated area because of the reactor’s passive safety and high reliability;

(5) Being simplified in system, compacted in layout and of small constructive scale;

(6) Because the high and large reactor building is cancelled and the foundation loading is lightened, the plant can be built in seabeach, soft soil, seismic areas and so on. On the plant site choice, it is similar to the fossil-fueled power plant;

(7) The heated steam for heat supply has no radioactivity and its radiation level almost equals that of natural radiological reference state;

(8) The reactor building is pressurized and airtight The environment around the plant would not be polluted in case of radioactive leakage.