In the NHR-5 some pipes are connected under the water level, such as the boron injection system pipe, the pipe for the hydraulic driving facility for control rods etc. A pipe break of this type is more dangerous. In the case all the water above the inlet of the pipe will be discharged as well as some water lower than the inlet because of siphonage.

A discharge test has been made to model a pipe breaking in the boron injection system. The boron injection pipe is opened under water at a level equivalent to the middle of the chimney.

In the test steam content, drain water quantity, liquid level, pressure and their transient characters have been measured.

The typical pressure change during the test is shown in Fig. 5. The change of drain water quantity is shown in Fig. 6.

From these figures it can be seen that at the beginning of the drain process discreasing of the pressure wasn’t very fast, but increasing of the drain water quantity was quite considerable. This can be understood, because at the beginning water was discharged, vaporized steam compensated for the change of water volume, so the pressure changed slowly. When the liquid level was lower than th. e opening of the discharge pipe, steam and gas began to drain. The pressure decreased faster, but the drain water quantity changed reasonably slowly. The decrease of the liquid level from linear changed to slower. Eventually draining stopped because of losing heat and pressure.

When the liquid level was lower than the opening of the pipe, some water was drained because of siphonage. The greater the distance between the pipe opening and the liquid level, the less the drain water quantity. Consequently there was a smaller the decrease of the level.

The relationship between the drain water quantity and the height (to the liquid level) of the opening is shown in Fig. 7. From this can be seen that when the height is more than 5cm, the height almost doesn’t effect on the drain water quantity.

(1) Adopting integrated self-pressurized water reactor.

(2) The primary circuit operational state is: the primary coolant is at low pressure; the reactor core has a low power density; the fuel in low temperature; while the coolant has a high average temperature. So not only the safety but also the economical steam is ensured. The third circuit steam pressure can reach 4MPa and thermal efficiency can come up to about 80% while the primary pressure is only 10 MPa.

(3) Adopting a water-water-steam technological process improves the operation reliability of the heat exchangers in the reactor. The middle circuit pressure is higher than the primary’s, so radioactivity leakage is prevented from the primary to the third. The size of the heat exchangers and related buildings is reduced.

(4) The reactor and the primary circuit have adoped an adopt integrated layout The inner diameter of the containment is only 16m, so the reactor building is reduced to a minimum and the nuclear island covers 2518m2, only 32% that of looped layout

(5) Large amount of concrete and steel will be reduced and the general cost will also be reduced because of the small constructive scale. And a half of million tons oil will be replaced every year. So the time to recover capital is estimated as about 7 years.

(6) High passive safety and reliability. About 20% thermal power can be removed by the natural circulation of coolant when the main pumps stop. So a core melt accident will not occur when the main pump motor electricity is lost It is almost impossibly difficult for radioactivity to leak into the third circuit because of the high pressure in the middle circuit The accident probability drops down because the systems are simplified and the number of valves and pumps also decreased. The water-water heat exchanger is more reliable than a steam generator in the reactor. All the above-mentioned can reduce unexpected plant shutdowns and improve the plant availability.

(7) The plant can be built near a city or a large enterprise because of the reactor’s high passive safety and two-layered containment.

(8) The plant layout is compact and the main and the auxiliary buildings take up less land, and the foundation loading is lightened. So it is favourable for the plant to be built on seabeach, soft soil or a seismic area.

(9) The main components of the reactor and the primary circuit can be manufactured and installed in the factory. On site installation and examination work become less. The nuclear island constructive scale is small. So the double-reactor constructive period can be shortened to 5 years.

(10) The shape of the main and the auxiliary buildings on the nuclear island is simple, it is easy to construct So the construction period is shortened.

Table 2 is the main performance comparison between the integrated and the looped pressurized water reactor. The parameters in table 2 indicate that this design meets the basic requirements of a nuclear co-generation plant

According to the above-mentioned, the small-sized nuclear co-generation plant is economically competitive. Referring to more than twenty users’ steam consumption, it is suitable to adopt 400-600MWt power for the single-reactor. For the nuclear co-generation plant, its main function is heat supply, while the amount of electricity supplied depends on how much the heat is supplied. To avoid causing a high turbine cost, the electric power proportion should not be too large. If the heat supply is the main function and the electricity supply is auxiliary, the three circuits and water-water-steam technological process should be adopted.

For the reactor choice, PWR and BWR have all been considered. But the former is put first to raise the reactor core temperature, produce medium pressure steam in the third circuit and expand the scope of steam supply. The reactor and the primary circuit adopt an integrated layout and forced circulation. It is more compact than BWR with natural circulation. It is more suitable that the operation pressure of the self-pressurized water reactor adopts lOMPa according to the steam parameter requirements of domestic chemical industry.

In accordance with the above discussion, for a small-sized nuclear co-generation plant, conclusions are made as follows:

(1) Adopt integrated, forced circulation, self-pressure-stabalized water reactor.

(2) Reactor operation pressure lOMPa.

(3) Single-reactor power 400-600MWt, maximum to 1000KWt.

(4) Adopt water-water-steam main technological process and electricity supply depends on steam supply.

(5) No boric acid and simplified system.

This design also suits a small nuclear electric power plant But the heat exchangers should be replaced by steam generators in the reactor vessel.

The design can also be used for centralized heat supply in a city. On condition of not changing the reactor and the auxiliary system, only decreasing operation pressure and temperature, and the steam generator being replaced by the heat exchangers in the third circuit, this design would turn to be nuclear heat supply plant, while the research and experiment are not needed furthermore.

Natural circulation and forced circulation should be compared further in the future and the control rod drive mechanism type needs improving. A core with a high conversion ratio in a small reactor is worth researching further.

In summary, a small nuclear heat and electricity co-generation plant could become an economic, safe and clean energy source in a city or a large enterprise in the future. Its basic capital investment is low, the construction period is short, it has a wide range of uses, it has good prospects and it should have a proper position in the energy resource development of China.

L. CINOTTI Nuclear Division,

NSALDO,

Genoa, Italy

Abstract

ANSALDO has conceived a reactor called ISIS (Inherently Safe Immersed System), an innovative light water reactor with easily understandable safety characteristics.

The main targets are: passively safe

behaviour, no pressurization of the Reactor Containment under any accident condition, control of plant capital cost and construction schedule by virtue of the modular concept and the compact layout.

The ISIS concept, described in general terms in the paper, builds up on the Density Lock concept originally proposed by ABB ATOM for the PIUS plant (ref. /If), featuring innovative ideas derived from ANSALDO experience and based on proven technology from both LWR and LMR.

1. MAIN TARGETS OF THE ISIS CONCEIT

1.1 Safety targets

Significant progress has been made in the last years towards nuclear reactors that rely to the smallest possible extent on safety-related active systems, which, even using up-to-date technology, are felt by the public as prone-to-fail, no matter how low the frequency target for their loss is set.

The ISIS concept, under development in ANSALDO, largely embodies this progress. The main safety targets may be summarized as follows:

— No core melt-down and negligible release of radioactivity in any accident condition, by virtue of the reactor concept itself.

— Prompt reactor shut down occurring naturally after any abnormal condition.

— Reactor cooling in natural circulation for unlimited time.

— Self-depressurization of the reactor after a postulated failure of the pressure boundary.

1.2 Economic targets

The economic target aims at a viable industrial power plant based on specific overnight-capital cost and construction time competitive with those of the Light-Water Reactors under development This is achievable by means of following features:

— Modular reactor.

— Integrated components (Compact layout).

— No pressurization to be taken into account in

the design of the reactor containment

— Primary system installation after reactor

building completion.

While studying accident situations, the major attention was focused on accidents associated with the depressurization of primary circuit components and the disturbance of heat removal over the secondary circuit.

4.3.1.1.Rupture of the pipeline between MM and pressurizer

Consider the behaviour of the MM channel in the case of the disrupture of the pipeline connecting the micromodule with the pressurizer. The peculiarity of this situation is that it is impossible to feed water to the MM from the pressurizer. The calculations obtained for this accident did not predict severe consequences; however, this fact needed an experimental verification. The problems to be studied were as follows:

— The MM depressurization results in ceasing or reversing of the circulation in the MM or degrading the heat removal from fuel elements in the first phase of the accident;

— the amount of water remaining in the MM compared to the amount at the moment the pressure reduces up to the atmospheric value in it;

— the maximum permissible power of the fuel assembly filled with water under the flooding conditions (zero flowrate);

— the rate of coolant losses at the final accident phase at a pressure near to the atmospheric.

Fig.4 shows the results of one of the experiments investigating the break-down of the pipeline connecting the MM with the pressurizer. The rupture is located just in the vicinity (100 mm) of the MM vessel and is simulated using a fast-responce device. The initial power of the channel amounted to 1070 kW and then reduced according to the law of residual heat variation. The delay time was accepted to be 10 s. which is considerably heigher than the design-basis value (i. e below 1 s).

The experiments indicated that the pressure in the MM reduces to the atmospheric pressure for 1-2 min. The water flow rate through the fuel assembly slitly increases at the beginning of the process (to — 1.9 kg/s) and then for 2-2.5 min is practically equal to the nominal one (1.5 kg/s), providing the reliable cooling of fuel elements. No negative outcomes were found in the temperature behaviour of fuel elements over this time period.

The amount of water remained in the MM to the moment of reducing the pressure to the atmospheric value amounted to 54 kg., the water level being ~ 2.6 m above the upper edge of the fuel assembly.

To define the limiting fuel assembly power under sarbotage conditions occurring after reducing the pressure, some preliminary tests were carried out using a 7-rod bundle with the inlet cross-section blocked, (i. e. under more severe conditions as compared to the MM). They revealed, that for a 30-rod fuel assembly MM the limiting power should be 120-130 kW, thich is higher than the net power of the residual heat in fuel elements and the heat influx from graphite (for the present 100 kW design). In the full scale MM without blockage at the fuel assembly inlet (i. e. for real conditions) at the atmospheric pressure, a power of 150 kW was reached without any indications of an> critical phenomena. Thus, the condition of the fuel assembly filled with water is sufficient to remove the residual heat under no-crisis conditions.

After the MM pressure is reduced to the atmospheric value, the further development of the process is characterized by the ongoing loss of coolant. The presence of the heat exchanger in the MM results in that the overwhelming portion of steam generated in the fuel assembly condenses on the surface of this heat exchanger and returns through the downcomming line of the circuit to the fuel assembly inlet. As a result, the rate of steam losses through the rupture is considerably lower than the rate of steam generation in the fuel

assembly. The measurement of the rate of mass losses in this phase of the accident is taken at different values of the fuel assembly power, and different levels of water in the MM. The obtained data allowed the time interval to be defined, over which the fuel assembly has reliable cooling, and the fuel element cladding temperature is close to the saturation temperature of water at the atmospheric pressure This time interval ranges from 36 to 44 hrs, and only then the fuel element heat-up will occur, but no higher than 610 C m accordance with the calculations

The reactor is refueled annually. Refueling is manual. Half of the core is replaced in every refueling. Spent fuel is placed in a pool. The pool has capacity for accommodating the spent fuel produced by seven years of reactor operation. After seven years, the older spent fuel elements will be removed to a dry storage.

Waste handling facilities are similar to the ones for conventional PWR. Solid waste management must be tailored to the site and country characteristics.

4.4 Decommissioning

No specific features related to decommissioning of integrated reactors has been identified. Actually, dismantling of the plant will probably be easier than the corresponding to a conventional PWR: dismantling of the primary system is achieved by just removing the pressure vessel.

3. CONCLUSIONS

The paper gives and overview of the status of CAREM design. The description makes emphasis in those topics that can imply important differences with respect to traditional PWR technologies.

Although some areas in which R&D activities are necessary have been identified, the problems specific to integrated type reactors are well within the present technological capabilities of the industry.

The following aspects need attention during the design phase:

• The integral water-cooled reactor complicates the in-vessel control detector arrangement. There is a need for new technical solutions regarding instrumentation design. There is no difference with other reactors in principle but there may be engineering problems.

• In-core instrumentation is needed on prototype reactors but should not be required on production reactors. This is because of the low power density and small size which gives stability to the power distribution

• There is a licensing requirement for in-core instrumentation in some countries

• Failed fuel detection can be left to detectors in the chemical and volume control system (CVCS) as in large light water reactors (LWRs)

The once-through steam generator is located within the reactor vessel in the annular space between the core support barrel and the reactor vessel inner wall. The secondary coolant is completely evaporated in a single pass through the steam generator. The steam generator tubes are divided into two groups that can be operated independently and the tubes in each group will be cross wound to reduce thermal twisting.

The steam generator consists of tube bundle, downcommer, two feed water and steam headers, shrouds to guide primary flow, and tube supporting structures. Approximately 1100 tubes are helically coiled in the effective heat transfer region and the effective coiling height is 3m. To assure equal steam quality from individual tubes, the length of tubes in the effective tube bundle is maintained as nearly equal as possible by varying the number of tube starts and helix angles in each tube column. The tube material is inconel 690 with 19 mm outer diameter through the whole tube length. The tube bundle is supported by eight perforated radial support plates, which transfer the load to the bottom support structure located on the supporting lug. Each tube can be accessed easily through the feed water and steam headers which are attached to the reactor vessel for in-service inspection and maintenance.

In accordance with modem NPP safety approaches, radiation exposure on personnel, population and environment in normal operation and design-basis accidents should not lead to excess dose rates for people, and in beyond design-basis accidents, this effect should reasonably limited. To this end, technical and organisational measures are taken to ensure safety with any initiating event envisaged by the design with superposition of one failure independent of the initiating event of any of the following safety system element: active or passive element with mechanical movable parts or one personnel error independent of the initiating event. Besides one failure independent of the initiating event, it is necessary to take into account the nondetectable failure of elements affecting safe operation, which are uncontrolled in operation, and influencing accident propagation.

Safety of the UNITHERM NPP is achieved by a complex of technical solutions among which the following are worth mentioning.

The NPP employs water-moderated, water-cooled reactor with inherent safety features which reflect its capability of keeping safety on the basis of internal feedbacks, natural physical processes applying passive residual heat removal systems and automatic protection devices which ensure chain fission reaction suppresion without intervention of operator. The UNITHERM NPP is also capable of self-controlling chain fission reaction due to negative temperature, power and void coefficients of reactivity. The core physical characteristics are so selected that the above coefficients are negative in the entire range of temperatures during the core life both in normal and emergency operation modes. This eliminates spontaneous core power excursion in normal startup and heatup and stabilizes operation in steady-state and transient conditions when heat consumer circuit modes of operation change.

After NSSS is started up and brought to a preset load, moving of shim groups upwards is mechanically blocked thereby eliminating possibility of unauthorized introduction of additional positive reactivity.

The NSSS design is such that all potential leak initiation locations are in the top part of the vessel with limiting equivalent leak diameter sufficiently small and not exceeding DN 20. The integral layout of NSSS unit with rather efficient iron-water radiation shield between the core and wall of NSSS vessel excludes vessel brittle fracture because of metal neutron irradiation. All this allowed to exclude accidents associated with large and medium leaks, and prevent dangerous propagation of accidents due to core dryout. To this end, the containment (Fig. 3) is used designed for localization of primary leaks within the inner volume. The use of three-section liquid

Fig.3. Reactor Plant

1 — Iron-water shielding tank; 2 — radioactive gases storage cylinders; 3 — liquid absorber supply system; 4 — containment; 5 — shock-proof casing; 6 — cooldown system heat exchanger; 7 — safeguard housing; 8 — steam generating unit; 9 — biological shielding blocks; 10 — liquid and solid wastes storage tanks; 11 — basement

absorber feed system ensures flooding this part of the containment in the emergency situation under consideration with liquid medium up to the level above potential primary circuit depressurization points which completely eliminates core drying in any design initiating events or accident scenarios. Thus, as a maximum design-basis accident it is possible to consider the primary circuit leak of conventional equivalent size of DN 20 max. The estimates showed that propagation of such an accident follows the scenario typical for the small leaks in containment without deterioration of its leaktightness and core damage. This is contributed by reasonable emergency cooling core system (ECCS) redundancy and its passive principle of operation using no forced circulation means. ‘

Incorporation into the UNITHERM NPP of additional localizing safety barrier — safeguard housing — enables even in the case of beyond design-basis accidents due to containment depressurization practicale to eliminate radioactivity release to the environment and risk of core drying. Beyond design-basis accident due to containment depressurization and postulated damage of 10 % of fuel elements and primary coolant and radionuclides discharge to the safeguard housing does not lead to significant radiation damage for population and individual exposure will not exceed 0.11 rem/y.

The UNITHERM NPP three-loop thermal-hydraulic design when consumers even with two consecutive interloop leaks can be reliably protected by reasonably redundant shut-off and cut-off localizing valves against radionuclides discharge to heat consumer circuit, and against harmful effect of ionizing radiation on personnel. Thanks to this, NPP personnel is beyond the area of ionizing radiation and the ionizing radiation rate on the NPP protection surfaces does not exceed the background values in normal operation. The dose rate of ionizing radiation in maximum design-basis accident 100 m away from NSSS is only by 10% above the background level.

Special attention has been paid to NPP UNITHERM emergency core cooling system (ECCS) which plays the important role in safety assurance. This was mostly due to the fact that because of inapplicability of traditional technical solutions we had to search for new ones taking into account not only general approach to NPP design, but climatic conditions of the NPP site as well.

The ECCS is designed as independent process loop associated with the intermediate loop. In emergency situations, the heat removed from the core via steam generators arranged within the NSSS vessel is fed to the intermediate loop, and further, through heat exchangers of the loop, is removed, via independent ECCS, to its heat exchange surfaces cooled by atmospheric air. The low winter temperature level in UNITHERM NPP application areas demands the selection of low-boiling coolant of the aforementioned independent loop of ECCS. To this end ammonia may be used.

Specific features of ECCS is that it does not have isolating and cut-off devices, i. e., the system is in continuous operation. Therefore, marked seasonal ambient air temperature fluctuations may greatly influence the amount of heat discharged through the system to atmosphere. So, in summer the capacity of heat removed through the system is 3-4 % of the nominal NPP heat capacity, the respective figures in winter may increase as high as 1.5-2 times. To reduce heat losses, the system of shudder-type gate valves is envisaged in the air duct. Switching of the gate valves from summer to winter operation is made during NPP preventive maintenance. Apart from its main functions, ECCS provides the possibility to keep NPP in hot stanby, i. e., at minimum possible core power level when power take-off is stopped.

Another improvement in reliability and safety of the UNITHERM NPP is the passive nature of core protection system. During NPP operation under load variations, core power is self — controlled, whereas variation of reactivity during continuous operation practically compensated for by burnable absorber and temperature effect, and only once a year reactivity is adjusted by remote relocation of absorber rods.

Emergency scram of NSSS and the core subcriticality is achieved by insertion of absorber rods in the core by motor-operated drives or by gravity and compressed spring energy in case of de-energizing of drives. Shutdown of NSSS with malfunction of the above absorber rods is ensured by using additional emergency protection based on alternative design philosophy. To prevent unauthorized withdrawal of control and protection system elements in commissioning the electromagnetic "arrestors" are provided in the drives limiting movement of absorber rods.

For quantitative evaluation of the UNITHERM NPP safety the following possible scenarios of severe accidents have been considered (unauthorized introduction of positive reactivity in the core, loss of preferred power, and primary circuit depressurizationjand probabilities of final states of the following categories have been determined:

• the first category — accident localized without violation of safety limits;

• the second category — accident localized with partial deviations from safety limits and without core damage;

• the third category — accident localized with significant deviations from safety limits and accompanied by transition to core steam cooling which in the case of long-term accident can lead to partial core damage.

Predictions showed that core damage probability in any of the above-mentioned accident situations will not exceed 10’5 1/y. In this case, probability of core damage with primary circuit

—12 * —X

depressurization is 6.7 • 10 , and with blackout is 5.4 • 10 1/y.

Now, cassette SGs are operated as a part of transport VVER reactor plants. More than 212000 straight-tube steam generating elements operate as part of cassette SGs.

Total operating experience of all active steam generators is more that 500000 hours. Simulating a standard operation model in the laboratory verified lifetime of more than 100000 hours.

Cassette SGs have never failed during operation.

2. CONCLUSION

Positive experience of theoretical and experimental investigations, accumulated in OKBM, the results of cassette steam generator operations in operating PWRs, wide layout potentialities of cassette steam generators in combination with high specific characteristics (compactness and specific power) allowed development of integral type reactor plants, of various powers (including 600 MW electric power) and purposes.

Because the activity and mass of radioactive equipment of the integral reactors considered are substantially lower than those ones of conventional NPP reactors, this influences irradiation doses.

According to calculations personnel irradiation dose during VPBER-600 and AST-500 unit decommissioning is 2.5-3.5 man. Sv taking account of plant-level systems).

The radiological effect on the inhabitants during decommissioning was evaluated proceeding from the amount of radioactive product releases which can enter the atmosphere. The result obtained was that during AST-500 equipment and reactor vessel disassembly the reactivity release to the atmosphere including cleaning in wet scrubbers and aerosol filters gives the value ~3 mCi for Co 60, that is -20% of the specified limit.

6. COMPARISON OF DECOMMISSIONING CONDITIONS FOR INTEGRAL AND

VVER-TYPE REACTORS

The advantages of integral reactor decommissioning emerge when comparing decommissioning conditions for VPBER-600 and VVER-440 "Lovisa" (Fig.9).

The concept of "immediate" dismounting in 2 years following termination of preparatory work is suggested both for "Lovisa” and VPBER-600 reactors. The main problem of NPP "Lovisa" decommissioning is the handling of the hot reactor vessel which must be broken up in hot chambers by remotely controlled equipment and be removed in large thick-walled casks to stores or the reactor vessel and equipment must be removed and disposed off as a whole. One should note that special manipulators and other expensive facilities are not required for breaking up the VPBER-600 reactor vessel.

Comparison results show that in spite of higher power a more prolonged operation period and shorter hold-up time following shutdown, safety characteristics (activity value, radwaste amount, irradiation doses) during VPBER-600 reactor decommissioning are tens and hundreds times smaller than those for VVER-440 reactor.

7. CONCLUSION

So, the structural features of AST and VPBER reactors (including the integral layout of their main primary equipment and location of a leaktighr reactor in an additional guard vessel) ultimately simplify the technologically difficult, radiologically dangerous and expensive decommissioning of PWR (VVER)-type NPP reduce the amount of radwaste and the costs and make integral reactor decommissioning safe for both personnel and inhabitants.

[1] Dimensions were taken from available literature. Heights and weights of vessels were determined approximately after subtracting the ones of covers.

[2] Dimensions and weights have been determined from Г21. 13].[4] by approximation after subtracting covers.

From the comparison of pressure vessels of PVRs. BVRs and those of integral reactors it follows, that diameters and lengths of last ones are comparable with those of BVRs. Due to the higher operating pressure which is equal to operating pressure of PVRs. walls of integral reactor vessels are more thick (in the range of 250 — 300 mm) and their weights exceed the weight of the to date heaviest RPV of ATUCHA 2 (975 t).