The decommissioning schedule is similar to that of existing PWR and BWR units and may include the following stages

• fuel assembly unloading,

• dismantling and breaking up of in-vessel highly radioactive structures,

• dismantling and breaking up of the RPV,

• dismantling of low radioactive and non radioactive equipment,

• radioactive equipment and spent fuel removal from plant site

Alternatively, the relatively large reactor vessel can be used for storage of all active components for about 50 years until their activity has reduced to a low level for easy handling for ultimate disposal This option is especially attractive if the vessel is below ground allowing removal of many structures above ground

4 CONCLUSIONS

In the technical committee meetings, a diversity of viewpoints and opinions was offered on the design and development of integral reactor concepts Consensus exists on many aspects of integral reactors The following are the conclusions

6 1 The presentations confirm the engineering validity and sound advancement of the integral design approach for advanced light water reactors (ALWR) Presentations have also covered new issues, areas and designs not covered previously

4 2 Integral reactor design activity is strong in many Member States Some designs have been built, some are in the detailed-engineering stage and most are in the conceptual design stage

6 3 For further development, a clear definition of user requirements is necessary, which will clarify design criteria and specifications

7 4 There are many similar designs for which realization in construction is unlikely Concentration of effort on fewer projects for the detailed design stage would be beneficial and cost effective

7 5 Integral designs cover from low to medium power range Low power reactors generally use natural circulation These designs may be more applicable than loop type designs for district heating and in remote locations

7 6 International cooperation is strongly recommended in carrying out further development of this type of reactors The efforts should include computer code validation and simulation of accidents in integral reactors, noting the specific problems of modeling natural circulation and the effect of non-condensables in accident condition

7 7 Other areas that should receive special attention include

a) Economic comparison of different integral reactor systems and identification of benefits covering the whole of the life cycle.

b) Ways to maximize the safety of integral reactors, especially to enable their siting near population centers

c) Requirements for plants in remote areas where more stringent restrictions on operation may be needed.

d) Decommissioning of integral reactors

e) Design of compact steam generators

f) Survey of market potential for integral reactors.

As showing :n Fig 1 the reactor core is located at the bottom of the reactor pressure vessel (RPV) 6 primary heat exchangers divided mto two groups are arranged on the penpnery m upper part of the RPV The system pressure is maintained by N2 and steam All penetrations of RPV are therefore of small size (the largest one sized D50) and located at the upper part of RPV No pipe goes mside and down to the lower part of the RPV so that should a pipe break occurs outside the RPV the water ejection wall last not so long, then turn to steam ejection This arrangement reduces the amount of losmg inventory to a great extent Meanwhile a guard vessel fits tightly around the RPV so that the core will not become uncovered under any break of pnmarv pressure boundary even at the bottom of RPV This vessel can also function as a containment

The reactor vessel is cylindrical, with a weld hemispherical bottom head and a removable, flanged and gasketed hemispherical upper head The RPV is 4 8m in diameter, 14m in height and 197 Tons m weight The cylindrical portion of the vessel is welded by 65 mm steel plate then lmed with 6 mm stainless steel layer The RPV will be manufactured and fabricated in factory then shipped to the site The guard vessel consists of a cylindrical portion with a diameter of 5m and an upper cone portion with maximum 7m in diameter The guard vessel is 15 lm in height and 223 Tons m weight The upper and lower portions are fabricated m factory then shipped to the site separately, then welded together on site

3.1 Safety concepts and safety systems

A large LOCA cannot occur in the MRX, since only small size pipes (^50mm) exist in the primary system. The emergency water injection systems are not provided in the MRX. The engineered safety system consists of the water-filled containment vessel (CV) system, the emergency decay heat removal system (EDRS) and the containment water cooling system (CWCS). During a small LOCA, the engineered safety system keeps the core flooding and removes the decay heat without emergency water injection. Figure 2 shows a basic idea of the engineered safety system. The decay heat in a LOCA is transferred to the atmosphere by natural convection of the primary water in the EDRS, the CV water and the CWCS working fluid. According to the PSA, this engineered safety system has a high reliability. The probability of the functionally disordered trouble is 2xl0-e at first one month after starting the operation of this system. The residual heat removal system (RHRS) is not necessarily essential in the emergency core cooling system of the MRX. The RHRS is used for controlling the temperatures of water in the RPV and the CV for long term cooling after a LOCA and is also used for long term decay heat removal during a scheduled reactor shutdown.

(1) Containment vessel (CV)

The design pressure of the water-filled CV is 4MPa to withstand a high pressure at LOCA. The size of the CV is designed to satisfy the requirements of radiation shielding and the maintenance of instruments installed in the CV. The RPV is surrounded with a watertight shell for thermal insulation. The shell can stand against the static pressure at anticipated transients. Pressure relief valves are installed to mitigate the rise of pressure in the space between the RPV and the watertight shell due to pipe breakage accidents in this area.

(2) Emergency decay heat removal system (EDRS)

The EDRS is a closed system which transfers decay heat from the core to the CV water. It includes three trains, each of which has a capacity to remove the core decay heat. Each train consists of a hydrogen reservoir tank, a cooler, two valves and piping. In the case of accidents, the valves of each train are opened actively by the signal of battery, then the primary coolant circulates by natural convection removing decay heat from the core and is cooled in the cooler placed in the CV water.

(3) Containment water cooling system (CWCS)

The CWCS is a heat pipe system for long term decay heat removal transferring the heat in the CV water to the atmosphere. It includes four trains. In the event of an accident, the water temperature in the CV will be kept lower than the design value by the arbitrary three trains operated using natural convection. For its working gas in the CWCS, anti-freezing gas such as

R22 (CHCIF2) will be used taking into account of low temperature condition in ice-sea atmosphere.

From the above experimental study the following conclusions can be made.

(1) The total drain water quantity during the safety valve discharge is no more than 20 % of the initial water inventory, no matter whether the safety valve is full open or not. ( this is confirmed by the heat balance) .

The drain velocity and flow rate can be calculated by the critical flow theory.

(2) For the discharge caused by breaking of a pipe from the steam plenum, the drain quantity depends on the position of the pipe in the steam volume. During the discharge process water may be drained discontinuously. The bigger the pipe diameter, the more the discharged quantity. The total drain water quantity is more than for the opening of the safety valve at the same discharge orifice diameter.

(з) For the discharge from the water volume, the drain quantity is more than that of a drain from the steam volume. All water above the pipe opening will be drained. When the level is lower than the opening of the pipe, the drain quantity depends on the height difference between the liquid level and the opening of the pipe. In a certain range the bigger the height difference, the smaller the drain quantity. Under the conditions of our tests, when the height difference is more than 5cm, its effect on the quantity can be negligible.

References

1. Ma Changwen, Bo Jinhai, Za Meisheng and Others:

“ Safety experimental study for the Nuclear Heating Reactor ” INET Report, 1991, 2.

2. Ja Haijun

Safety experimental study on the Boron injection system of the 5MW Heating Reactor, doctoral thesis. 1991, INET, Tsinghua Univ.

Nuclear power has played a significant role in the supply of electricity over the past two decades. The two major nuclear accidents, namely Three Mile Island and Chernobyl, have considerably affected its further growth. Reconsideration of reactor design and safety aspects of nuclear power has been an active area of development. Elimination of some accident scenarios, simplification of systems and reliance on natural phenomena have been a major part of these activities, which led to the development of new design concepts.

Under the sub-programme on non-electrical applications of advanced reactors, the International Atomic Energy Agency has been providing a worldwide forum for exchange of information on integral reactor concepts. Two Technical Committee meetings were held in 1994 and 1995 on the subject where state-of-the-art developments were presented. Efforts are continuing for the development of advanced nuclear reactors of both evolutionary and innovative design, for electricity, co-generation and heat applications. While single purpose reactors for electricity generation may require small and medium sizes under certain conditions, reactors for heat applications and co-generation would be necessarily in the small and medium range and need to be located closer to the load centres.

The integral design approach to the development of advanced light water reactors has received special attention over the past few years. Several designs are in the detailed design stage, some are under construction, one prototype is in operation. A need has been felt for guidance on a number of issues, ranging from design objectives to the assessment methodology needed to show how integral designs can meet these objectives, and also to identify their advantages and problem areas.

The technical document addresses the current status of the design, safety and operational issues of integral reactors and recommends areas for future development.

H. J. BOADO MAGAN, J. P. ORDONEZ INVAP S. E.,

San Carlos de Bariloche

A. HEY CNEA,

Buenos Aires

Argentina

Abstract

The CAREM Project is a iow power NPP of 25 Mwe, with an integrated self pressurized primary system The cooling of the primary system is of the natural circulation type and several passive safety systems are included The owner of the Project is Argentina s CNEA (Comision Nacional de Energia Atomica) and its associated company, INVAP, is the main contractor

The present status of the CAREM Project is presented The possible evolution of the CAREM project is mentioned in relation with a new containment design A short description of the Experimental Facilities, listed below, already in operation and under construction are also included

• CAPCN High Pressure Loop Natural convection loop to verily dynamic response and critical heat dux

• RA-8 Critical Facility, designed and constructed for the CAREM Project (that may be used as a general uses facility)

• RPV Internals The whole assembly of absorbent rods, connecting rods and the rode guides are being constructed in a 1 1 scale The aims of this experimental facility are vibration analysis and manufacturing parameters definitions

• Control Drive Mechanisms A sene of venficalion and tests are being earned out on these within RPV hydraulically dnven mechanisms

Other development activities are mentioned in relation with the tliermalhydrauhcs, Steam Generators and Control 1. CAREM Project Present status.

The Primary System of the ISIS reactor is of the integrated type (fig. 2), with the Steam Generator Unit (SGU) housed in the Reactor Vessel, to which feedwater and steam piping are connected.

Within the Reactor Vessel, an Inner Vessel provided with wet metallic insulation separates the circulating low-boron primary water from the surrounding highly borated cold water.

Hot and cold plena are hydraulically connected at the bottom and at the top of the Inner Vessel by means of open-ended tube bundles, referred to in the following as Lower and Upper Density Locks. The Inner Vessel houses the Core, the Steam Generator Unit and the Primary Pumps.

Outstanding feature is the complete immersion of the Pressure Boundary, made up, for each module, of a Reactor Vessel and of a separated Pressurizer with interconnecting Pipe Ducts, in a large pool of cold water.

During normal operation, the heat generated in the core is transferred to the SGU via the water circulated by the Primary Pumps, which are located at the top of the Inner Vessel. In case of unavailability of this heat transfer route, the cold and highly borated water of the Intermediate Plenum enters the Primary Circuit from the bottom, mixes up with the hot primary water, shuts down the reactor and cools the core in natural circulation. The same process, by heating the intermediate plenum water and the Pressure Boundary metal, activates the natural heat transfer route towards the Reactor Pool, which contains approximately 6.000 cubic meters of cold water.

Fig. 2 — ISIS Reactor Module

The water inventory in the Reactor Pool is large enough to allow the water itself to remain below the boiling point after removal of the decay heat for about a week.

Cooling down of the plant pool is guaranteed, anyway, for an unlimited time, by virtue of two loops provided with water-air heat exchangers in natural circulation, sized to reject to the atmosphere, at steady state, approximately 2 MW and thereby capable to prevent the pool water from boiling.

Similarly to the PIUS reactor concept, the shut down and cooling functions of the core are carried out, in any condition, by the highly borated cold water of a plenum, which is hydraulically connected to the primary system by means of density locks.

However, unlike the PIUS, the intermediate plenum of ISIS contains a relatively small inventory of cold water (approximately 300 cubic meters per reactor module) at primary system pressure.

While breakage of the MM vessel, and the link between the MM and the pressurizer remains intakt, the MM is provided with water from the pressurizer. However, with a large break of the MM vessel, the flow rate of the loss of coolant exceeds the flow rate of the supply; therefore the pressure and the amount of the coolant in the MM decreases until the amounts of losses and supplied liquid are equal

To study such the situation, the MM vessel was provided with pipe connections of which the breaks at different elevations were simulated (i. e 5190 mm and 1140 mm above the upper edge of the fuel-assembly) All experiments were conducted at an initial power of 1070 kW reducing in accordance with the residual heat expected.

It is experimentally established that if the disrupture of the MM vessel occurs at these elevations m the initial phase of the accident, the heat removal from the fuel assembly is effected in crisis-free conditions with the fuel assembly filled with water Therefore the subsequent cooling of the fuel assembly is beyond question even if only the feed is present in these cases. The contrary is the case when the MM vessel ruptures below the fuel assembly where the MM is incapable to retain the supplied water and the flow rate entering it from the pressurizer is insufficient to provide a continuous flow over the entire fuel assembly cross-section. It is possible under these conditions that water will not flow around a portion of the fuel elements, which will lead to their overheating.

The experiments were conducted as follows: A preset flow rate of water ot room temperature (15- 20OC) was fed to the upper part of the micromodule, then the power of a predetermined value was supplied to the fuel assembly and the micromodule vessel; the fuel element temperature being recorded at four points over the fuel assembly cross-section and at ten points of the entire elevation. Considerable attention was paid to the investigation ot the influence of the feed water conditions (e. g. below or above the "flow-over" wiridoiqs of the heat exchanger; from one or two sides of the MM vessel) Tables 3 and 4 present some results of these experiments It is obvious from Table 3 that there is a significant scatter in tuel assembly temperature (65-400°C) This shows a non-uniform distribution of inlet water over the tuel assembly cross-section. Of great interest is the fact associated with reduction of the fuel element surface temperature as the power increases from 10 to 15 kW in the present case. To our mind, this is due to the boiling-up of water on those simulators where water is available, and its more uniform distribution over the fuel assembly cross-section At a certain stage, further increase in power logically results in a rise ot the rod temperature

With increasing the flow rate of water being ted to the MM low tuel assembly temperatures are observed at sufficiently high powers (Table 4) In the RKM-150 reactor, the parameters of water flowing around the secondary circuit and the net MM power (together, with the heat influx from graphite, amounting to — 40 kW) meet the conditions ot the test the results, ot which are shown in the second column ot temperatures, Table 4. However in the RKM-150 reactor the flow rate of water flowing to the MM from the pressurizer is significantly higher (no less than 4000 kg/hr) than in the above mentioned case (700 kg/hr). Thus, the reliable cooling of the fuel elements without their overheating will be provided in each phase of accidents accompanied by the MM vessel rupture at any elevation.

In addition to the above consideration, experiments were conducted, simulating accidents with MM vessel rupture at the elevated pressure and a power level of ~ 600 kW. These experiments confirmed the absence of the fuel element overheating under conditions ot this accident even at flow rates of feeding water considerably ‘ower than in the RKM-150 MM •

Table 3

THE THERMOCOUPLE INDICATIONS AT WATER CIRCULATION ROUND THE MM AT A FLOW RATE OF 100 KG/HR AND ONE-SIDED FEED.

(The MM vessel is not heated and the water flow rate over the secondary circuit is equal to zero)

4.3.2. Secondary Circuit Zero Flowrate,

Experimental study has been carried out of an accident with ceasing the second circuit water flow rate through the heat exchanger of one MM, which can be caused by blocking the flow area by an outside subject. The. initial parameters of the micromodule were set up to correspond — to its operation at a maximum design-based power of 1070 kW. After terminating the flow rate of the secondary circuit water, the MM power reduced according to the law of residual heat variation with a delay of 10 s.

Experiments showed that in such an accident, the pressure over the primary circuit slightly exceeds the nominal value (by 0.2 MPa) during a short period of time 15s) The circulation over the primary circuit provides a crisis-free cooling for the fuel assembly As for the whole of the reactor, the pressure in the emergency MM reduces after decreasing the power The duration. of the experiment was 10 mm By that time about 32 kg of water retained in the MM whereas to fully cover the fuel assembly, 18 kg is sufficient (Note, that according

Table 4.

FUEL ELEMENT TEMPERATURE AT WATER CIRCULATION AROUND THE MM WITH A FLOW RATE OF 720 KG/HR (ONE-SIDED FEED.

(The water flow rate in the secondary circuit is 11 i/iir and inlet temperature is — 65 C)

to calculations, the time required to decrease the water amount in the MM to 18 kg is 13 min). Thus the experiment reasonably well justified the predicted time of starting the fuel assembly uncover, which is assumed in the fuel element temperature behaviour evaluations.

J. KRALOVEC

Skoda Nuclear Machinery Plzen,

Plzen, Czech Republic

Abstract

Integral water-cooled reactors are typical with eliminating large-diameter primary pipes and placing primary components, і.e. steam generators and pressurlzers in reactor vessels. This arrangement leads to reactor pressure vessels of large dimensions: diameters. heights and thick walls’and subsequently to great weights. Thus, even medium power units have pressure vessels which are on the very limit of present manufacturing capabi1 іties. Principal manufacturing and inspection operations as well as pertinent equipment are concerned: welding, cladding, heat treatment. machining. shop-handling. non-destructive testing, hydraulic pressure tests etc.

The transport of such a large and heavy component makes a problem which effects its design as well as the selection of the plant site. Railway, road and ship are possible ways of transport. each of them having its advantages and limitations.

Specific features and limits of the manufacture and transport of large pressure vessels are discussed in the paper.

INTRODUCTION

Though the Czech Republic at present time does not intend to develop and construct integral reactors, a contribution on problems of manufacturing and transportation of large pressure vessels which is based on the experience from the manufacture of VVER reactors can be inspiring. In addition to that. the situation in the Czech Republic can be representative to other countries which could potentially utilize integral reactors. The situation is typical with following features:

— inland position.

— complicated terrain.

— geological situation limiting the choice of construction sites.

— lack of water transport ways.

— complicated network of railways and roads having limited transporting profiles as well as limited bearing capacities.

In addition to that, heavy machinery factories which are qualified to produce nuclear components are located out of the direct reach of water-transport.

Integral water-cooled reactors are typical by the elimination of large — diameter primary components i. e. steam generators and pressurizers in reactor vessels. This arrangement leads to reactor pressure vessels of large dimensions: dimeters, heights and thick walls and subsequently to great weights. Thus, even medium-size power reactors have pressure vessels vhich are on the limit of present manufacturing capabilities.

There are 424 reactors in operation in the world, from which are 239 PVRs. 89 BVRs. 2 PHVRs (with a pressure wessel). Under construction there are 48 PVRs and 6 BWRs. for which pressure vessels have been mostly manufactured. Totally. 384 RPVs have been manufactured.

Not ice

Materials will be used which have for PVRs and properties of which are uniformity of chemical composition technological properties. even for very and extremely heavy forgings.

Following materials should be considered:

Ni-Мо-Сг type: SA 508 C1.2. 22NiMoCr37. SQV2A.

Mn-Mo-Cr type: SA 508 C1.3. 16MND5. SA 533 Gr. B.

Cr-Mo< N і)type; 15Ch2MFA,15Ch2MNFA.

and corresponding Japan steels. eventually improved mod і f і cat і ons-

The technology of casting extremely large ingots Cup to 600t) from these steels have been mastered by leading firms assuring the required homogen іty of chemical composition and high purlty.

Most integral reactors being considered have some type of control rods and also have soluble boron This provides diversity in the physical means of shut-down and in the technical means of implementing it

System-integrated PWR (SPWR) has no control rods and relies on boron systems only There are three independent systems Philosophically, this is similar to the fast breeder reactor (FBR) situation where there are only control rods for shut­down but there are diverse control rod drive mechanisms