R. HÜPER Fast Breeder Project, Kernforschungszentrum Karlsruhe

STATUS OF FAST BREEDER REACTOR DEVELOPMENT 500 °C. After several years of operation with a thermal reactor core (KNK I), IN THE FEDERAL REPUBLIC OF GERMANY, the plant has been run with fast, i.e. unmoderated, reactor cores since 1978 (KNK II). BELGIUM AND THE NETHERLANDS February 1987* While the fuel assemblies of the first fast core, KNK II/l, were identical with those of the Mk.Ia first core of the SNR 300 in their key parameters, the fuel assemblies of the second core loading, KNK 11/2 , correspond to the SNR 300 Mk.II R. HÜPER type, which is characterized by a high fuel density, 7.6 mm fuel rod diameter, Fast Breeder Project, and spark eroded spacer grids which, in the fuel assemblies of the test zone, Kernforschungszentrum Karlsruhe GmbH, are attached to the wrapper tube by means of flow aprons. Karlsruhe, Federal Republic of Germany The first fast core had been unloaded in 1982 after a peak burnup of 100.000 MWd/te. The fuel assemblies of the second core, which was commissioned iri mid-1983, have meanwhile reached an average burnup of 47.000 MWd/te. One test Abstract fuel assembly transferred from the first to the second core has reached an average burnup of 125.000 MWd/te by late 1986. In 1967 and 1968, the Federal Republic of Germany, the Kingdom of Belgium, and the Kingdom of the Netherlands ("DeBeNe") agreed in a joint program to develop In the course of the burnup of the second core so far, a change in the outlet breeder reactors. The following research organizations have taken part in this effort: temperatures of the Mk.II fuel assemblies has been observed. Probably as a result of some redistribution of the coolant flow from the test zone to the driver zone due to increased pressure loss in all fuel assemblies, the outlet Kernforschungszentrum Karlsruhe (KfK), INTERATOM, Bergisch Gladbach, temperatures of the test zone fuel assemblies rise, while those of the driver ALKEM, Hanau, assemblies drop. An extensive program of research is being conducted to eluci­ date the causes of these temperature changes. SCK/CEN, Mol, Belgonucléaire, Brussels, Four fuel assembly failures have so far occurred in operation of the KNK 11/2 core. Two test assemblies and two driver assemblies were involved. Operation of ECN, Petten, three of the assemblies was continued up to 35 days at full load and part load, TNO, Apeldoorn, respectively, after the failure had been noticed. Spotting the failed fuel NERATOOM, The Hague. assemblies in KNK II is done by a combination of load tilting operation and a dry sipping test conducted by the fueling machine. The failed fuel assemblies are subjected to extensive post-irradiation examinations at the Hot Cells of The three institutions mentioned above have been associated since 1977 in the KfK. Entwicklungsgemeinschaft Schneller Brüter. KfK, INTERATOM, and the French Commissariat à l'Energie Atomique entered into contracts in 1977 about close cooperation in the fast breeder field, to which the Belgian and Dutch partners Irradiation and post-irradiation examinations of specimens of the 1.6770 type material of the KNK II reactor vessel, which had been loaded in the first acceded. materials test assembly, have been concluded. The integrity of the vessel wall has been found to be unimpaired by irradiation effects, even in the long term. The results of activities carried out by the DeBeNe partners in 1986 have been compiled in this report. The report begins with a survey of the fast reactor plants, followed by an R&D summary. In an additional chapter, a survey is given Post-irradiation examination of a KNK 11/2 absorber assembly has also been of international cooperation in 1986. completed. The absorber rods of the first core were filled with boron carbide granulate. Only some test rods contained B.C pellets of the kind used for KNK II follow-on loadings and for the SNR 300.

I. SUMMARY An investigation is being carried out of the KNK II steam generator in order to see whether the noise of the water-sodium interaction, which would arise in a steam generator failure, can be detected. For this purpose, a leak simulation 1. Operating Experience of KNK II in Karlsruhe source and ten acoustic sensors have been installed.

KNK, a compact sodium-cooled reactor, is an experimental nuclear power station of 20 megawatts electric power, whose core is cooled by liquid sodium at approx. 2. Commissioning of the Kalkar Nuclear Power Station (SNR 300)

In 1973, construction near Kalkar was begun of the 300 MWe SNR 300 prototype * Translated by R. Friese. breeder power plant by Internationale Natrium-Brutreaktor Bau GmbH (INB). The plant owner is Schnellbrüter-Kernkraftwerksgesellschaft mbH (SBK). Both firms 3. Planning of the SNR 2 are joint German-Belgian-Netherlands enterprises; in addition, the state-owned British CEGB holds an interest in SBK. The company awarding the contracts for planning and construction of the SNR 2 (1.500 MWe pool type facility) will be the Europaische Schnellbrüter-Kernkraft- Construction of the plant was largely completed in 1985 after lengthy delays, werksgesellschaft (ESK). Its shareholders are SBK, the Italian ENEL, and the most of them caused by political conditions. After the sodium systems had been French EdF. The project is based on the Breeder Convention agreed upon in 1973, filled, functional tests were carried out within the framework of pre-nuclear under which construction of the SPX 1 in France, now commissioned, was to come commissioning. A number of problems were overcome in this phase in 1986: first.

Because of a number of leaks observed, the welds of the four dump tanks and Planning work for the SNR 2 is to provide an assured development status re­ leakage interception vessels on the secondary side were redone. As a consequence presenting an attractive basis on which to decide about the next European of similar technological, design and welding conditions as in the sodium dump breeder. Points of primary interest are engineered safeguards (passive safety and leakage interception vessels, also the partly loaded cold traps were re­ concept) and economic considerations. paired and regenerated. On the basis of the well-known inherent properties of breeder facilities, the To reject the higher-than-rated heat load, new blowers were installed in the safety potential is to be exploited further, e.g., by passive natural cir­ reactor closure head cavity. Additional tanks designed to collect leakage oil culation systems for decay heat removal. Studies of the component design va­ were installed near the primary pumps. riants of decay heat removal systems performed so far, and also the arrangements of these components, are at present being evaluated in detail by the French and Labor-intensive rerouting of cables took until mid-February 1986; it had become Italian partners. Also the British side contributes to this effort. necessary in order to meet the tighter redundancy requirements in the decay heat removal trains. Passive day heat removal is one of the key features of an accident management concept designed to minimize active measures. The studies devoted to this In January 1986, a vibration measurement lance introduced into the reactor complex of problems are intended to result in a new kind of emergency power vessel was ruptured in an attempt to dismantle it for reworking. The disrupted supply concept and a consistent control panel concept. part of the lance and several small parts which had stuck in the reactor vessel were retrieved. In March 1986, the primary system and the reactor vessel were The concept for handling core assemblies has been advanced far enough to allow again filled with sodium. consistent handling under sodium of irradiated core assemblies and, hence, solutions for passive decay heat removal to be defined. Conceptual proposals An abnormally high moisture content was measured in the cover gas of the re­ have been made for the layouts of the secondary systems and the handling de­ actor. It came from the basalt granulate used to shield the reactor closure vices. head. The basalt was dehumidified at a sodium temperature of about 240 °C. Breeder plants achieve clearly higher burnups than light water reactors. One of The tests of the emergency power supply system were completed. The systems the targets in planning for the SNR 2 ist a considerably higher burnup also designed for decay heat removal (decay heat removal specifically for each leg relative to the SNR 300. Besides this point, which is important especially with through the heat sink in the water/steam system and the emergency core cooling regard to fuel cycle costs, lower construction costs are expected to arise from system) were tested successfully. the passive safety concept.

Since the end of 1986, the Kalkar facility has been ready for acceptance of the The SNR 2 allows cost reductions to be achieved especially for the following fuel assemblies. reasons :

In July 1986, the North-Rhine-Westphalian State Minister of Economics, who is No safety related decay heat removal through the "balance-of-plant" sy­ responsible for issuing reactor permits, surprised everyone with the announce­ stems; consequently, conventional components and process engineering ment of reservations against further permits, because of his new assessment of simplifications are possible. facts well-known for quite a time. The German Advisory Committees on Reactor Safeguards (RSK) and Radiological Protection (SSK) as well as the Rhineland Absence of extensive emergency power supply installations because the Technical Inspectorate (TQV) had not raised any objections on grounds of tech­ sodium systems can be used for passive decay heat removal. nical safety against loading the fuel assemblies and conducting zero power tests in the SNR 300. Reduction in size of buildings and systems to be protected against earth­ quakes and airplane crashes by arranging inside the reactor building all The Federal Government expects the permit for storage of fuel elements to be safety related systems, including the accident control panel. issued in the first half of 1987. The R&D program for further development of the SNR has been revised in the light The German-French breeder agreements of 1977 have resulted in close cooperation of the R&D programs of the foreign partners. This R&D program was subdivided with France. Under an agreement signed in London on March 2, 1984, industries into work packages by the joint Working Groups with the British, French, and and research institutions in Germany, the United Kingdom, France, Italy and Italian partners so as to achieve broad international division of labor. Belgium have combined their breeder activities. The possibility to accede to the agreement has been left open to the Netherlands partners. The agreement had been preceded by a Memorandum of Understanding among the governments of the partici­ 4. Research and Development Work pating countries dated January 10, 1984. An R&D agreement was initialed in November 1984. R&D work was concentrated on fuel and materials development, safety, physics, and components development. Part III of this report contains the most important In the spring of 1986, INB/Interatom, Ansaldo, and Novatome signed an agreement R&D carried out in 1986. They can be summed up as follows. on industrial cooperation in evaluating the passive safety concept proposed for the SNR 2. The purpose of the agreement is to reach a harmonized proposal for Materials for fuel, blanket, and absorber assemblies were studied and further the design concept of the next European breeder. developed. One major objective of development is an extension of the in-pile time of components in order to reduce the operating costs of breeder plants. The British manufacturing industry cooperated intensively in the further deve­ lopment and optimization of the SNR 2 in 1986. Some major points of attention R&D work on reactor physics above all focuses on critical experiments conducted were the support of the reactor closure head, the conceptual and technical to determine the nuclear parameters of breeder cores, on evaluation of the design of the steam generators, core structures, and decay heat removal. reactor physics data accumulated in the commissioning of breeders, the prepara­ tion of improved nuclear data and joint European reactor physics codes for The DeBeNe group and France so far have mainly cultivated trilateral relations nuclear core design, and contributions to design basis and safety calculations. each with Japan and the USA, which have resulted in an information exchange, the execution of joint projects, and the delegation, among others, of German experts One major aspect considered in safety studies is the complex of fuel rod fail­ to foreign research teams in a number of areas of breeder research. For the ure, coolant failure, and power transients as potential causes of accidents. future, the partners will seek to achieve joint agreements for cooperation in Studies performed to control credible accident consequences deal with ensuring breeder R&D between Western Europe and Japan, and the USA, respectively. decay heat removal from fuel melts, among other aspects. The behavior of aero­ sols, effects of sodium fires and sodium-concrete interactions, constitute Through the International Working Group on Fast Reactors (IWGFR) of IAEA a another major point of' research under this heading. To meet technical safety limited exchange of knowledge and experience has been organized with other criteria, e.g to ensure decay heat removal after failure of the tertiary system, countries. International conferences support the development of breeders world­ more use is to be made of the inherent safety advantages of sodium cooled wide, especially in the safety sector. systems.

Methods of measurement in plant surveillance are important in the preventive II. REACTOR PROJECTS measures taken against failures in the reactor core, the cooling systems and their components. The data supplied by the measuring instruments are processed, in the future by means of pattern recognition techniques, to enable accident 1. Operation of the KNK II Experimental Breeder Reactor conditions to be diagnosed. An in-core monitoring system of this type is to be demonstrated in KNK II. 1.1 Reactor Operation

Materials of the vessel and of other components are being investigated es­ The first fast core, KNK 11/ 1, of the compact sodium cooled nuclear reactor was pecially for their long-term characteristics (structural integrity), in parti­ unloaded in 1982 after a peak burnup of 100.000 MWd/te. The fuel assemblies of cular in the light of the impact of sodium and of irradiation effects. Fracture the second core, KNK 11/2 , which had been commissioned in mid-1983, have mean­ mechanics methods are employed to detect the failure modes of vessels and pipes while reached an average burnup of 47.000 MWd/te. A test fuel assembly trans­ arid to achieve acceptance of the leak-before-break criterion for future licen­ ferred from the first to the second core had attained an average burnup of sing procedures. 125.000 MWd/te by late 1986. This is close to the irradiation dose level of 68 dpa NRT, which is typical of the SNR 300.

5. International Cooperation While the KNK 11/1 fuel assemblies were identical with those of the Mk.Ia first core of the SNR 300 in their main parameters, the KNK 11/2 assemblies correspond Ever since 1967, Belgium and the Netherlands have cooperated closely with the to those of the Mk.II type of the SNR 300, which are characterized by a high Federal Republic of Germany within the framework of the R&D Programs Working fuel density, a fuel rod diameter of 7.6 mm, and spark eroded spacer grids Party. which, in the fuel assemblies of the test zone, are attached to the wrapper tube by means of flow aprons. In the course of the burnup of the second core so far a change has been observed particular importance in repair is the simple, reliable identification of failed in the outlet temperatures of the Mk.II fuel assemblies. Probably as a result of fuel rods. One fuel assembly of the third core has been equipped with Hydra a coolant flow redistribution from the test zone to the driver zone due to an bellows indicators as a means of testing one of the techniques of potential use increased loss of pressure in all fuel assemblies, the outlet temperatures of for this purpose. As is seen in Fig. 1, these bellows are attached to the bottom the fuel assemblies of the test zone rise, while those of the driver assemblies end plug within the fuel rod. The fission product gas evolving in the fuel rod drop. During reactor outages the temperatures occasionally return to their during irradiation compresses the bellows, which causes a ferrite core inside original levels, but it has not yet been possible to attribute this to a specif­ the bellows to change position. In a failed fuel rod, the fission product gas ic event. will escape from the rod still in the reactor; the bellows will return to their original shape. The position of the ferrite core consequently is different in a Four fuel assembly failures have so far occurred in operation of the second failed fuel rod than in'an intact one. The position of the ferrite core can be core. Two test fuel assemblies and two driver fuel assemblies were involved. determined in a hot cell by means of a coil measurement system even without Operation of three of the assemblies was continued for another 35 days at full disassembly of the fuel rod bundle. load and partial load, respectively, after the failures had been detected.

The operation in 1986 of KNK II was determined largely by investigations carried out to clarify the causes of temperature changes and fuel assembly failure. Most of these investigations were conducted with the nuclear part of the reactor shut down.

In view of the planned extension of in-pile time of the second core, a public inquiry was held in May 1986 on the objections raised by the public. The expert appraisal by the licensing authority of the documents submitted on the extension of the in-pile time has been completed, the Baden Technical Inspectorate (TÜV) having finished its work. Serina LanЛгЛ GeniUioa JeUows lotto* End ТЦ Most of the fuel rods and all of the structural parts are ready for the third loading of KNK II. However, the fuel assemblies still need to be made. The FPg as TV«s5art ia ^ design basis reports have been submitted for expert appraisal to the Baden TOV. Irta-ivcttí^ InbutftyL No FP Got ’Vftssui’t In 1.2 Pinpointing and Repairing Failed Fuel Assemblies IrrfcAnibA FalU In KNK II, failed fuel assemblies are detected by a combination of load tilting operation and a dry sipping test conducted by means of the fueling machine. Both Fig. 1 : Detection of failed fuel rods techniques have proved to be reliable; merely in looking for the fuel assembly failed last, the prediction made by load tilting operation did not agree with by the ferrite core technique. the assembly later detected by means of the sipping test.

The fuel assembly check system using the wet sipping method, which had been planned for the SNR 300, had not met expectations in KNK II in 1985. However, in 1.3 In-plant and Post-irradiation Examinations of Fuel Assemblies those trials it was found that simply lifting the fuel assemblies individually by means of the reshuffling device, which allows them to stay in the coolant, Model calculations performed on the observed temperature changes seemed to will release fission product gas from a failed fuel rod. If sufficient sodium indicate that these could be explained to be due to partial blockages of the flows through fuel assembly, that gas will reach the cover gas in the upper coolant flow in the region of the spacer grids of the fuel assemblies caused by plenum, where it can be detected. A device using this principle of measurement particles of a certain size and in sufficient quantities. However, experiments is at present under construction. performed with mesh inserts in the main coolant flow with mesh sizes of 11 pm showed that, despite the presence in sodium of particles of the assumed size, This fuel check system is another instance in which the importance of KNK II their quantity was in no way sufficient. The bulk of the undissolved impurities within the framework of the German Fast Breeder Development Program was demon­ is made up of very fine sludge of particle sizes below 1 pm. strated. In post-irradiation examination of the failed fuel assemblies a coating was As a rule, the failure of assembly means that only one out of more than 100 fuel found on the surface of the upper region of the fuel rods, especially in the rods has become defective. In view of the high fabrication costs it therefore test assembly, the nature and cause of which has not yet been found out.Para­ appears to be useful, especially for fuel assemblies with low burnups, to meter studies have shown that the temperature changes could well have been develop a concept of repairing fast breeder fuel assemblies. An aspect of caused by a thin coating. Two of the four failed fuel assemblies have so far been subjected to post­ assembly, has been completed. The integrity of the vessel wall is not impaired irradiation examinations in the hot cells. Test assembly NY 308 revealed a by irradiation effects, not even in the long term, as was found out. failed rod with a wide longitudinal crack caused by the formation of uranate- plutonate. Compared to experience in the first core, there were very pronounced Irradiated specimen capsules containing the Np237, Am241, Th232, Pu239, and U233 wear marks at the points of contact between the spark eroded spacer grids and actinides have been removed from the second materials test assembly. The speci­ the fuel rods. In addition, marks were observed which seem to indicate that mens are still being evaluated. Irradiation was carried out to determine the there had been contacts among the fuel rods. The driver fuel assembly examined integral cross section and also to support work on the design of the fuel cycle. subsequently showed comparable findings. Several failed fuel rods had transverse Other actinide specimens are still contained in one fuel assembly of the second cracks in the spacer region. core.

1.4 KNK II Experimental Program Also post-irradiation examination of a KNK 11 /1 absorber assembly has been completed. The absorber rods of the first core had been filled with boron An extensive irradiation and experimental program is being conducted in KNK II. carbide granulate. Only some test rods contained B^C pellets of the type used All experiments mounted in the facility are fully covered by the provisions of for the follow-on loadings and for the SNR 300. the German Atomic Energy Act, i.e., a complete licensing procedure must be carried out on any modification to the facilities or the mode of operation While TV cameras and endoscopes are used for inspecting components below the covered by a permit, which might be required for the experiments. water level in water-cooled reactors, different methods of inspection must be used for the non-transparent fluid sodium at temperatures of about 200 °C. Among Irradiation of installations described earlier was continued in the core of other devices, an ultrasonic viewing device under sodium has been developed, KNK II also in 1986: which is to be tested in KNK II. The image is produced by means of the pulse Pressure tube test device. echo technique, which is also employed in oceanography. A multiple US probe Test tubes in the fuel assembly structure. scans in concentric circles the object to be imaged, transmitting pulse type Test bundle with carbide fuel. signals and receiving echos from the component under inspection. These echo Two materials test assemblies with structural material specimens and signals are evaluated in a microcomputer system with respect to transit time and actinide specimens. amplitude and then plotted point by point by a plotter as a function of the Materials irradiation plugs. position of the multiple US probe. In order to achieve sufficient resolution of Depressurized absorber rods. the image, the ultrasound needs to be focused. Fig. 2 shows the focusing probe, which consists of a central transceiver and eight receiving sensors around it. For the TOAST (Tolerance Expansion Study) in-pile experiment planned for the The pulse echo technique is used also in one of the newly developed sodium third core, the test fuel rods and the structural parts have been fabricated in filling level probes. Fig. 3 shows the installation of an ultrasonic probe in the meantime, both for the test bundles and for the carrier fuel assembly. The the bottom part of the vessel, the filling level of which is to be monitored. design basis report for these assemblies has been drafted and harmonized with The filling level is determined from the transit time of the acoustic signal. In the reports on the overall core. order to compensate for the influence of sodium temperature on the velocity of sound, also the transit time to a transceiver installed at a known distance is The fabrication data of all test fuel assemblies of the third core loading, such included in the evaluation in addition to the echo received from the liquid as fuel composition, numbering of mixed, sintered, and cladding tube batches, level. densities, and weights per unit length have been processed and entered into the BESEX data base system. These data are now available for post-irradiation A KNK II steam generator is being used for studies of the possibility to detect examinations. The system has proved to function satisfactorily in work on fuel the noise produced by the water-sodium interaction associated with steam ge­ assemblies of the second core. nerator damage. For this purpose, a leak simulation source and ten acoustic sensors have been installed. The source simulating the leak consists of a steel Comparison of the designs of the fuel assemblies of the three KNK II cores vessel filled with water. Nitrogen at high pressure (200 bar) is introduced into indicates that not only more and more aspects of fabrication technology were the water. The measuring systems were tested outside KNK II and coupled to the considered, but also possibilities were taken into account of disassembing spent steam generator in 1986. Some preliminary measured results show that the cou­ fuel assemblies. Moreover, as methods of localizing failed fuel rods within a pling of the units to the steam generator pipes still needs to be improved. fuel rod bundle are being tested within the KNK II experimental program, a repairable fuel assembly has now been planned for the third core. In this case, A project study was carried out to develop the concept of a diagnostic system it must be ensured for the exchange of a failed fuel rod that the top and bottom for monitoring the steady-state operation of a large fast breeder reactor. The sections of the fuel assembly can be separated in a hot cell and then rein­ system is designed to recognize in time incipient failures which could lead to stalled. Items requiring further investigation are problems of quality assurance fuel failures and also to propose countermeasures. In addition, a scram analysis and corrosion. is to be conducted which determines, out of the flood of signals, that signal for more frequent selected cases which was primarily instrumental in initiating Irradiation and post-irradiation examination of specimens of the 1.6770 reactor a scram. This monitoring system must also cover part of the heat transfer vessel material of KNK II, which had been used in the first materials test system, malfunctioning of which can have a direct impact on the core. The oo feasibility of such a system is to be demonstrated in KNK II. The measuring and evaluation systems already existing at that plant, plus additional ones, will be used for this purpose.

2. Construction of the Kalkar Nuclear Power Station (SNR 300)

2.1 Assembly and Pre-nuclear Commissioning

The 300 MWe breeder prototype nuclear power station, SNR 300, has been under construction at Kalkar on the Lower Rhine river by the Internationale Natrium- Brutreaktor Bau GmbH (INB) since 1973. The owner is the Schnellbrliter-Kernkraft- werksgesellschaft mbH (SBK). Both firms are joint German-Belgian-Netherlands enterprises; also the state operated British CEGB holds an interest in SBK.

Prenuclear commissioning of the Kalkar nuclear power station is nearly com­ pleted. The rerouting of cables for redundancy and fire protection purposes (Fig. 5 ) has been completed. All repair work on the sodium dump tanks and leakage interception vessels of the primary and secondary sodium systems has been carried out. The vessels have been back in regular use since February 1986. Because of similarities in the technological, design and welding conditions with the sodium dump tanks and leakage interception vessels, also the cold traps, which had already been partly loaded, were designed for repair and regeneration.

Fig. 3: Fi.a-.ii ---- F i q. 5” : Focusing ultrasonic probe with a Operating principle Component hall of the SNR 300 reactor building (above the primary central transceiver and eight of the ultrasonic pumps and the intermediate heat exchanger cavities) with auxilia receiving sensors. level probe. support structures for final fire protection activities. On January 7, 1986, a vibration measurement lance, which had been inserted in "Spots" detected on the outer wall of the reactor vessel by means of optical the reactor vessel, was broken in an attempt to retrieve it for reworking. The inspection devices were explained by analyses and comparison with the surface of broken part of the lance and several small components, which had remained in the a primary pump to be uncritical phenomena not causing any detriment to a com­ vessel, were collected from the yessel after the sodium had been drained. On ponent. The "spots", the contrast of which had been grossly exaggerated by the March 10, 1986 the primary system and the vessel were refilled with sodium. optical means of inspection (TV camera), are phenomena occurring in high tem­ perature operation on technically pure surfaces of Cr-Ni steel grades. A final As a result of findings made in the prenuclear commissioning phase, also modi­ report of the investigation has been compiled and submitted to the authorities fications of the plain bearings for the pipes of the sodium auxiliary systems and expert consultants. became necessary. The work required for this purpose has been completed. In line with the progress in assembly and commissioning activities, the number The design of the ventilation system of the reactor instrumentation rooms has of personnel was reduced from approx. 1750 to 440 between early 1986 and late been modified and the appropriate technical modifications have been made. Also August 1986. The advance completion of some residual work in the fields of changes to the cooling system of the fresh fuel assembly storage facility have mechanical and electrical engineering systems and fire protection caused a been completed. Most of the in-service inspection lances have been completed and personnel increase to approx. 750 people by the end of 1986. delivered to Kalkar. 2.2 Licensing and Supervisory Procedures under the Atomic Energy Act The plant has been ready technically for acceptance of the fuel assemblies since late 1986. At its meeting in February 1986, the German Advisory Committee on Reactor Safeguards gave a positive recommendation on the first step of nuclear commis­ Approximately 97 % of the items required for prenuclear commissioning of the sioning of the SNR 300 (loading and zero power tests). power plant have been completed. Among other activities, handling of the core assemblies under sodium was tested, and the core assemblies containing no In July 1986, however, the North-Rhine-Westphalian State Minister of Economics, nuclear fuel (reflector, absorber and dummy assemblies) were loaded in the who is responsible for reactor permits, surprisingly announced reservations reactor vessel. Most of the commissioning activities implied the repetition of against further permits for the fast breeder in Kalkar. tests necessary as a consequence of the modifications made to some systems. Some first tests still need to be carried out in the sectors of handling, control of In the reasons given for this decision, the following items, among others, were the decay heat removal systems, and the protection systems. mentioned as being in need of clarification: Bethe-Tait accident, quality checks and in-service inspections, spots on the outside of the reactor vessel, sodium In testing the emergency core cooling system, satisfactory agreement with the deposits at the ferritic/austenitic mixed welding seam of the vessel, and planning data was found with regard to effectiveness and behavior of the system. safeguarding the back end of the fuel cycle. Also the dump test of the secondary sodium system was conducted successfully. Tests of the recirculation mode in operating the reactor building ventilation In a statement of August 15, 1986 by SBK/INB, the objections raised by the system were completed and the trial phase of operation of the water-steam system authority were refuted. A discussion on principles held with the authority on was finished. Backfitting is required in the tertiary system with respect to August 26, 1986 did not show any more profound arguments and reasons, for these control possibilities in the decay heat removal mode. objections.

After some malfunctions had been detected in the linkage of the primary and On the basis of studies conducted by KfK and Interatom, the Bethe-Tait accident secondary shutdown systems within the framework of prenuclear commissioning had been the subject of an extensive expertise as early as in 1982. On the basis tests, a systematic analysis of possible causes of defects was conducted, which of that work, decision 7/5 of September 22, 1982 by the authority had found that revealed that residual moisture had been contained in the basalt boxes of the the design basis limit defined for. the mechanical energy released covered, to rotating reactor plug system and in the immovable closure head ring, which was the necessary extent, the spectrum of physically meaningful accident sequences released through the bores and gaps, respectively, existing vis-à-vis the cover compatible with the technical systems properties of the reactor facility. Since gas plenum. The same phenomenon is the reason for the high plugging point in the that permit had been issued, breeder research on the development of uncontrolled primary system, which has been found since the change in the cover gas of the power excursions has advanced further. The main results were presented at reactor vessel. To remove the residual moisture (on the order of approx. 1 % bx conferences on breeder safety in Lyons in 1982 and Knoxville in 1985. They are weight) contained in the basalt, a device was installed which operated largely also reflected in the analysis of the Clinch River Breeder Reactor carried out by bypassing the cover gas and the reactor vessel and, hence, the reaction with by US experts on behalf of the Nuclear Regulatory Commission, which were pub- the primary sodium. For this purpose, tubes were introduced through existing 1ished in March 1984. openings in about 15 positions of the closure head system so as to end a short distance below the basalt boxes in the cover gas plenum. Through these tubes the Because of a requirement made by the authorities, the German Federal Minister hot argon continuously filled through the closure head gap system was removed; for Research and Technology acting in coordination with the German Federal the moisture leaving the basalt boxes through existing gaps and bores was Minister of the Interior had requested KfK in the spring of 1985 to review the extracted along with the argon. The action was finished in January 1987. more recent scientific literature on the Bethe-Tait accident and indicate in a comment on the SNR 300 whether there were new references in the scientific literature that would require a reassessment of the provisions made against damage due to an uncontrolled mechanical energy release of the reactor core. The For the SNR 2, the Europaische Schnellbrüter-Kernkraftwerksgesellschaft (ESK) result produced by KfK in late 1985 was this: and Interatom use a pool type primary system. This makes use, on the one hand, of the lead in operating experience accumulated with the French Super Phénix The Bethe-Tait accident continues to be a hypothetical accident, i.e., (SPX 1), which is also a pool type plant, and is also a clear step towards the there is no argument in the literature which would give rise to the need to development of a joint European type of breeder reactor. There are no signi­ assume a higher probability of occurrence than had been postulated earlier ficant technical and cost differences relative to the loop type design. on. With respect to its probability of occurrence and its consequences, the Bethe-Tait accident is still covered in the description of the uncontrolled Here are some other major characteristics of the SNR 2 concept: loss-of-flow accident. Concentration of safety related technical systems upon the nuclear steam The expected sequence of accident events continues to be non-energetic. raising system and, consequently, less stringent criteria to be met by the conventional part and the auxiliary systems. Recent experimental findings allow the range of parameter studies in boundary case assessments to be restricted much more severely. Minimization of the capacity of the important safety related emergency power loads, e.g., by using natural circulation for decay heat removal in Mechanical energy releases of more than 100 to 150 MJ still cannot be case of a breakdown of the plant electricity supply system. substantiated by any physically conclusive line of arguments. Mechanical energy releases above 370 MJ therefore can be excluded to all practical Concentration of measures of protection against external impacts on as few intents and purposes. buildings and systems as possible.

On permit 7/6 (storage of the core assemblies in the facility), all expert A preventive safety concept obviates the need for designing against hypo­ opinions were submitted to the authority by the end of January 1986. A list of thetical sequences of events. pro memoria items compiled by the authority on the state of deliberation of the scope of permit 7/6 (April 22, 1986) was updated repeatedly (June 23, 1986; Utilization of LWR experience in streamlining the licensing procedure. September 25, 1986); all conditions, some of them new, were completed by the applicant/manufacturer by late October 1986. All major expert opinions have been 3.2 The Nuclear Steam Supply System (Fig. 6) produced also on permits 7/7 and 7/7 (1) (reactor loading, zero power tests, power operation). The nuclear steam supply system of the SNR 2 is characterized by key data as represented in the "Status of the DeBeNe Fast Breeder Reactor Development" 2.3 SNR 300 Test Fuel Assembly Program report published in January 1986.

Some Mk.II fuel assemblies are to be tested in the first core (Mk.Ia) of the In line with the pool type 'design concept, the reactor vessel contains the SNR 300 for the benefit of the second core. The design, including the completion reactor core, the four primary pumps, and the intermediate heat exchangers of fabrication drawings, of the Mk.II test fuel assemblies has been completed. (IHX). The vessel has no nozzles; it is surrounded by an outer vessel. All specifications for the blanks and structural components have been examined and cleared by the expert consultant. The fuel assembly specification with the The reactor core rests on the core support structure, which is a box type respective item specifications for the fuel pellets, cladding tube, wires, etc. structure carrying the grid plate for positioning of the core assemblies and for have been agreed upon with the fuel assembly manufacturer. The pretesting coolant supply to the core assemblies. documents required for fabrication have been completed in part and, in some cases, have already been cleared for fabrication by the expert consultant. The core contains the hexagonal fuel assemblies arranged in two enrichment zones and surrounded by the blanket and reflector assemblies. The 25 absorbers of the The fabrication of cladding tubes, wrapper tubes, and spacers has been completed primary and 12 absorbers of the secondary shutdown systems are located on fuel with the acceptance and documentation of the individual parts. Fabrication of assembly positions. Both shutdown systems are independent and diverse. the other structural parts of the assemblies has been started. The vessel will also accommodate the storage space for spent fuel assemblies the heat of which will be transmitted by natural circulation. 3. Planning the SNR 2 Cold and hot sodium are separated by a double walled internal tank. After having 3.1 Technical Planning Status passed through the core, the sodium flows through the hot collector, enters the IHX windows at the top and flows down the IHXs, leaving them for the cold Although no contract for the detailed planning phase has as yet been awarded, collector, where it is taken in by the primary pumps and returned to the core conceptual planning of the SNR 2 has been continued also in 1986. These ac­ through the pressurized pump lines. tivities have meanwhile led to an advanced status of technical planning. Eight secondary sodium systems connect the tops of the IHXs above the reactor closure head to the eight steam generators located in two separate, opposite buildings. All sodium pipes are enclosed in double jackets and run in pipe conduits, respectively, within the reactor building. Consequently, there can be no safety related sodium fire in the reactor building in case of a leak in a sodium pipe.

The two steam generator buildings contain the eight straight tube steam gener­ ators as well as the secondary pumps and the supporting facilities of the secondary systems. The steam generator buildings are designed to withstand the safe shutdown earthquake.

3.3 Further Development and Optimization

Planning work at present is concentrated on the following sectors, in which the planning status is to be advanced and optimized:

Flow pattern in the primary system. Design of the reactor vessel closure head. Seismic behavior of the reactor core. Handling concept. Accident management concept.

Optimization of the Flow Patterninthe Primary System

The startup behavior of natural convection in the transition from the power.mode to the decay heat removal mode of the reactor largely determines the thermal loads acting on the core assemblies and reactor vessel internals. The startup behavior is influenced very much by the position of the immersion coolers. Consequently, five different arrangements have been studied. In order to be able to assess properly the complex results, a catalog for evaluation was compiled Fig. 5: SNR 2 cross section. complete which weightings. The most favorable arrangement of the immersion coolers will be determined in cooperation with foreign partners.

The studies of a decay storage facility for spent core assemblies inside the vessel, which would be cooled exclusively by natural circulation, were completed The hot collector also contains the four immersion coolers, which can remove the with a positive result. This concept, which is also called an insular storage decay heat in case of failure of the main heat transfer system. Their optimum system, therefore will be considered as a reference solution for the project. layout will be defined in connection with the natural circulation characteris­ tics of the plant. Design_of_the Reactor Vessel Closure Head

The whole reactor block is suspended in the reactor closure head, a self-sup­ The closure head is part of the primary containment and carries components porting welded structure, which is air cooled and filled with shielding concrete important for shutdown and decay heat removal. on the inside. The IHXs, primary pumps, and immersion coolers are suspended in the outer, fixed part of the closure head. In the center there is the rotating First calculations of the temperature development in case of breakdown of the reactor plug system consisting of two eccentric rotating top plugs which can be air cooling system show differential temperatures between the top and the bottom turned independently. The inner top plug carries the instrumentation plug and sides of the closure head, which can be considered to be barely permissible in the transfer arm used to move core assemblies from a core position to the the light of a simplified stability assessment. transfer chute and vice versa. The short margin to the permissible limit requires more in-depth investigations Spent fuel assemblies to be removed from the reactor are transported through a by analytical 3D-methods and, perhaps, the construction of a test model. chute to a transfer position in the reactor building, fresh fuel assemblies come in the opposite direction. Seismic Behavior of the Reactor Block the power required by these systems to be estimated, thus permitting a concept to be elaborated for an emergency power supply system not dependent on large The SNR 2 has a core with free-standing elements. Extensive three-dimensional quick-starting diesel engines. calculations of the movements of these elements under earthquake conditions point to major deflections of the tops relative to the shutdown rods and link­ 3.4 International Cooperation in Planning the SNR 2 ages arranged above these. This is due to the close proximity of the natural frequencies of the soft pool type reactor structures and the core assemblies. In the spring of 1986, INB/Interatom, Ansaldo, and Novatome signed an agreement The first attempts at "detuning" the natural frequencies by means of design on cooperation in evaluating the passive safety concept proposed for the SNR 2 changes have shown that major steps need to be taken. under which a harmonized proposal for the conceptual design of the next European breeder is to be elaborated. The work is carried out by the firms at their Handling Concept respective sites and under their own responsibility within the framework of a joint study group. As a consequence of the passive safety concept, special studies were conducted on handling the core assemblies under sodium. This work related in particular to The Ansaldo program focuses on an analysis of the feasibility of the safety the conceptual design of a shipping and storage vessel. Also thermal studies of decay heat removal system operating in the natural circulation mode and on the the cooling concept of the handling cells were carried out. special loads and stresses arising to the reactor block out of this concept.

Ongoing investigations are concentrated on the ordinary handling of core assem­ The Novatome program pursues similar goals and subjects. It mainly differs in blies in the reactor and the handling of failed core assemblies. also including a critical review, with the assistance of the CEA, of the load criteria of the core. A special R&D program has been defined to investigate the feasibility of wrapper tubes not wetted by sodium. In 1986, also the British manufacturers (NNC) cooperated intensively in the further development and optimization of the SNR Z. Some major areas of work were Accident Management_Concegt the support of the reactor closure head, the steam generator concept and design, the core structures, and decay heat removal. Technical safety considerations are concentrated on the accident management concept designed to III. PROGRESS IN RESEARCH AND DEVELOPMENT limit the concept of "safety relatedness" to those systems, which are indispensable to meeting the limits specified in the German Radiation Protection Ordinance, and The following survey is a summary of the most important research and development projects and the results obtained in work by the "DeBeNe" partners for the Fast classify as "operating systems with safety functions" those systems, which Breeder Project in 1986. only serve to minimize the impacts of failure.

Studies were conducted of the following failures: 1. Core Assemblies

Maloperation of shim rods giving rise to local excess power levels while Irradiation of two SNR Mk.II fuel assemblies in the French Phénix breeder leaving the global core power unchanged (load tilting). reactor has been most successful. The burnups and dose levels attained by the Malfunctioning of the primary pumps. Mk.II assemblies to date are 93 MWd/kg and 80 dpa NRT and 107 MWd/kg and 87 dpa Release of primary cover gas as a result of leakages of the cover gas NRT, respectively. The continuation of this irradiation depends on the long­ system or the closure head. itudinal growth of the wrapper tube due to swelling. Because of the satisfactory Sodium leakages in the reactor building. behavior of the assemblies presently loaded, studies are now being conducted on Secondary sodium leckages in the steam generator building. modification and further irradiation of an assembly to even higher burnups. Steam generator leackages up to a 2-F break of a pipe. Failure of ventilation and recooling systems, respectively, especially as a The DUELL-II operating transient experiments have also furnished favorable consequense of breakdowns in the plant power supply system, and temperature results for the Mk.II fuel of the SNR 3Û0. The DUELl in-pile experiments were rises in components and in the reactor building, respectively. conducted to study the behavior of fuel rods typical of the SNR 300 under the influence of power increases after steady-state operation in the early stage of The studies are conducted parallel with the planning of the system, and as this in-pile irradiation (so-called startup ramps). The irradiations were carried out has not yet been carried to the same level of detail for all systems, they will in the Pool Side Facility of the Petten HFR. be continued. The question is to be clarified which systems and installations, respectively, must be classified as "safety related" and thus must be initiated In a joint KfK/CEA/ENEA program, 11 single-rod experiments with various test and controlled from the accident control room. In addition, the studies allow rods carrying intended defects were run in the Si loe test reactor (Grenoble). Major parameters of these experiments were the burnup in a fast neutron flux No major axial fuel migration. (pre-irradiation in RAPSODIE, PHENIX), size and position of the defect, the fuel No longitudinal growth of the fuel column. geometry, and the cladding material. Nine of these experiments (S2, S3, S4, Intensive fuel/cladding interaction in all rods of low density. RS-1, RS-5, CYCLO-1+2, PS-0, PS-1) have already been subjected to post-irradi­ ation examination (PIE). Fuel entrainment was found to be slight, even in the In the IDEFIX joint CEA-DeBeNe experiment, two rod bundles were irradiated in presence of major initial defects. This experience is valuable for the operation the Phénix reactor. The rod cladding tubes were made of oxide dispersion of KNK II and the SNR 300. In the KS-1 experiment, a KNK 11 /1 rod was operated strengthened ferritic steel grades. Their main asset is knwon to be the resi­ up to natural overload failure after 4,8 % burnup and then irradiated further in stance to neutron induced swelling at doses of 80 - 100 dpa NRT and above. The the Si loe loop for another three reactor cycles. PIE has not yet been completed. main drawbacks of conventional ferrites are their unsatisfactory mechanical pro­ perties at the operating temperatures of breeders. These disadvantages can be The PS-1 test rod (with 1.4970) was shortened, fitted a central thermocouple, avoided by certain types of dispersion strengthening which, however, is likely exposed to internal helium pressure, its cladding weakened and then specifically to complicate fabrication and requires correspondingly accurate fabrication con­ taken to cladding failure in the Si loe reactor. Here are some of the major trol. The planned IDEFIX experiments with 2 x 216 rods serve to test both results of the PIE conducted at KfK: in-pile behavior and the method of fabrication.

17 secondary cracks in the highly embrittled cladding due to the formation DeBeNe materials irradiation in the PFR includes the PORRIDGE and PFR-M2 in-pile of Na,M0., projects. They contain specimens with and without pressurization to measure approx. 1 g loss of fuel (from in-pile measurements in Siloe), radiation induced creep in combination with potential swelling due to radiation. a considerable number (approx. 50) of mixed oxide particles had dropped PORRIDGE continues the RIPCEX-2 experiment, containing also Belgian specimens of inside into the central void at the level of the primary defect during oxide dispersion strengthened ferrite (0DS) at temperatures of 400 and 500 °C. irradiation in Siloe, The specimens at 400 °C consist of an earlier variant of the 0DS series of slight plutonium enrichment at the edge of the central void, which had alloys. They have attained doses between 45 and 68 dpa NRT and will be unloaded already been found in the reference rod in Mol, was confirmed also in PS-1. for PIE by the end of the 13th cycle. The test rig at 500 °C contains German and Belgian specimens. Their irradiation will be carried to a total dose of 50 - 63 Some preliminary results of non-destructive assays of selected fuel rods of the dpa NRT at the end of the 13th PFR cycle. German-French CHARLEMAGNE bundle irradiation project have indicated a very sTTght increase Tñ diameter and length of the German 1.4970 cladding tube 2. Reactor Physics material at the dose attained of approx. 80 dpa NRT. In addition, a number of batches with Si and Mo contents higher, within the given limits of specifica­ The last assembly studied in SNEAK, 12C, represented a reactor core with a tion, than the reference formula showed the highest dimensional stability. central zone of plutonium platelets. The assembly had been used for experiments Consequently, 19 selected rods were prepared for further irradiation in the on the effectiveness of control rods, especially heterogeneity and moderation Phénix reactor up to a target dose of approx. 150 dpa NRT in cooperation with effects, and for material relocation experiments. Investigations conducted on a the CEA. rod with different levels of moderation achieved by the replacement of absorber material by zirconium hydride showed that the use of natural boron can increase In the defective rod program, the Mol-18 series has been completed successfully rod effectiveness by moderation by some 18 %\ in the case of fully enriched in the BR 2. Two experiments are still under evaluation: In Mol-18-ВЗ, a full- boron, the effect is only mild and, moreover, tends to reduce effectiveness. The scale KNK II/2 rod with 2 t burnup had been exposed to several transients results of subsequent calculations by means of transport theory are hardly extending to more than 150 % of the initial power and, at the end, to a high better than those obtained by diffusion calculation. cladding temperature transient of up to 850 °C sodium temperature. In Mol-18-A2, a rod preirradiated to 4.8 % in KNK 11/ 1 safety survived eight power transients In determining control rod worths within the commissioning phase of SPX 1, the to 150 % of the design power followed by 850 °C sodium temperature. difference between precalculated and actually measured reactivity worths for the secondary shutdown system initially was found to exceed that of the control shim The FARFAÜET in-pile experiments are being evaluated by SCK/CEN. In those system. Closer examination showed this difference to disappear if more precise experiments, which are so-called "beginning-of-life" irradiations, fuel rod account was taken in the calculations of the complicated structure of the behavior and the fuel-cladding interaction are investigated as a function of the secondary shutdown rods. fuel density and the fuel-cladding gap width. FARFADET-1 contained two fuel rods of low density (fabrication by BN), FARFADET-2 two high-density rods (fabricated The ID and IE RACINE assemblies (Fig. 7) were designed to support the cal­ by Alkem); gap widths differed in each case. Irradiation took 15.6 days. The culation of control rod worths in SPX 1. While RACINE ID contained only one envisaged rod power of 500 W/cm was attained. In-pile measurements showed central rod, 12 rods of various designs were studied in IE. Agreement between considerable axial heat transport. Non-destructive PIE revealed the following calculation and experiment is good or satisfactory in all cases. However, for facts : enriched boron the С/E values (ratios between calculation and experiment) are found to be higher. The differences between the С/E values found in Super Phénix Formation of a central void in all fuel pellets. (SPX I) and those found in control rod experiments in RACINE ID and IE, which Pu-enrichment of the fuel surface of the central void in the hottest region are close to 1.0, may be due to one or more of the following reasons: of the fuel column. 34 RAC1E sufficiently accurate for application calculations, unless step widths change hc=90 cm too much. The numerical method developed at KfK for the description of uniphase Control \ 77 A Blankets and multiphase flows under the impact of sudden changes in flow cross sections rods is better equipped to describe accelerated and steady state flows than the (ZZl Co re zones method applied in the standard version of SIMMER 11.10. This version was used for parameter studies of the expansion phase of energetic accidents in the SNR 300. As the pressure in the core region is reduced relatively slowly at high initial temperatures of the fuel, also the load removal effects in the radial and downward directions must be considered, in addition to the kinetic energies of the sodium accelerated in the upper coolant plenum, in assessing the vessel integrity. SIMMER is not able to calculate these effects accurately, because it is based on the assumption of rigid structures (i.e., those incapable of de­ formation) . Fig. _7: RACINE IE critical assembly Single experiments are carried out at KfK to verify the models used in SIMMER. cross section Within the framework of experiments performed on the expansion of the sodium pool into the cover gas plenum, the influence on the expansion of the bubble of such parameters as volume, filling level, and pressure was determined syste­ Uncertainty in determining heterogeneity corrected cross section data for matically. The rate of injection of the filling liquid was determined (fast gas the SPX 1 control rods. injection experiments). To clarify the transition phase, the experiments on the thermal reaction of U0„/steel were supplemented by studies of the influence of Difficulty in calculation, expecially with the asymmetrical rod con­ various parts bv volume of the individual components (multiphase-multicomponent figurations in SPX 1 (e.g., only one rod in or out). tank experiment). The work will be applied increasingly in accident analyses in the longer term. Different values for the effective fractions of delayed neutrons in RACINE and SPX 1; RACINE had been partly built up of enriched uranium. The SIMBATH thermite experiments simulating transient materials movements have illustrated details of the flow phenomena encountered in core disruptive acci­ Uncertainties in the cross sections of BIO, for RACINE and SPX 1 have dents. Stratification and freeze-out behavior, respectively, had been studied different neutron spectra. also in the TRAN B1 (Sandia, USA) in-pile experiments in a simple geometry. In order to establish a comparison of the behavior of fuel with that of thermite, Construction of another large critical zero power assembly is under discussion. the same experiments in the same geometry and, where possible, with the same It would serve to clarify some discrepancies remaining in the calculation of parameter settings were repeated with thermite at KfK. The use of X-ray cine­ control rod configurations, for studies of axially heterogeneous cores, and to matography allowed transient movements of materials to be observed, showing that verify the determination of the power profile by detectors set up outside the the initial flow of material into the test section occurs in a droplet flow and core. As the critical will still be much smaller than the breeder core to be not as a slug flow, as modeled in the PLUGM Sandia code used for TRAN recal­ simulated, a compact core has been suggested by the DeBeNe side, which generates culation. A code suitable for recalculating the experiments has become available the lower neutron leak rate found in larger reactors by doing without the with the BUC0GEL computer program acquired from CEN Grenoble. The first preli­ simulation of the coolant. Such assemblies have already been studied on a minary results show satisfactory agreement with respect to depths of penetration smaller scale in the SNEAK 10 series. and velocities.

At present, four different code systems in the field of nuclear core design and The Mol 7C-in-pile blockage experiments serve to study the behavior of sodium the analysis of critical experiments are used by the present European partners cooled fuel rod bundles in a major local loss of coolant. The Mol 7C/4 experi­ in cooperation, which greatly aggravates this cooperation. Since early 1985, a ment was expanded, the Mol 7C/5 experiment using irradiated rods (burnups of 48 study group had been looking at the existing systems and, in the light of those and 17.5 MWd/kg of metal, respectively) was examined after irradiation. Theo­ findings, presented the draft of a new joint code system in September 1986. retical studies with the BACCHUS code were conducted to describe thermohydraulic Although the introduction of this system would result in some temporary extra events in the blockage region. expenditure, it would certainly lead to a considerable long term reduction in the present manpower requirement in this sector. The first few ceramographic sections for the Mol 7C/5 experiment are now avail­ able. They show that a melting cavern had formed, as in the Mol 7C/4 experiment. 3. Fast Reactor Safety The fuel crust is slightly thicker, but the melting cavern is clearly smaller than in Mol 7C/4. Fig. 8 shows cross sections through the blockage. The upper The SIMMER II program for analyzing major core disruptive accidents has been part of the picture shows the crust broken up and liquid fuel escaping also in tested critically. For various irregular lattices it was found that the tech­ Mol 7C/5. The fuel crust obviously "healed" after breaking. The escaping fuel nique implemented in SIMMER 11.10 for averaging densities and velocities is was carried away by the sodium flow and deposited in the next spacer grid in the The thirty transient experiments in the French CABRI experimental reactor provided for under the present international program were completed in 1986. Evaluation of the experiments will continue until 1987/88. The following para­ meters, out of the multitude affecting the sequence of accident events, were studied in the CABRI test program:

The energy added to the fuel during an excursion (variation between 0.4 and 2 kJ/g of oxide). Burnup (fresh fuel, low and medium burnup). The time of initiation of the power excursion relative to the onset of sodium boiling in the cooling channel (before, during and after the onset of boiling) in experiments conducted at a reduced coolant flow.

The results of the CABRI test program have greatly added to the knowledge of the phenomena associated with accidents. They have supplied extensive measurement data for further advancement and verification of models to be used in the safety analysis of fast breeder reactors.

The experimental program agreed upon between KfK and CEA on studies of combined sodium fires has been started. Three experiments have been performed so far, which are at present in the evaluation stage. The experiments are to be com­ pleted in 1987.

After modification of the FAUNA facility, experiments will be performed on the sodium-concrete interaction (hydrogen problems).

Methods of calculation for containment analyses are being upgraded. For this purpose, the CONTAIN computer program system was adopted from Sandia National Laboratories in 1984. That integral accident analysis code treats events oc­ curring in a reactor building after failure of the vessel and leading to failure of the outer containment and to activity being released into the environment. The FAUNA 5 and 6 experiments were recalculated to test the sodium pool fire model in the CONTAIN code. After improvements in the appropriate equations and constitutive data satisfactory agreement between the computed and the measured results was found in a consistant way for both experiments, at least with respect to the global development. With respect to some other CONTAIN models, e.g., core melt-concrete interactions, fission product transport and decay, and gas transport between cells, test calculations and parameter studies were carried out. In addition, an ESMERALDA experiment is to be recalculated.

The FAUST water experiments devoted to studies of aerosol behavior in bubbles rising during accidents in which the reactor core is destroyed completely or in rig. 8 : Ceramographic sctions through Mol 7C/5 blockage part have been completed. The new test rig designed for investigations of sodium a) Upper third of blockage in the same problem category has been commissioned. Two experiments were per­ b) Middle third of blockage formed with solid particles above 1 pm in size.

Experiments planned to study the influence of aerosols on heat transfer in the region of the upper third of the blockage. The local temperature elevation top region of the reactor vessel were harmonized with Interatom, especially brought about by the plugging of the flow channels caused a defect in one fuel questions of the instrumentation required. These experiments, and the FAUST rod in the outer row of the bundle. However, in contrast to the Mol 7C/4 ex­ experiments, are performed on the same facility. The facility has been modified periment, there was no major increase in size of the defect. appropriately for this purpose. The first experiments were begun in late 1986.

35 Thermite experiments were performed on the freezing of fuel melts. A narrow gap has been observed to form betwen the crust of the fuel'1 and the walls of the "internals". This gap, together with thermal conduction in the crust, determines radial heat transfer by thermal conduction and radiation. The American PLUGM computer program describes this condition satisfactorily. Experiments with superheated metal and oxide melts were performed on the penetration of melts into structures. A first few series with oxide and metal melts have already been run to study the behavior of a fuel jet impinging upon steel structures. The melt/concrete interaction will be investigated mainly in 1986 and 1987 after modification of the SUSI test facility. These experiments are of particular importance in the light of the Chernobyl accident.

To demonstrate decay heat removal in the SNR 2 by natural convection only, studies are conducted in water with three-dimensional tank models on a 1:5 scale (NEPTUN) and 1:20 scale (RAMONA) (Fig. 9). For the NEPTUN model, the main components (tank vessel, immersion cooler) and the power supply system have been ordered. The tank vessel has already been assembled. In the test rig meanwhile built to test the fuel assembly simulators, the test series designed to deter­ mine pressure drops in the fuel assembly simulator internals, such as heater rod bundles and B.C shielding, have been completed. On the basis of these findings, the type of fuel assembly simulator will be selected for installation in the model core. In the RAMONA model, temperature distributions in a veriety of zones were measured in some first few series of experiments and flow patterns observed by means of tracers and video recordings.

The first thermohydraulic analyses of the transient behavior of the decay heat removal system after failure of the plant power supply system carried out with the Italian NATURA code on the reference concept (four immersion coolers su­ spended in the hot colector) have confirmed the results achieved with the DYANA code of Interatom. According to these, and taking into account the thermal capacity of the main secondary systems, the core is cooled within safe limits also by three out of four immersion cooling circuits. In addition, Ansaldo already considered an arrangement of immersion coolers in the cold collector, which allows a "cold" immersion cooler secondary circuit and may perhaps offer the advantage of clearly directed natural circulation flows with lower core temperatures. The points still open include questions of stratification effects in the cold collector and the consequences arising with respect to the design of the intermediate heat exchanger, whose outlet window would have to be raised in this solution.

4. Methods of Measurement for Plant Surveillance

The boiling generator experiment performed in KNK II in 1984 served to determine the acoustic transfer function from the -core to the upper plenum of the reactor. Theoretical determination of the transfer characteristics is not possible because of the acoustically complex core internals with their large number of steel-sodium transitions along the path of sound propagation. For this reason, the boiling noise was measured both close to the source of boiling and at realistic probe positions in the upper plenum of a power reactor. Boiling noise was recorded in all of these positions. The general shape of the transfer function shows that an acoustic detection system must work preferably in the frequency range between 5 and 50 kHz, as the higher-frequency fractions of the original broad band boiling noise (far above 100 kHz) is attenuated too much because of the low-pass characteristics of the transmission path. It may be concluded that local coolant boiling could be detected by monitoring A concept has been developed for a diagnostic system designed to monitor the for noise in the upper plenum. steady state operation of a large fast breeder reactor. The system is to be able to recognize early incipient breakdowns and propose countermeasures. In addi­ Pinpointing failed fuel assemblies in KNK II is achieved by a combination of tion, scram analysis is to be carried out to determine, for selected frequent load tilting and a dry-sipping test using the fueling machine, both techniques cases, that signal out of the flood of signals which was primarily responsible have been proved to be satisfactory. However, in looking for the fuel assembly for initiating the scram. Surveillance must cover not only the core region, but in KNK II which had failed last, the prediction made by load tilting did riot also those parts of the heat transfer system, disorders of which might have a agree with the assembly later found by means cf the sipping test. direct impact on the core region. A system of this type is to be demonstrated in KNK II. The measurement and evaluation procedures available at that plant will Two gamma spectrometriс assays of the radionuclide inventory in the primary be used for this purpose. sodium of KNK II were carried out in February 1986, shortly before the plant was restarted. The only fission products detected were Сs-137 and Cs-134, which were 5. Components and Structural Material measured by their minor residual activities. Further levels were measured in May, shortly before the reactor was shut down for the removal of a failed fuel Further theoretical studies were conducted to clarify the 1eak-before-break assembly. The release of fission products into the sodium had given rise to behavior. To demonstrate the reliability of the main coolant pipe 07 tïïë higher levels of Cs-137 and Cs-134; also other fission products were measurable. SNR 300, a pipe elbow in the hot leg of the primary system was considered. The However, the activity concentrations of the fission products continued to be failure probabilities of circumferential and longitudinal welds were calculated clearly below the specific activity of the Na-22 activation product. The ratio on the basis of plausible assumptions about crack size distribution. Both leak between Cs-137 arid Cs-134 had remained at 5. This indicated that the fifth and break probabilities rise but slightly during the period of operation. This cladding tube failure in KNK II is similar to the fourth one, i.e., that the means that most of the cracks able to penetrate through the wall will do so assembly involved again is a driver assembly. after the first few load cycles. After that they will continue to grow as one-dimensional cracks, perhaps up to breaking (leak-before-break criterion). The sorption of radionuclides from sodium was investigated in KNK II with The contributions made to the leak probability by crack growth during operation material specimens. Niobium sheets were Toaded into the hot primary sodium in are comparatively slight. Break probabilities rise steadily in the course of two experimental plugs besides Ni and stainless steel specimens. This was done operation. in order to verify findings made in the Siloe defective rod program which had indicated selective sorption of fission product cesium to niobium specimens. The EASY code is a very efficient instrument for computing the elastic behavior, Experiments with Nb foils conducted in KNK II had only been run as far as 320 °C fatigue behavior, and plastic instability behavior of cracks extending through and had not indicated any sorption of Cs-137. The Nb sheets in the tést plugs the wall and surface cracks. were subjected to sodium temperatures above the core was still high enough to clearly activate with Ta the sheet metal and its slight contamination. Fission Even complex, locally variable loads with two-dimensional stress gradients are product cesium was sorbed selectively; the only two measurable fission products, reduced to one simple numerical integration by means of a modified weighting Cs-134 and Cs-137, were calculated per unit area of the sheet metal. function technique. The library of weighting functions contained in the program is being expanded continuously. A method of sodium removal and decontamination of loops was tested. For sodium removal, a method has been developed in France which uses ethylcarbitol (EC), a Studies of the behavior of the SNR 2 reactor roof in case of cooling failure heavy alcohol. This was to be applied technically in decommissioning the RAP­ resulted in a "moderate" insulation of the closure head. The cooling of the roof SODIE reactor. CEA and Interatom think the new cleaning procedure will find and of the rotating reactor plug are not to be supplied with emergency power; broader application in sodium technology. An opportunity to learn more about the consequently, several cooling failures must be postulated to occur throughout use of EC was found at the mock-up of an Interatom test loop, which was to be the service life of the unit. In this respect, various load cases must be dismantled anyway. The low radionuclide inventory existing in that system distinguished with respect to the duration of a failure, the failure rate and allowed not only sodium removal by means of EC, but also decontamination by the number of available emergency power systems. It is doubtful whether mere slight caustic wastage to be tested. analytical proof of reliability in these load cases will succeed. It willhave to be found out whether a roof structure and a roof insulation matched to these The removal of cesium on carbon fiber s-tructures was demonstrated. As this conditions, respectively, can be found which, in case of a cooling failure, procedure is limited in operation to temperatures of 200 °C, it can be used only results in clearly calculable and tolerable loads and stresses. If the result with the reactor shut down. Niobium does not suffer from this drawback; it is were negative, some back-up experiments would be required. The proposals out­ being loaded in KNK II test plugs for experimental prurposes and tested in the lined so far, and investigations based on them of a roof structure insensitive NATAN test loop. Particles removed from the primary sodium of KNK II were to cooling failure, have clearly added to our basis of knowledge and have also studied morphologically and chemically. led to new goals of optimization.

A tritium probe developed by Interatom for use in the cover gas was commissioned The AKB large scale sodium experimental facility (a facility for structural core for trial purposes in the primary helium of the AVR reactor at Jlilich. Two parts 57 Interatom) has been beefed up after some 40.000 hours of operation. advanced oxygen probes with liquid In/In-O, reference electrodes were delivered AKB, which was built in 1970 for tests of actuating rods and fuel assemblies of to CEA for joint testing in a sodium loop under primary sodium conditions. the SNR 300, will play an important role also in the further development of the evaluation of the CHARLEMAGNE, OLIPHANT, Phênix-Mk.II (high-dose experi­ SNR line. As the 50 MW facility at Hengelo has meanwhile been dismantled, ments with 1.4970), and HILDEGARDE (1.4970 as wrapper tube material) further components for the SNR 2, above and beyond those listed above, are to be experiments have been planned. tested in AKB. Components and operating facilities no longer required have been removed in the meantime. The remaining components were subjected to in-service In the IDEFIX joint CEA-Debene experiment, two rod bundles are being inspections and most of the electric installations redone. A modern process irradiated in the Phénix reactor, the rod cladding tubes of which are made control system has been installed for control and surveillance of the unit. In of oxide dispersion strengthened ferritic steels. addition, also the facade lining of thefacility will be replaced for reasons of safety at work. In the Si loe experimental reactor, eleven single-rod experiments with various failed test rods have so far been carried out in a joint program by KfK/CEA/ENEA on such main parameters as burnup in a fast neutron flux (pre-irradiation in IV. INTERNATIONAL COOPERATION RAPSODIE, PHENIX), defect size and defect position, fuel geometry and cladding material :

On nine of these experiments, post-irradiation examination has been com­ 1. Cooperation in the West European Region pleted. Fuel entrainment was found to be slight even if the initial damage was great. This experience is valuable for the operation of KNK II and The memoranda and agreements signed with France in 1976/77 have resulted in a SNR 300. major expansion of cooperation. An extension of that cooperation to the United Kingdom and Italy has been initiated by the Memoranda of Understanding of In the KS-1 experiment, a KNK 11 /1 rod was run up to naturally induced January and March 1984. In July 1984, a European Steering Committee was ap­ overload failure after 4.8 % burnup and then irradiated in the Siloe loop pointed for R&D coordination, to which eleven European working groups (AGTs) now for another three reactor cycles. KS-1 has been completed successfully; report. post-irradiation examination still needs to be done.

Some of the most important joint R&D ventures in Western Europe in the field of In the ELISE loop in Siloe, the PASSION (seven Phénix rods), TRANSIT (three in-pile experiments, physics and safety will be summed up briefly below. SPX 1 rods), OCARINA (three SPX 2 rods) and PS-2 (7 Charlemagne rods) experi­ ments are to be carried out jointly with CEA/ENEA. 1.1 In-pile Exgeriments_of_Core_Assemblies and_Materials The British prototype fast reactor, PFR, is used, among other purposes, for Several Debene experiments have been concluded in the French Phénix prototype irradiating cladding material specimens of the Debene group: breeder or are under way and being planned, respectively: The PFR-M2 and PFR EXCHANGE test series (creep, swelling). PFR-TRANSIENT (mechanical properties under transient reactor conditions). Irradiation of assemblies to confirm the Mk.II concept of the SNR 300 was completed in 1986. As the in-pile time was controlled by the swelling The range of Debene materials irradiations in the PFR also includes the PORRIDGE behavior of the French wrapper tube material, a possibility is at present project, with specimens and without pressurization, to measure the radiation being investigated to load the respective fuel assembly in a new, un­ induced creep developed in combination, potentially, with swelling due to irradiated wrapper tube and then continue irradiation. irradiation. PORRIDGE is an extension of the RIPCEX-2 experiment and also contains Belgian specimens of oxide dispersion strengthened ferritic materials In 1987, the POUSSIX and GIGONDAS experiments will be terminated, which are (ODS) at temperatures of 400 °C and 500 °C. The specimens at 400 "C consist of being carried out in cooperation with the CEA. They serve for studies of an early variant of the ODS series of alloys. They have accumulated doses the mechanical interaction between the fuel and cladding during startup. between 45 and 68 dpa (NRT) and will be unloaded for post-irradiation exa­ mination by *he end of the thirteenth PFR cycle. The test rig at 500 “C contains In the BARSAC end-of-life experiment, moving a failed fuel rod with high German and Belgian specimens. Their irradiation was continued up to a total dose burnup was studied for the first time under storage conditions inside the of 50 - 63 dpa NRT by the end of the thirteenth cycle. reactor vessel. Additional tests of the behavior of failed fuel rods under storage conditions will follow, and contributions with also be made to the 1.2 Neutron Physics and Nuclear Core Design RAISIN experiment. One experiment under planning will use a failed small bundle assembled out of SNR-Phénix rods under storage conditions inside the Before and during the commissioning of SPX 1, a KfK delegate to Cadarache did vessel. Both the SNR 2’ and the SPX 2 are to have internal fuel assembly preparatory calculations and evaluations of experiments with CEA data and storage facilities within the reactor vessel. The Soviet BN-600 is equipped methods. These activities concentrated on the reactivity worth of control rods. with 124 storage positions inside the vessel. Subsequently, KfK together with the Debene partners will evaluate the experi­ ments with KfK data and methods. Besides the control rod experiments, also the To study fuel assembly materials, the CROMIGNON (10Cr25Ni) and SAMARCANDE power distribution (reaction rates) and reactivity coefficients will be treated (1.4914) experiments as well as the post-irradiation examination and in these calculations. The joint large critical assembly, CONRAD, is in its definition phase. KfK for sensitivity and reliability within the framework of qualification tests. The proposed to include in this series also a compact core. After experience with result of these trials in KNK II will be used as the basis for deciding whether SNEAK 10, one is able to approximate, in a relatively small compact core, and how the detectors and broad-range measuring channels tested are to be used properties of a larger reactor core, among these is control rod interaction. in the SNR 2 and SPX 2 plants, respectively.

Other control rod experiments performed in SNEAK 12C, and sodium void experi­ 1.3 Safety Studies and Sodium Technology ments carried out in the British BIZET-D assembly, have been evaluated. Eva­ luation of the international comparison of reaction rate measurements performed The experimental part of the CABRI transient test program conducted so far was in the last RACINE assembly in the French MASURCA zero power assembly is pro- completed in 1986 with the 30th experiment of the test matrix. Evaluation and gressing wel1. full documentation will require considerable expenditures also in 1987/88.

For the JEF 2 nuclear data file to be established jointly by Europeans and CEA considers continuing the CABRI project under the name of CABRI-2. The main Japanese, evaluation was updated for Am 242m and begun for inelastic scattering results of the CABRI-1 program and a first proposal of a CABRI-2 program were of U238, for Pu241, and for Fe isotopes. KfK has adopted parts of the THEMIS presented to the Steering Committee. In June 1986, AGT-4 elaborated a preli­ processing code system and tested them. minary European CABRI-2 test matrix at Cadarache which was then discussed in detail at KfK, among other places. Contributions were made to the user specification for a new joint European neutronics code system. For the ECCO joint heterogenity code, routines were made KfK did comparative experiments with thermite on selected TRAN experiments of available for calculating collision probabilities in cylindrical geometries. the Sandia series. They were recalculated by means of the PLUGM code. The Under the same heading, contributions are at present being made to determine BUCOGEL code developed by CEA was adopted by KfK and will replace PLUGM. diffusion coefficients for void and streaming situations. Results are available of the Mol 7C/1 to 5 cooling blockage experiments con­ In the large three-dimensional diffusion codes in rectangular and triangular ducted in the Belgian BR 2 reactor. A decision has been taken in the meantime geometries, respectively, various algorithms are at present being tested for about the programs and the funding of the Mol 7C/6 and 7 experiments; the effectiveness within the framework of vectorization. Comparison of European participants will be KfK, SCK/CEN, and IRC Ispra. Mol 7C/6 is to be conducted in diffusion codes with regard to effectiveness and precision has been started with 1987, Mol 7C/7 in 1988. The results of the first three experiments (unirradiated the objective of standardizing the codes used. rods) have been passed on to the Japanese PNC and the USNRC (in exchange for results of the American P4 blockage experiment). In the meantime, PNC has shown The efforts initiated by AGT-3 towards the harmonization of methods and computer an interest also in Mol 7C/4 and 5 (irradiated rods). UKAEA participated in the codes for the nuclear design of fast breeder reactors have come to a first evaluation of Mol 7C/5. conclusion, after having led to a practical comparison, in the form of work­ shops, among the four European program systems, COSMOS (UK), CCRR (CEA), KAPROS As the experiments conducted so far have demonstrated the profound influence of (KfK), and IANUS (IA), in 1985. burnup on the sequence of failure events, Mol 7C/6 is to be carried out with rods of 94 MWd/kg burnup. The detailed design report for Mol 7C/6 is now avail­ The results and recommendations stemming from that evaluation were presented at able. a specialists' meeting at Karlsruhe in September 1986 and the conclusions were discussed. A cost-benefit analysis shows that changing to a future joint system Mol 7C/7 will be planned with fresh KNK II fuel rods which, in addition, will will offer major advantages especially against the background of an expected new have outer diameters of 7.6 mm instead of 6 mm; instead of a central blockage, FORTRAN standard (FTN 8X). In this way, the costs of maintaining and updating there will be a blockage in the boundary region. computer codes could be clearly reduced for all participants in the longer term, uniform standards of quality assurane could be ensured, and the productivity and The experimental program of' studies of combined sodium fires agreed upon between possibility of division of labor in project development could be improved. The KfK and CEA has been started. Three experiments have so far been conducted, attending representatives of all industries involved expressed themselves in which are currently under evaluation. The experiments are to be completed at KfK favor of introducing such a system. in 1987. In addition, an ESMERALDA experiment is to be recalculated.

For monitoring reactor power by means of neutron flux measurement, proposals for In the commissioning of SPX 1 some KfK delegates perform measurements and the SNR 2 as well as for the SPX 2 seek to have the neutron flux detectors for accompanying studies, some of them related to the observed temperature vari­ all neutron flux measurement ranges installed inside the reactor vessel close to ations in fuel assemblies. the core. High temperature fission chambers designed for these measuring jobs are being developed, optimized and will be made available by CEA und UKAEA. A method for cleaning circuit tubes has been tested at Interatom. France had Under conditions approaching as closely as possible the future use in the SNR 2 developed a new method of removing sodium based on the use of the ethyl carbitol with respect to temperatures at the point of measurement and also neutron and (EC) heavy alcohol. This was applied technically in decommissioning the RAPSODIE gamma radiations, these fission chambers are to be loaded in KNK II and tested reactor. In the opinion of CEA and Interatom, the new cleaning technique is likely to find wider application in sodium technology. One opportunity to learn and to releases of activities into the environment. FAUNA experiments were more about the use of EC existed in the "Mock-up" Interatom test loop. Because recalculated to check the sodium pool fire model in the CONTAIN code. After of a low radionuclide inventory, decontamination by means of slight caustic improvement of the appropriate equations and constitutive data it was possible wastage was tested simultaneously with sodium removal by means of EC. to find satisfactory agreement, consistent for the experiments, between the computed and the measured results, at least with respect to global development. 2. Cooperation with the USA Test calculations and parameter studies were carried out for a number of ad­ In view of the budget cuts suffered by breeder activities in the USA and the ditional CONTAIN models, e.g. those relating to core melt-concrete interactions, realignment of some of that work to smaller power plant modules, different fission product transport and decay, and gas transport between cells. nuclear fuels and integrated reprocessing procedures, the DeBeNe side carefully observes further developments in the United States. However, there is also The thermite experiments performed by KfK to simulate transient materials active cooperation with the USA in a number of sectors. movements illustrated details of the flow processes encountered in core dis­ ruptive accidents. Stratification and freezing out, respectively, had been The Schnell-Brüter-Kernkraftwerksgesellschaft mbH (SBK) and the American Elec­ studied also in the TRAN 81 in-pile experiments (Sandia, USA) in a simple tric Power Research Instisute, Inc. (EPRI ) had signed an agreement in 1985 on geometry. In order to insure comparability of the behavior of fuel with that of the exchange of knowledge in a number of fields. Under that agreement, a com­ thermite, these experiments were repeated in thermite at KfK in the same geo­ prehensive discussion was organized in Washington, D.C., on November 11 - 14, metry and with parameters as nearly identical as possible. Transient materials 1986 about the following topics: movements were observed by means of X-ray cinematography. The evaluation of ACRR The PRISM and SAFR modular concepts and their safety aspects. experiments will be concluded at KfK in 1987. New developments and safety advantages in the field of metal breeder fuels for high burnups. Integrated fuel cycle facilities and their costs. Special attention was devoted also to the future uses of FFTF. The event was 3. Cooperation with Japan attended, among others, by representatives of EPRI, GE, Rockwell International, Bechtel, ANL, SBK, KfK and Interatom. Debene and CEA had a trilateral agreement with the Japanese PNC since 1978 on cooperation in a number of areas of breeder development. Efforts are under way Proof of decay heat removal by natural circulation requires numerical simulation to include also the United Kingdom in this cooperation; at present, problems are by a computer program able to simulate with sufficient accuracy the complex being clarified in connection with an existing UKAEA/JAERI agreement. geometry of all components integrated in the reactor vessel. The COMMIX code supported by the US Nuclear Regulatory Commission at ANL is being considered for At a German/French/Japanese experts' meeting at Oarai on April 14 - 18, 1986, this purpose. C0MMIX-1A was the basis on which the three-dimensional two-phase which was attended also by Italian and British observers, radiological conse­ COMMIX-2 program had been set up, which had been developed jointly by ANL and quences of reactor accidents were discussed. Agreement was reached on the KfK. exchange of findings on

Analyses of major accidents by means of the SIMMER code adopted from LANL relate sodium spray fires, to energy release, materials transport, mixing and heat transfer events, and the sodium pool fires at low temperature, source term. The accuracy of SIMMER-II solutions for various irregular lattices combined sodium spray and pool fires, was investigated in more detail at KfK. sodium fires on an inclined steel liner, sodium-concrete interaction and hydrogen release, KfK is running individual experiments for verification of the models used in source term determination. SIMMER. Within the framework of experiments on the expansion of the sodium pool into the cover gas plenum, the influence on the expansion of the sodium bubble Since 1985, also a Japanese delegate has been participating in the evaluation at of volume, filling, level, and pressure was investigated systematically. The KfK of the joint KfK/Toshiba experiments in the FPL-2 fission product loop. The rate of injection of the filling liquid was determined (fast gas injection experiments were conducted to clarify events associated with the release, experiments).. To clarify the transition phase, the KfK experiments on the transport, and deposition of radioactive impurities in the primary sodium U0„/steel thermal reaction were supplemented by studies of the influence of system. various volume fractions of each of the components (multiphase-multicomponent vessel experiments). The Japanese PNC indicated its interest in active participation in the dis- assembly of the French RAPSODIE fast experimental reactor, including evaluation With respect to major accidents, also methods of computation for containment of the experiments. The experts’ meeting on seismic design, which had been analyses were upgraded. For this purpose, the CONTAIN computer program was planned for June 1986, was postponed until some future date by request of Japan. adopted from Sandia National Laboratories by KfK in 1984. This integral accident analysis code deals with events occurring in a reactor building after failure of European-Japanese contacts exist also in other areas of breeder safety, thermo­ the reactor vessel and potentially leading to failure of the outer containment hydraulics, structural materials, and steam generators. The 5th DeBeNe/CEA-ENEA-Japanese FBR Cooperation Review Meeting was held at PNC A REVIEW OF THE INDIAN FAST REACTOR PROGRAMME on January 20 - 23, 1987. The discussions focused on questions of safety in­ cluding sodium water reaction, the CONTAIN code, experiments on decay heat removal by natural convection, and CABRI follow-on, reactor operating ex­ S.B. BHOJE perience, components and structural materials, fuel and core materials, and fast Indira Gandhi Centre for Atomic Research, reactor physics. Several specialists' meetings'were agreed upon. Kalpakkam, India

1.0 GBHBRAb The total installed eleotriolty generation capacity in India, aa оf end 1986, was 47.6 GWe,

made up of 64$ coal baaed thermal, 3 3 hydel and

3$ nuclear. Total eleotricity generated during '86 vaa about 180 bKVfh. The present annual compound growth rate of the installed capacity is about Q.5%

and the total planned capacity by the end of 1990 ia

about 66 QVe. The demand for the electrical energy ia growing continuously and the planned capacities will not be adequate to meet the demands. This year,

there ie power out in parta of the country due to drought.

Indigeneously designed and built DHRUVA reactor

(100 MWt) at BARC is operating satisfactorily at 40 MWt. Severe fuel element vibration problem has been solved by modifying the fuel support. The power level will be further raised in steps. Two BWR unite at Tarapur have achieved capacity factora over 5Ф». The second unit of Rajasthan has operated very well during the year with a capacity factor of 70^4 and has

41 operated without an interruption for 163 days. The