BNES international conference on fast reactor power stations: Conference report

Annals of Nuclear Science and Engineering,

Vol. 1, pp. 463 to 474. PergamonPress 1974. Printed in Northern Ireland

BNES INTERNATIONAL CONFERENCE ON FAST REACTOR POWER STATIONS: CONFERENCE REPORT A . R . BAKER Fast Reactor Systems Directorate, U K A E A Reactor Group, Risley, Warrington, Cheshire

A b s t r a c t ~ A t the Conference, held in London in March 1974, there were five plenary sessions devoted to past and current experience and to future plans for incorporating fast reactors into the electrical supply networks and there were six specialist sessions devoted to: steam generators; nuclear performance; core including fuel; control, dynamics and hydraulics; coolant management; and fuel handling and other mechanisms. The main points raised in the papers to the Conference are summarized and, in a final section, the writer makes his own assessment of the conclusions to be drawn from the information presented.

INTRODUCTION during commissioning with the pumps and with the internal fuel handling machine. The experience in dealing A conference was held on Fast Reactor Power Stations with these faults demonstrated the feasibility of handling under the auspices of the British Nuclear Energy Society, large components in and out of sodium using simplified on 11-14 March 1974, at the Institution of Civil Enginhandling techniques and of carrying out major deeers, London. The large number of delegates, about contamination and dismantling operations. 450, from about 20 countries, showed the widespread In his oral presentation, Mr. Evans concentrated on interest in the subject. There were five plenary sessions events since the printed paper had been prepared. Fuel devoted to past and current experience and to future loading was completed and criticality achieved at the end plans for incorporating fast reactors into the electrical of February and the reactor was then under test at very supply networks and there were six specialist sessions on low power with the primary and secondary sodium particular topics arranged as two sets of three in parallel circuits at approaching 400°C. The first loading was sessions. itself a substantial test of the fuel handling system since The conference was timely as the construction of three it represented a number of operations that would occur large reactors had recently been completed. The Russian in future over a period of about 18 months. The reBN 350 had been producing power for about a year, the actor was expected to produce electrical power in the French P H E N I X reached full power during the consummer. ference and the U.K.'s P F R at Dounreay had just preThe next presentation was a sad occasion for those in viously been made critical. The main parameters for the audience who had worked on fast reactors for many these reactors and others mentioned at the conference years. Mr. E. L. Alexanderson described the contribuare collected together in Table 1. tion to fast breeder technology of the FERMI-1 reactor In his introductory address, Dr. M. Davis, Director which was then being decommissioned. He pointed out of Nuclear Power, E.E.C, summarized the importance of fast reactors and their place within the world programme that the experience over many years was well documented of nuclear power. He said that the recent rise in the and should provide a valuable reference to other deprice of oil left no doubt of the economics of nuclear signers on what had been successful. Many components, power. The expected rapid build-up of thermal reactor such as the pumps, had operated for the full life of the power stations would lead to greatly increased demand reactor. There was considerable expertise in maintenance for uranium. Fast reactors offered the promise of including the major operations necessary to deal with the economy in the world's uranium requirements by two melt-down some years previously and it was important to make full use of what could be learnt. orders of magnitude. A sizeable proportion of fast reThe paper gave a brief account of recent experience. actors would be needed as soon as their reliability was established and it would be unwise to wait until the A new oxide core had been needed. When the finance necessary for the new core and for continued operation economic benefits were fully demonstrated. had not been forthcoming, the reactor had to be finally shut down. The decommisioning itself presented some PLENARY SESSIONS 1 A N D 2: CONSTRUCTION, unexpected problems; how to dispose of the slightly COMMISSIONING AND OPERATING radioactive primary sodium and of 30 tons of breeder EXPERIENCE WITH PROTOTYPE PLANTS uranium containing 6 kg of plutonium. He suggested The first paper was presented by Mr. A. D. Evans, that designers should give thought to the problems of on behalf of the host nation, on the design, construction decommissioning when designing new reactors. Mr. N. V. Krasnoyarov presented the paper describing and commissioning of PFR. The previously circulated paper had described experience up to December 1973. four years' experience of operating the BOR 60 fast If difficulties rather than achievement are emphasized reactor experiment. A large number of oxide fuel elein this conference report this is because they are apt to be ments had operated beyond 10 per cent burn-up. Twelve more interesting to readers, who wish to learn by other sub-assemblies had failures which led to loss of gas tightpeople'sexperience. Mr. Evans pointed out the difficulties ness which were located by a movable fission product 463

P

P

142 1000

5"8 325

G

45

11

142 914

5-8 325

G

45

8-5

620 540 490 16

103

342

78

660 565 538 16

3

3

650 560 510 16

7"2

42

W

6'6 217

124 850

3 6

3 6

P

563 250

3140 1300

600 250

France 1969 1973

UK

PHENIX

1966 1974

UK

CFR

620 530 490 18

8'6 271

358

4 8

P

3000 1200

France

530 490

3.7

56

W

6-0 37

44 400

2

L

60

1969

USSR

SUPERPHEN1X BOR 60

* BN 350 is also designed to produce 120000 m3/d of water from the sea. 1" The number given for F F T F is that of driver fuel.

Country Dates Construction start Criticality Output Gross thermal Gross electrical Type Pool (P) or Loop (L) Number of units Primary pumps Intermediate heat exchangers Steam generators Core Number of fuelled SA Outside dimension of wrapper across flats (mm) Fuel height (mm) Pin outside diameter (mm) Number of pins per SA Grid (G) or wire wrapped (W) Max linear power (kW m -1) Max flux (1015 n cm -2 s -1) Temperatures (°C) Nominal max clad mid-wall Primary sodium outlet Steam at turbine Steam pressure (MPa)

PFR

500 350 3.2

W

6"1 169

96

6

L

1000 150"

1964 1972

USSR

BN350

SNR 300

SNR-2

545-580

W

6-9 127

96

P

551 510 8

L

60 20

1966 1971

546 500 17

G

6.0 166

110 950

205

9

3 9

L

300

1973 1979

7.6

4 4

L

5000 2000

1980

USSR Germany Germany Germany

BN600KNKII

Table 1. Comparison of reactors discussed at the conference.

640 525

4.7

39

6-7 91

82 650

L

1973 1978

Italy

PEC

4or 8 8

P

1000

Italy

Demo Plant

0

3 3

7

482-593

W

217

115 910

73t

L

400

1977

USA

111 900

196

3

3 3

L

714 300

1975 1980

Japan

636 535 482 10

48

W

540 487 13

4

46

5"8 6"5 217 169

910

198

3

3 3

L

975 380

1974 1981

USA

Clinch F F T F River M O N J U

rn

e~

).

4~

BNES International conference on fast reactor power stations : Conference report detector (described in Specialist Session 5) at the subsequent shut-down period. The secondary circuit survived sodium leaks without a fire and an oil leak into it which was gradually cleaned up by cold trapping. Mr. J. Megy outlined the design of PHENIX. The reactor had been only 5 months late in connection to the grid, a success which he attributed to good co-ordination of all concerned and adequate testing beforehand of major components including a 45 MW modular steam generator. Mr. B. Guillemard continued the story of P H E N I X with more details of commissioning and start-up. Again it was the unexpected incidents, all of which were dealt with without too much difficulty, which were of most interest to the audience. There was an oil leak into the reactor before sodium filling and later there were minor leaks from valves in the secondary circuit. One type of control rod failed by a few millimetres to fall by the full traverse due to jamming by sodium deposits in a small clearance. The three control rods concerned were removed, cleaned and modified to prevent a recurrence. This incident showed the wisdom of having two types of control rod since, in this case, only one type was affected. During the discussion period, Mr. Krasnoyarov gave further details of the operation of BOR 60 with failed fuel. He said that fuel pins failed by splitting apparently due to irradiation swelling and not to fission gas pressure. He added that the reactor had been operated with failed fuel for as long as a month. To another question, this time on BR 5, he replied that when two failures occurred in one subassembly they were in separated pins, indicating that the failures were unrelated. PLENARY S E S S I O N 3: PLANNED POWER

PLANTS Mr. J. A. Kyger described the use of EBR-II in support of breeder development. This reactor has been used as a test facility since 1965 and oxide fuel has already been tested up to 15 per cent burn-up. Nitride and carbide fuels have been tested too. In some cases, the fuel was run to failure and it was shown that end-of-life failures were benign. The xenon/krypton tagging method for detection of failed fuel was tested on EBR-II and will be implemented on later U.S. reactors. Continued operation of the reactor demonstrated the durability of the pumps and the fuel handling equipment. An intermediate heat exchanger in which one of the tube supports failed was successfully repaired. Mr. J. E. Nolan described the progress of the construction of the Fast Flux Test Facility (FFTF) then 35 per cent complete. The oxide fuel has xenon/krypton tags for failure detection as developed on EBR-II. There will be closed and open loops for testing of special fuels. There is a passive restraint system for the control of swelling and bowing of sub-assemblies. Mr. W. B. Behnke described the U.S. L M F B R activities and the Clinch River Breeder Reactor Plant (CRBRP). He said that the objectives were to demonstrate the reliability and safety and to gain confidence in the economics. It was not intended to compete with LWR's. The organizational structure and the main parameters (included in Table 1) were described. It is interesting to note that, at this stage, provision is made in the design for the choice of one or other of three different types of steam generator. The design leans heavily on F F T F experience. Mr. H. Kitagawa introduced the paper describing

465

developments in Japan. In his oral presentation, he dealt briefly with the reactor JOYO for which criticality is expected in 1975. This reactor will give experience in construction, sodium technology and fuel irradiation. His circulated paper described the main parameters of M O N J U and summarized recent experience in research and development. Mr. A. Brandstetter introduced the paper describing the German reactor SNR 300. It is noteworthy that in this loop-type reactor there is no pipe in the reactor vessel below the sodium level required to cover the core. There is a multiplicity of containment structures including a final one surrounding the whole plant designed to withstand the crash of an aircraft. The same licensing rules are being applied to this prototype plant as would be applied for a fully commercial reactor. This will assist the procedure for later licensing of larger reactors. Initially the reactor will have 6 mm fuel pins similar to designs from other countries but it is proposed that the second charge shall have larger pins to reduce the fuel fabrication costs. Mr. G. Brudermuller continued the description of the German work with a paper on the operating experience with the K N K reactor and the preparations for a fast mixed-oxide core (KNK-II). This paper was particularly interesting because of its frank discussion of the various problems that had had to be dealt with. These included sodium frost which prevented correct operation of the rotating shields and which led to lengthening of the drop time for one type of control rod. The latter problem, which affected only one type of control rod, again showed the importance of having two separate types. There was a failure in the steam generator which led to the leaking of 50 kg of sodium into the water side. The unit was repaired by cutting out the failed tube and was then run with some neighbouring sodium passages plugged with reaction products until these passages cleared themselves during normal operation. Finally the subassembly thermocouple indicated a blockage in one subassembly. Despite extensive tests, no evidence for a blockage nor explanation of the temperature rise was found before the trouble cured itself. Mr. G. Cicognani of Italy presented a paper describing the reactor PEC which was designed to be a small irradiation test facility. As well as giving fuel irradiation experience, it will also give experience in the design and operation of components. Three closed loops with independent coolant circuits will enable fuel to be tested as far as failure. Mr. B. Musso, in a second Italian paper, described a limited study of a 1000 MWe demonstration plant based on the PHENIX design. The problems concentrated on were measures to reduce the size of the tank in a pooltype reactor and the dimensional stability of a core under neutron irradiation. During the discussion, there was interest in the requirements for continued core cooling in the event of primary pump failure. There was evidently difference of view on the need for pony motors to back up the primary pump motors and on the need for separate decay heat removal loops with natural circulation. In answer to a question why it was now thought possible in the U.S.A. to have passive rather than active restraint against voidage swelling, Mr. Kyger said that new evidence had shown that irradiation creep was great enough to make passive restraint possible.

466

A . R . BAKER

SPECIALIST S E S S I O N 1: STEAM GENERATORS

The Italian paper by Mr. P. Casalini et al. identified the steam generators as key components and listed the main design aspects. Chief among these were the precautions to prevent major sodium-water chemical reactions, the detection of minor ones and design precautions against thermal shock. After several design studies, the straight tube type was selected for development. The prototype units under test in rigs might be linked either in oncethrough or recirculating modes. There was also a brief mention of a design study with helical tubes. The Japanese paper by Mr. Y. Nakai et al. gave a detailed specification of the M O N J U steam generators. A once-through helical-coil design was proposed with sodium reheat. It was possible that the same generator might be used in a recirculating mode at part load. For the evaporator, 2 ~ Cr-1 ~ Mo ferritic steel was proposed with stainless steel type 316 for the superheater. They apparently hoped to prevent stress corrosion by operational control of the water coming over into the stainless steel sections. The generator was designed to withstand the pressures caused by simultaneous rupture of four tubes. The research and development programme was described. A quarter-size (50MW) M O N J U generator was under construction. The next paper by Mr. A. J. H. Mante et al. described the tests in the Netherlands and Germany of the design of major components for SNR 300. The components included pumps, valves, intermediate heat exchangers and the top rotating shields complete with fuel-handling equipment. There was considerable interest in the incident during the test of a 50 MW prototype steam generator with straight tubes and bellows to allow for expansion. There was a major sodium-water reaction due to a faulty weld in the reheater which, however, did not fracture the bursting disc nor give the expected large signal on the hydrogen-detection system. Future designs would remedy these defects and would incorporate emergency pressurization with nitrogen on the water side after blowing off steam. A 50 MW prototype of a helical-tube type generator had also been built and recently was put under test. A second separate paper by Mr. C. A. J. van der Krogt gave a more detailed description of the two types of steam generator proposed for SNR 300. It was claimed that the present state of generator technology required a multiplicity of units for economic reasons. Six of the straight-tube type and three of the helical-type were proposed for SNR 300. The U.S. paper by Mr. J. S. Armijo et al. cited the available detailed evidence to support the choice of the reference material (2¼~ C r - l ~ Mo unstabilized steel) for the Clinch River steam generator, a forced circulation unit which allowed dry-out in the evaporator tubes. Incoloy 800 was selected as back-up material. A small amount of decarburization of the reference material was expected but the designers were confident that allowance could be made for the small changes of strength in consequence. The over-riding concern was stress corrosion and all materials, except the ferritics, had been shown in accelerated life-tests to suffer to varying degrees, particularly in the superheater sections if wet-steam contamination took place. The paper contains a useful table comparing the material choices for tubing in more than 20 steam generators. The French paper by Mr. M. G. Robin et al. described the unique type of modular steam generator success-

fully utilized in PHENIX. Each module consists of a coil of 7 water tubes (without a tube sheet) and 12 of these make each of the 3 steam generators. This design was preferred because of its ease of fabrication and transport, its expected reliability, its ability to limit the consequences of sodium-water reactions and also because of the possibility of previous testing of a full-scale module prototype. During commissioning and start-up, good agreement was demonstrated between predicted and operational conditions. The generators were stable at 5 per cent flow at normal decay heat removal temperatures and at flows with significant operational margin at higher operating temperatures. The water impurity levels were quoted for the start-up conditions of the P H E N I X generators, indicating that the levels of all impurities were high by a factor of 5-25 on the specification. All impurities were brought to lower but still substantial levels in a short space of time. The pH of the water was kept between 9.2 and 9.4 by dosing with ammonia. In the U.K. paper by Mr. D. Taylor et al., it was claimed that the P F R was unique among fast reactor prototypes in that its steam generators had a shell-and-tube design which made them realistic forerunners of those to be expected for a future large plant. The steam pressure and temperature were suitable for use in a large-turbine cycle normally associated with fossil-fired power stations. The ' U ' form of the tubes placed the welds to the tube plate in gas spaces where they could be easily monitored for leaks. An explosive plugging technique for dealing with defective tubes was demonstrated during commissioning. The commissioning experience was detailed, including the successful demonstration of the hydraulic performance on the sodium side during a water test. There was considerable interest in the Russian presentation because of the national press reports just before the conference of a major explosion in BN 350. It was stated that there had been leaks in steam generators which had given rise to sodium-water reactions of various magnitudes. It was understood that the leaks occurred at the welds of tube to tube-plate and of end caps to bayonet-type tubes. There was no explosion nor fire and all had been accommodated by the protection system installed. It was said that the two major incidents occurred after continued operation with indications of a small leak into the sodium and as much as several hundred litres of water were involved in the interaction. Of the six steam generators, two were out of service after the incident, one other was shut down as it showed a similar indication of a small leak but three were still in operation, allowing the reactor to continue to operate on part load. There was also an interesting account of the experience with a leaking steam generator in K N K - I I reported in a paper to Plenary Session 3. SPECIALIST SESSION 2: NUCLEAR

PERFORMANCE Mr. M. F. Troyanov introduced the paper which described the comparison of calculations with measurements of physics parameters on the BN 350 reactor. The calculations were performed with a revised version o f the 26-group ABBN set published in 1964. It was disappointing that the previous critical assembly on the zero-power reactor BFS had not led to a closer extrapolated value of kerr for BN 350. Therewas also a need for improvements in the accuracy of control rod calculations.

BNES International conference on fast reactor power stations: Conference report Mr. Y. Maeda introduced the paper describing the nuclear performance of M O N J U . Numerical values were given for the usual range of parameters. The major areas left to be resolved were stated as the sodium void coefficient, subassembly bowing, the change in reactivity due to burn-up and the improvement of the form factor (power-factor). Mr. J. G o u r d o n presented the paper describing the experimental physics results at the start-up of PHEN1X. Information was given on the performance of the installed neutron source and the low power counters. These counters were sited below the reactor tank and the neutrons had to reach them through a considerable depth of sodium. The counter sensitivity was found to depend markedly on the temperature of the sodium between the core and the counters. It was also interesting to note the power distributions produced by measurements with the installed outlet thermocouples on each subassembly. These were accurate enough to check calculations of power distributions and the information from them would complement the data obtained with much greater effort from scans with activation foils during the low-power start-up phase. Mr. D. C. G. Smith concluded the session with a paper describing the initial measurements on PFR. He reported that nothing unexpected had been found from the key core parameters measured at low temperatures. These included the critical approach, control rod worths and isothermal temperature coefficients. SPECIALIST S E S S I O N 3: CORE INCLUDING

FUEL Mr. J. F. W. Bishop opened the session with a paper on the performance development of the P F R fuel. He described how the demonstration in D F R of the then target burn-up for oxide fuel of 5 per cent paved the way for the building of PFR. The operation of P F R itself would pinpoint the remaining development areas. He described how the requirements for boiling-noise detection had influenced the mechanical design of the subassembly and had led to detailed vibration testing in water loops. It had become apparent that the key points influencing the performance were voidage swelling due to irradiation damage, irradiation creep and chemical interaction between the clad and fuel. Mr. J. A. G. Holmes continued the U,K. presentation with a description of the mechanical design of the P F R subassembly. The P F R had a so-called free-standing core with the subassemblies supported only at the bottom. This was chosen originally in order to obtain a negative coefficient of reactivity due to bowing. When the importance of irradiation-induced voidage growth became apparent, it was clear that this type of core had disadvantages. These might be removed, to some extent, by rotation of subassemblies in the course of their irradiation but, for PFR, a restrained core was being considered for installation later as a more fundamental attack on this problem. Mr. Krasnoyarov presented the Russian experience with fuel irradiations in the BOR 60 reactor. The small number of failures were described in the 1240 fuel pins taken to a burn-up of up to 11 per cent. Successful performance was also reported of 80 per cent boron-10 carbide control rods for which the burn-up of B10 was over 5 per cent. The experience on BOR 60 suggested

467

that pin failure occurred near to the point of high rating and burn-up. Mrs. Menshikova continued the Russian presentation with a discussion of the development programme for fuel elements for BN 350 and BN 600. She made the important statement that mass failure of fuel pins was not observed up to 10 per cent burn-up. As a rule, she said, single fuel pins failed. She concluded that the analysis of failed fuel pins did not permit an unambiguous determination of the cause of cladding failure. The German paper by Mr. J. Hochel et al. presented the design of the fuel element for the SNR reactor, the mechanical tests, the irradiation experiments and the proposals for fuel manufacture. The reactor was to have a passive core-restraint system consisting of two ferritic steel rings located respectively above and below the core. A computer study had shown that irradiation-induced creep was quite effective in reducing both the stresses in the wrappers and the forces in the core-restraint system. The paper presented economic figures to justify the decision to change from a 6 mm pin to a larger 7-6 mm pin for later cores and also for SNR-2. German speakers stressed the importance of the choice of a larger pin during discussion periods throughout the conference. There was a detailed review on failed fuel behaviour by G E authors, Mr. P. E. Bohaboy e t al. This report indicated that oxygen in the coolant was unlikely to be a controlling factor in fuel-coolant chemical interaction. Some estimates of failure rates as a function of burn-up were attempted and fission product release data were given. They concluded that considerable research remained to be carried out in order to evaluate the operating, safety, environmental and economic factors associated with failed fuel behaviour. The paper by Mr. M. Barigozzi et al. outlined the design criteria for the core of the PEC reactor and the theoretical and experimental work performed in support of the core design. The experimental programme included hydraulic testing of fuel elements, thermal hydraulic testing in sodium, an irradiation programme and experiments on the vibration behaviour of a group of seven subassemblies. The final paper by Mr. Estavoyer et al. described the P H E N I X core and the observations made during the early commissioning stage. SPECIALIST SESSION 4: CONTROL, DYNAMICS AND HYDRAULICS Mr. Y. Nakai opened the session with a paper describing the M O N J U heat transport system. It was noteworthy that emergency core cooling was to be provided by the normal primary circuit via a separate section of the intermediate heat exchanger. The paper concluded with a description of the aspects that required a major development effort. These included pumps, valves and intermediate heat exchangers. Mr. H. Hoffman presented a paper describing the work in Germany on thermal performance and hydraulics. He listed the computer codes used. He also described the experimental studies, mostly using gas, of pressure drop in rod bundles, interchannel mixing and heat transfer. He also described the studies in a water model of flows in the reactor tank and, using sodium, of studies of emergency core cooling. Mr. J. Branchu described the commissioning tests on

A. R. BAKER

468

the primary and secondary circuits of P H E N I X during the preceding 13 months. The tests included checks of the correct operation of major components, including tests on vibration, measurement of sodium levels and pipe displacements. Modifications were found to be necessary in the non-return valves in the primary circuits. This work provided an opportunity to check the handling of large components into the reactor vessel. Mr. U. Ciriegi described the main characteristics of the PEC reactor and then went on to describe the special features required in a control system for a fuel-testing reactor. Mr. G. B. Collins introduced the paper on the dynamic modelling and control studies for the P F R plant. It outlined the dynamic models that had been developed for control, fault and performance studies. The principles for the choice of control system were discussed together with a description of its implementation on the plant computers. The methods for commissioning the controls and the programme of dynamic tests were also described. Mr. A. Hopkinson continued the U.K. presentation with a detailed description of the P F R simulator. This was to be used for training operators and for study and development of operational procedures. The detailed information given will be of value to all those with a similar project. A paper by Mr. 1. A. Kuznetsov described the dynamic measurements made on BN 350 during the commissioning and early stages of operation. He was mainly concerned with flow measurements with pumps in various states and with the performance of non-return valves. He described the appreciable reactivity addition on shutting off pumps due to elastic displacement of subassemblies. Mr. Acket described the dynamic tests that have been performed on P H E N I X to confirm the theoretical models used to predict transient behaviour. The three main problems described were the thermal shock at the heat exchanger on shut-down, the temperature transient following disconnection of a primary pump and the behaviour of the steam generators following an electrical supply failure. The models were found to give a good description of events but some adjustment of parameters was necessary. S P E C I A L I S T S E S S I O N 5: COOLANT

MANAGEMENT Introducing the session, the Chairman, Mr. L. Vautrey, outlined the topics of concern to operators. These derived from the presence of impurities in the coolant or blanket gas, which promoted corrosion and offered the possibility of released corrosion products causing blockage of channels. He stressed the need to control impurity levels by on-line monitoring or analysis and the need to monitor for failure of fuel elements and of steam generators. Mr. P. Chouard described the sodium purification and sampling systems of both the primary and secondary circuits in P H E N I X and the experience obtained to date. Stress was laid upon the presence of two plugging temperatures determined by the plugging meter. The lower one was always below 150°C and was attributed to sodium oxide, while the higher one could occur at temperatures of 230°C or more and its origin was not

identified. Whereas it was established that the efficiency of cold traps for oxide impurity was virtually 100 per cent, the same could not be said for the unidentified impurity. Hydrogen injection tests on the secondary circuit at temperatures of 350-450°C demonstrated the influence of hydrogen concentration on plugging temperatures and the ability of the cold traps to remove hydrogen. The results were also described of the commissioning tests of the sodium sampling system to determine the initial levels before fuel loading of such materials as carbon, caesium, manganese and fissile nuclides and to determine the sensitivity and reproducibility of the equipment. The trapping of sodium vapours from the cover gas was also briefly discussed. In presenting his paper, Mr. R. A. Davies defined the purpose of monitoring as identifying changes which would occur during reactor life which might affect the properties of structural materials; as reaching an understanding of mass transfer processes; and of determining the regions most susceptible to the build-up of radioactivity. To these ends, P F R had been equipped with sampling and distillation equipment, a plugging meter, carbon meters, a particulate collection unit, removable specimens and blanket gas sampling. Multiple plugging temperatures were reported as in the preceding paper on PHENIX. Since the concentration of carbon in sodium did not indicate the carburising potential, instruments of two types were to be installed on PFR to measure this directly. One type of measurement was based on diffusion in a permeable iron membrane and the other used an electrochemical carbon-concentration cell. The secondary circuit was fitted with hydrogen meters, based on the diffusion of hydrogen across a nickel membrane, to detect small water leaks from the steam generators. Mr. J. Jung described the development in Germany of instruments, in a specially designed sodium chemistry loop. Typical results were shown from continuously measuring instruments for the determination of hydrogen (using a permeable membrane), of oxygen (by electrochemistry) and of carbon (measuring the carburising potential). Cover-gas impurities, hydrogen, oxygen, nitrogen and methane, were determined by on-line gas chromatography, an argon ionization detector and an enrichment stage being used when more precise data were required. In the paper, by Mr. 13. de Clerq et al., which followed, the design, construction and testing of prototype sodium cold traps in the Netherlands was described. The following Japanese paper, by Mr. Y. Nishikawa et al., was concerned with quite different topics; fuel failure and the required safety instrumentation on individual subassemblies. After a brief calculational survey of the consequences of various types of subchannel blockages it was concluded that an adequate provision for each subassembly of M O N J U would consist of two temperature sensors, one flow sensor and one unit for determination of failed fuel by delayed neutron and fission product gas detection. The results of research and development on fast-response thermocouples, other temperature sensors and three types of flow-meter were described. Mr. H. A. Rohrbacher presented a paper on the development of ultrasonic and acoustic detection methods for LMFBR's in Germany. The continuous wave method in ultrasonics had been extended to be capable of detecting bubbles of diameter, 0"3 to 10 mm, travelling at

BNES International conference on fast reactor power stations : Conference report several metres per second. The pulse reflection method in ultrasonics was being developed to locate precisely the position of fuel element heads and to detect any obstacles which protruded. Special lithium niobate sensors had been produced capable of operating at temperatures up to 600°C. The acoustic measurements in K N K , of noise mainly due to pumps, and the proposed trials of boiling noise detection in that reactor and in an electrically heated loop were described. A Russian paper, by Mr. N. N. Aristarkhov et al., described a unique system for detecting failed fuel elements which had been demonstrated on the test reactors, BR 5 and BOR 60. At shut-down, the equipment was moved over each subassembly in turn so that argon gas could be bubbled through it. The failed fuel elements were detected via the radioactive inert gases flushed out with the argon. On BOR 60, the mean time for testing one subassembly was 15-20 rain and 100 per cent detection efficiency was claimed. The final paper, by Mr. E. Benoist et al., described the special instrumentation for detecting hydrogen in the P H E N I X heat exchangers. A sodium sample was taken from the outlet of each of the 12 modules of a steam generator. A single measuring device (based on a permeable nickel membrane) served via switches the group of 12 samples from each of the evaporator, superheater or reheater. The smallest continuous water leak that could be detected was estimated at about 0'004 g/s. The module at fault could be detected for a continuous leak of 0.03 g/s. The discussion periods disclosed a general concern about detection of failures both of steam generators and of fuel pins. In answer to a question, Mr. Davies said that a P F R steam generator would be shut down for a water leakage of 0.5 gls (compared with 0.18 g/s for PHENIX). In the course of a very active exchange, it became clear that acoustic methods, such as those described by the French, were being developed elsewhere to detect steam generator failure. The Japanese work on fuel element failure stimulated a number of comments. It was said that the reliability (failure rate) of the instrumentation described by them had not been established and there were no definite proposals yet for the instrumentation for MONJU. The P H E N I X delayed-neutron monitors were said to be capable of detecting 0-2 cm z of exposed fuel. Interest was shown in the nature of the material responsible for the higher of the multiple plugging temperatures hut, while it was suggested that this was due to hydride, no specific confirmation was forthcoming. Both the Russians and the French were asked if, in spite of their cold trapping facilities, they had detected any changes in impurity levels during power changes. The French said that none had been observed in P H E N I X . Neither was information forthcoming on any relationship between oxygen levels in blanket gas and deposits on surfaces. Interest was shown in the extensive materials monitoring programme which has been mounted in PFR. Nothing comparable appeared to be envisaged elsewhere. SPECIALIST SESSION 6: FUEL HANDLING AND OTHER MECHANISMS

Mr. W. Jansing opened the Session with a paper on the experience with fuel handling in K N K , the test

469

programme for SNR 300 and the proposals for SNR-2. He described the difficulties that were experienced with sodium deposits in the rotating shield system of K N K which had necessitated design changes. The SNR 300 components were being tested full-size in a test rig at Bensberg. The work with inflatable rolling seals for the rotating plugs were of interest. During the discussion later, he said that these seals had a life of 5 yr. Major changes from the choices for SNR 300 were proposed for SNR-2 to accommodate the larger size of the reactor, the larger subassemblies and the shorter handling time for replacing fuel subassemblies that was being aimed for. It was noteworthy that the Germans were proposing to abandon a rotating-plug system with direct-pull fuel handling for a transfer arm type of fuel handling system as used in P F R or P H E N I X . Mr. S. Basile continued with a paper on the design and research and development for the PEC core-related mechanisms. He described the control rods, the core hold-down mechanism and the refuelling machine. The handling of fuel was made more difficult because of the presence of the independent loops in this test reactor. Hence there was need for extensive testing in a special rig completed in 1973 which was described. Mr. N. V. Krasnoyarov introduced the paper describing the development of the refuelling system mechanism on the BN 350 reactor. He outlined the research programme to choose the pairs of bearing surfaces for use in mechanisms under sodium and the work done on out-ofpile test rigs. On the reactor itself, there were extensive tests of the refuelling system both in the hot and cold states. There was no trouble with the actual loading of the real fuel elements Mr. Nishikawa presented the Japanese paper on the fuel handling system for the reactor MONJU. A very comprehensive list of design parameters was given in this paper. The research and development programme on components and equipment was outlined. It was interesting to note that both internal and external clamping systems were being considered for dealing with voidage growth. Mr. E. Benoist presented a paper describing development of the fuel and special handling equipment for P H E N I X and the commissioning tests. The commissioning was performed without major incidents. He attributed this success to the use of simple machines, with separate machines each with few functions. The cost of the test programme in sodium was well justified by the saving in down-time on the reactor itself. The benefits would be even greater on the larger reactors of the future. it was interesting to note that the new fuel elements were put into the reactor and the spent fuel elements were discharged from the reactor via a sloping tube similar to the one used in the BN 350 reactor. Mr. P. L. Riley in his presentation showed slides illustrating the progress of construction of PFR. He described how useful it had been to have a combined test rig built beside the reactor which enabled commissioning tests of the major components to be performed in parallel with construction. He outlined the various difficulties that had been met during construction. The discussion period as well as the papers themselves revealed a widespread interest in the time taken for refuelling. In answer to a question, Mr. Krasnoyarov said that a fuel subassembly could be replaced in BN 350 in 40 minutes. There was also a considerable discussion

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on the special provisions necessary for removing debris. However, it was finally agreed that, although it was necessary to have such operations in mind at the design stage, it would always be necessary to produce special purpose equipment for removing the actual components that had to be dealt with. PLENARY SESSION 4: FUTURE PLANS

The session was begun by Mr. R. Carle who described future programmes and plans for prototypes for CFR's in France. A proposal for a S U P E R - P H E N I X was expected late in 1974. The parameters of this reactor are compared with those of P H E N I X and other reactors in Table 1. Somewhat lower core temperatures were intended in order to increase the burn-up of the fuel. He outlined the various aspects which would have to be investigated in a research and development programme, in order to support the new design. He described the proposals for a somewhat strengthened primary containment and also for a strengthened building to safeguard the entire active circuit against an aircraft crash. A double rotating shield was preferred to a single shield with internal fuel-transfer arms, in anticipation of possible trouble with removal of subassemblies affected by voidage swelling. The French had two different designs of steam generator which they would evaluate before making a final choice. In the oral presentation only, a description was given of a stretched version of P H E N I X of power output 450 MWe. This would have 8 intermediate heat exchangers and the same type of steam generator as P H E N I X . He concluded with some brief economic studies. He noted that the capital cost differential when compared with LWR's might be in the range 7-40$/kWe. The savings on fuel costs would more than offset this disadvantage and the benefit of fast reactors when compared with LWR's would widen, for the likely higher ore costs of the future. Mr. J. W. Crawford described the U.S. fast breeder reactor development programme. He began by noting that, in the U.S.A., there were now over 200 nuclear power plants in operation, under construction or on order. The preponderenc¢ of these were of the light water reactor type (LWR) which operated on the uranium-plutonium cycle. This required that initial use oI breeding must be on the same fuel cycle and explained why the highest priority effort was on the liquid metal fast breeder reactor (LMFBR). He noted two phases in the programme, the first was research and development and the second was utility commitment. The most recent step in the U.S. breeder programme was the initiation of work on the first breeder Demonstration Plant. A partnership agreement between the AEC and two major utility systems was signed in July 1973 to build the 380 MWe Clinch River Breeder Reactor. The paper outlined the work and past contributions of the various laboratories and organizations which contributed to the U.S. L M F B R programme. There was also a brief survey of alternative breeder concepts. These were the light water breeder reactor (LWBR), the molten salt breeder reactor (MSBR) and the gascooled fast breeder reactor (GCFR). The next paper was given by Mr. K. A. Roe, who described the participating organizations of the Clinch River Breeder Reactor project and the management

structure required to co-ordinate their efforts. He noted that a project-orientated sub-division of responsibilities was preferred to the commonly adopted sub-division by discipline. Mr. F. Pierantoni introduced the Italian contribution. He noted that the recent energy crisis and its attendant balance-of-payments problem had given renewed importance to the Italian interest in fast breeder reactors. He noted that it was impracticable for Italy to conduct an entirely independent programme. He suggested that there were two aspects worth participation for Italy: the acquisition of know-how and the promotion of industrial participation. Experience on fuel irradiations would be gained from the PEC reactor and experience in sodium technology would be gained from the reactor and from specially built test rigs. Mr. E. W. Carpenter introduced a paper from the (U.K.) Central Electricity Generating Board giving the perspective on sodium-cooled fast reactors from the point of view of an electric utility. He noted that there were already many economic studies showing the benefits of fast reactors but a utility would require further study on two factors, namely the influence of the longterm price of ore and the capital cost differential between fast and thermal reactors. He commented on the high cost of unreliability. Thus a one month construction delay or 1 per cent loss in load factor were equivalent to a cost increase of about £3/kW. Hence a utility's interest in replaceability, redundancy and tested repair methods. Larger units often failed to give the expected cost saving because of reduced availability. In summing up, it was stated that particular attention must be given to fullscale demonstrations of vital components such as pumps and long-term tests of materials behaviour in typical reactor environments. Mr. J. Moore began his review of the U.K. development programme after P F R with some economic studies. He showed that the introduction of the fast reactor could lead to a saving of over 5,000 tonnes per a n n u m of U308 by the turn of the century. An increase in the ore cost by a factor of 2 from 1973 prices would lead to an increase in generation cost of 5 per cent for thermal reactors but by a negligible amount for fast reactors. He went on to consider the proposals for C F R and the choice of its main parameters. It was to be a conservative design with somewhat lower temperatures than P F R to make a less demanding environment for materials. He completed his paper with a review of the status of research and development in the main technical areas relevant to CFR. Mr. J. Traube presented the paper on the plans for SNR-2, the successor to SNR 300, to be built in 1980. The same group of utilities is to be concerned with S U P E R - P H E N I X and SNR-2. The reactor size, 2000 MWe, was decided by extrapolation of the trend of power station orders in the U.S.A. The larger pin size as in the second charge of SNR 300 and in SUPERP H E N I X was chosen to reduce the fuel fabrication costs. In a final word of caution, Mr. Traube pointed out that full competitivity with LWR's could not be expected until many fast reactor power stations had been built. Governments, utilities and industry would have to bear the very considerable cost of the phase of commercial introduction over a period that would extend beyond construction of the first large demonstration plant. In the discussion periods, Mr. Simon of G G A took

BNES International conference on fast reactor power stations: Conference report the opportunity to make a prepared statement about gas-cooled fast reactors. He pointed out that very little had been said about breeding in fast reactors and pointed out that a helium-cooled, gas-cooled reactor would give improved breeding performance. He put on a few slides which showed what a 300 MWe plant would look like. In answer to a question why current designs of sodium-cooled fast reactors had such a low breeding gain, Mr. Dyos of Westinghouse, from the floor, pointed out that there was a choice in making a design: Did you want a low cost for the first plants or did you want good breeding? The methods of achieving a good breeding gain were well-known. It was a matter of engineering. He believed that a linear doubling time of 8 to 10 years was achievable with oxide fuel and some further improvement might be possible with carbide and nitride fuels. Mr. Wensch, of the USAEC, pointed out that a new report would be issued in June which would contain cost-benefit ~tudies. He said that the recent oil embargo had shown the necessity for self-sufficiency for the U.S.A. within its own boundaries. Currently the U.S.A. was considering only U.S.A.-mined uranium and it had been concluded that this would be scarce in 15 yr. Mr. Hafele added that in his opinion there was no option other than the fast breeder. Mr. Garvey, of AEC Canada, then commented that there were alternatives:--a gas-cooled fast reactor, or a D 2 0 reactor with plutonium recycle could give good neutron economy. Several speakers from the floor contested Mr. Traube's extrapolation of plant size to 2000 MWe in 1980. It was felt that a plateau in the range 1000-1500 MWe might exist for many years. Another questioner then suggested that smaller reactors might be needed for under-developed countries but Mr. Traube replied that the fast reactor did not lend itself to production in small sizes in the range 50-150 MWe. PLENARY S E S S I O N 5: THE STATE OF THE ART The final session was a panel discussion on the state of the art on the introduction of fast reactor power stations into electrical generating networks, under the chairmanship of Mr. W. Hafele (Germany). This provided an opportunity for the panel members to reiterate the important points made in papers presented earlier. It was generally agreed that the recent oil crisis had confirmed the need for nuclear power and the future rise in ore prices made the development of fast reactors essential. Mr. R. H. Campbell (U.K.) suggested that a fast reactor could afford an additional capital cost compared with thermal reactors (because of its low fuel consumption) of £50/kW when ore reached a price of $25 per lb U308. It was, however, noted that the immediate task was to get the prototype reactors working and that existing fast reactors were optimized for low generating cost now. The development of improved breeding performance would come later. The representatives of electricity utilities confirmed their interest in reliability, availability and flexibility. l)r. G. Schuster, Director General of Research, Science and Education in the EEC, who was Chairman in the Clodng Session, noted that the European Community and Japan had been harder hit by the oil crisis than the U.S.S.R. and the U.S.A. He urged that Europe should

471

pull together in a harmonized effort to avoid waste in meeting this crisis. In the closing speech of the Conference, Mr. R. V. Moore (U.K.) summarized the main features and achievements of the work of recent years. He noted that the overall timescale was long and conferences such as this one had been most useful in pooling the knowledge and experience gained. He thanked all those who had participated in making this such a successful Conference. COMMENTARY The Conference was called to mark the start-up of prototype reactors in the U.S.S.R., France and the U.K., and there was widespread interest in the experience gained during construction and operation. The plans for the construction of the next generation of near commercial plants were also described. In most cases, the plans called for reactors of about 1000 MWe size and some doubts were expressed about the need for reactors as large as 2000 MWe which were suggested from Germany. It was felt that the econo,nies claimed for large components were not always realized because of their unreliability and the very large units which thereby were put out of service. Now that fast reactors are at the threshold ~,f commercial acceptance, the influence of electrical utilities is beginning to show. Their knowledge of the cost of replacement electrical power led them to call for availability, reliability and repairability and also, as far as possible, for the full-size test of components. Several speakers noted that therewere two broad policies available for reactor development: the progressive development of a series of reactors or the initial development of a reactor suited for replication. The first strategy implied an earli,~r introduction of reactors with a higher degree of economic risk in the early stages but with the promise of improved designs in the future. The similarities and differences of current and future reactor designs can be gauged from Table 1. The desire virtually to avoid the possibility of sodium loss from the core has led to the adoption of a pool-type of reactor in many countries, in particular in P F R and in P H E N I X as illustrated in Fig. 1 (from the paper by Clauzon et al.). Elsewhere there was interest in the loop-type reactor in which the reactor and major components were in separate tanks, for example in M O N J U as illustrated in Fig. 2 (from the paper by Nakai et al.). This arrangement was claimed to give improved accessibility for repairs. The designers of M O N J U had sufficient confidence in their guard vessels to place access pipes to the reactor tank even below the core level. All reactor designs showed two zones of enrichment in the core to increase the specific power. Boron carbide control rods were the invariable choice and tantalum, as used in one type of control rod in PFR, seemed to have disappeared from future designs. There ~ some interest in europia control rods as a possible choice if there proved to be difficulty with boron carbide rods. In many designs, for safety reasons, two independent designs of control rod were called for. This ,~iew was justified by recent incidents with prototype react~,r~ where faults were found which affected one type of control rod but not the other. It is still one of the outstanding problems of core design

A. R. BAKER

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ing the reactor. The pin size in current reactors was mostly about 6 mm, but larger pins of about 8 mm size were being considered in some cases in order to reduce the cost of fuel fabrication. The pins in all cases were set in a triangular array in hexagonal subassemblies. There was no agreement on the ideal size for these subassemblies nor whether the pins were best spaced by grids or by spirally-wrapped wires. Experience to date suggested that the modes of fuel failure were benign. The Russians particularly had experience of operating reactors with failed fuel (in BR 5 and more recently in BOR 60). As the Russians were prepared to operate a reactor for a month or so with failed fuel elements, it was adequate for them to identify failures only after shut-down. Elsewhere there was interest in detecting the whereabouts of failures as soon as they occurred by delayed neutron sniffing or by a xenon/krypton tagging procedure. The temperature limit for a core was normally expressed as the nominal value of the clad at mid-wall with a maximum in the range 620-650°C. This led to a mixed coolant outlet temperature in the range 530560°C and steam temperatures at the turbine in the range 490-510°C. Thermocouples on the outlets of individual subassemblies on K N K showed their worth in detecting potential blockages as previously they had in FERMI-1 even though, in this case, the indications proved to be spurious. The approach to power station operation on a commercial basis required consideration of operation with part flow (for example, three primary loops out of four

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tubes• All countries contributing papers, except the U.K., had a 1-5 M W steam generator rig operating and a 50 M W rig either commissioned or in building• There was no indication that the material problem had been resolved except that most countries had settled for tubes of 2¼ per cent Cr ferritic steel, either stablized or unstabilized. Although this Conference was not concerned with safety, it was often mentioned in passing• It was clear that safety criteria in many cases were still undecided• There was a difference of view on the need for separate decay heat removal and for provision for natural circulation. There was widespread interest in international collaboration in this area of safety and licensing• Many speakers also suggested international collaboration as a means of reducing development costs and there was a concrete example of what was possible in the collaboration of Prench, German and Italian electrical utilities on the SUPER-PHEN1X and SNR-2 projects• The international oil crisis was enough to ensure a resurgence of interest in liquid-metal fast breeder

474

A. R..BAKER

reactors, which will conserve fissionable-fuel supplies, and to ensure the success of the Conference. The Proceedings will form a valuable record of the state of the art at the time when prototype reactors came into operation in the U.S.S.R., France and the U.K.

Acknowledgements--The author would like to acknowledge assistance with three of the specialist sessions:

Mr. D. Taylor (TNPG) on steam generators; Dr. J. F. W. Bishop (UKAEA) on fuel; and Mr. G. K. Dickson (UKAEA) on coolant management. A. R. BAKER

Fast Reactor Systems Directorate U.K.A.E.A. Reactor Group Risley, Warrinston, Cheshire.