Prototype Fast Reactor Research Articles

Gas bubble adhesion to heat transfer surfaces

A model of gas bubble adhesion to heat transfer surfaces is developed, by considering the drag, surface tension and buoyancy forces acting on a gas bubble attached to a solid wall with liquid flowing past. The model is consistent with the fuel pin failures that were observed in the irradiation tests of fuel sub-assemblys for the British Prototype Fast Reactor, and predicts that changing to upward flow will not in itself necessarily prevent such failures in the future.

Procedures used in the UK for the ceramographic examination of (UPu)O 2 oxide fuels for the dounreay prototype fast reactor

Ceramographic information on PFR fuel forms part of a comprehensive data bank established to record significant parameters on fuel, cladding and component materials loaded into the reactor. Collection of the data is essential for assessment of the effects of fuel and material variables on fuel element performance. The methods developed for conversion of ceramographic data into computer compatible form for storage on the data bank, are described. The uniform code of practice used by the various laboratories involved in the manufacture of fuels for the PFR is explained and the methods by which the ceramographic information is fed into the data bank are indicated. The autoradiographic method used to establish plutonium distribution is discussed and the results compared with those from electron microprobe examination of unirradiated (UPu)O 2 specimens. Autoradiographic work on irradiated (UPu)O 2 is also discussed. Examples of Pu re-distribution are given and the results compared with those from EPMA examination of the specimens.

Corrosion behavior of sodium-exposed stainless steels in chloride-containing aqueous solutions

The corrosion behavior of sodium-exposed stainless steels in chloride-containing aqueous solutions was investigated. Results showed that sodium-corroded Type 316 stainless steel (prototypic Liquid Metal Fast Breeder Reactor (LMFBR) fuel cladding) maintains its integrity after five months exposure in these solutions at 82°C and with chloride content up to 500 ppm. In contrast, sensitized and sodium mass transfer deposit-containing Type 304 stainless steel failed in the high chloride solution (500 ppm) within ten days at the same temperature. The failure was initiated by pitting and subsequently accelerated by intergranular attack. The results also show that high pH tends to reduce the susceptibility to failure while procedures commonly used for sodium removal have no significant effect on the water corrosion behavior of the test material. Based on the current results, it is concluded that water storage is feasible for spent fuels in a LMFBR reprocessing plant.

Effective one-group coarse-mesh method for calculating three-dimensional power distribution in fast reactors

An effective one-group coarse-mesh method for calculating three-dimensional power distribution in fast reactors has been developed. This method uses direction dependent one-group diffusion coefficients in each region so as to preserve neutron leakage rates before and after energy condensation. Furthermore, to correct coarse meshes of one point per hexagonal assembly used in the three-dimensional diffusion calculation, one-group cross-sections are modified by applying Askew's method. Based upon this method an effective three-dimensional diffusion code has been made. Results obtained in test cases on a prototype fast reactor indicate that this method is as accurate as six-group fine-mesh diffusion calculations using six mesh points per assembly in each radial plane and the solution of this method is a factor of 40 faster than the six-group fine-mesh calculations.

Western Europe pushing ahead to develop fast breeder reactor

In Western Europe, the U.K. was the first to build and operate an experimental electricity-producing FBR. The French and West Germans have put much engineering effort into FBR design. With the successful completions of Rapsodie and Phenix in France, French and Italian electric utilities are now collaborating to build Superphenix, a full-scale commercial FBR plant. Participants have expanded to West Germany, Belgium, Netherlands, and the U.K. Meanwhile, at Karlsruhe, West Germany, a 200-MW experimental fast breeder, KNK-2, was built in 1968 for fuel-irradiation studies. A 300-MW prototype FBR plant, SNR-300, is now being planned by Schnellbruter Kernkraftwerkgesellschaft. The intention is to install a commercial-scale 1200-MW plant, SNR-2, more or less matching Superphenix. Like Superphenix, SNR-2 would come within the framework of the Europewide international collaborative program. The broadening of the research, development, and operating effort to embrace a large number of countries truly places the FBR development program in a Europewide context. A characteristic of the various FBR development programs is the considerable care that is being exercised to reduce the element of operating risk. The reprocessing plants planned are discussed; at Dounreay, Scotland, a unit is nearing completion for reprocessing spent fuel from an adjacent prototype fast reactor. Itmore » is the first reprocessing plant in Western Europe designed for coping with spent FBR fuel. (MCW)« less

An Effective Method for Estimating Multiple Control Rod Worths in Large Fast Reactors

A method for estimating the multiple control rod worth in large fast reactors has been developed using the higher order perturbation technique. With this method, physical significance of the interaction effect between multiple control rods can be understood by decomposing it into independent components with different physical processes. Furthermore, a new technique is proposed, whereby the interaction effect between multiple control rods can be estimated from a combination of the interaction effects between two or three control rods depending on the strength of the effect. The validity of the method is examined with the experimental data of PHENIX and numerically for MONJU. The results show that the interaction effect between multiple control rods in prototype fast reactors can be estimated from a combination of the effects between 2 control rods with a good accuracy. The numerical examinations for a 1,000 MWe fast reactor showed that the interaction effect is too large to apply the present method for large fast reactors. To overcome this difficulty, a new concept of “quasi-control-rod” is introduced in the method. With this improved method, the worths of 19 and 18 control rods in the 1,000 MWe fast reactor are estimated with accuracies of ±2%.

An Effective Method for Estimating Multiple Control Rod Worths in Large Fast Reactors

A method for estimating the multiple control rod worth in large fast reactors has been developed using the higher order perturbation technique. With this method, physical significance of the interaction effect between multiple control rods can be understood by decomposing it into independent components with different physical processes. Furthermore, a new technique is proposed, whereby the interaction effect between multiple control rods can be estimated from a combination of the interaction effects between two or three control rods depending on the strength of the effect. The validity of the method is examined with the experimental data of PHENIX and numerically for MONJU. The results show that the interaction effect between multiple control rods in prototype fast reactors can be estimated from a combination of the effects between 2 control rods with a good accuracy. The numerical examinations for a 1,000 MWe fast reactor showed that the interaction effect is too large to apply the present method for large fast reactors. To overcome this difficulty, a new concept of “quasi-control-rod” is introduced in the method. With this improved method, the worths of 19 and 18 control rods in the 1,000 MWe fast reactor are estimated with accuracies of ±2%.

Evaluation of Desirable Accuracy for Nuclear Data in Fast Reactors by Optimization Method

The desirable accuracy of nuclear data for minimizing the design margin in fast reactor nuclear designs is evaluated through an optimization method, in which a parameter termed the “degree of difficulty” has been introduced in order to represent the relative degree of difficulty expected to be encountered in improving the accuracy of the data. This parameter serves to determine the desirable accuracy, by which the design margin could be reduced the least onerously, to obviate uselessly severe requirements being imposed on the accuracy demanded of data with high sensitivity coefficients. Application of this method to the equilibrium core of a prototype fast reactor leads to the conclusion that nuclear data that should to best advantage be improved in accuracy are: Furthermore, it is indicated that the ratio between the desirable accuracy and the present uncertainty level of nuclear data is roughly inversely proportional to the square root of the product of the sensitivity coefficient and the present uncertainty level of nuclear data.

Evaluation of Desirable Accuracy for Nuclear Data in Fast Reactors by Optimization Method

The desirable accuracy of nuclear data for minimizing the design margin in fast reactor nuclear designs is evaluated through an optimization method, in which a parameter termed the degree of difficulty has been introduced in order to represent the relative degree of difficulty expected to be encountered in improving the accuracy of the data. This parameter serves to determine the desirable accuracy, by which the design margin could be reduced the least onerously, to obviate uselessly severe requirements being imposed on the accuracy demanded of data with high sensitivity coefficients. Application of this method to the equilibrium core of a prototype fast reactor leads to the conclusion that nuclear data that should to best advantage be improved in accuracy are:σ41f(4000.5 keV), σ49f(1.4 keV) and a=σ28c(1000.5 keV). Furthermore, it is indicated that the ratio between the desirable accuracy and the present uncertainty level of nuclear data is roughly inversely proportional to the square root of the product of the sensitivity coefficient and the present uncertainty level of nuclear data.

Requisite Accuracy for Hot Spot Factors in Fast Reactors

In the thermal design of a fast reactor, it should be most effective to reduce hot spot factors to the lowest possible level compatible with safety considerations, in order to minimize the design margin for the temperature prevailing in the core. Hot spot factors account for probabilistic and statistic deviations from nominal value of fuel element temperatures, due to uncertainties in the data adopted for estimating various factors including the physical properties. Such temperature deviations necessitate the provision of correspondingly large design margins for temperatures in order to keep within permissible limits the probability of exceeding the allowable temperatures. Evaluation of the desired accuracy for hot spot factors is performed by a method of optimization, which permits determination of the degree of accuracy that should minimize the design margins, to give realistic results with consideration given not only to sensitivity coefficients but also to the present-day uncertainty levels in the data adopted in the calculations. A concept of “degree of difficulty” is introduced for the purpose of determining the hot spot factors to be given higher priority for reduction. Application of this method to the core of a prototype fast reactor leads to the conclusion that the hot spot factors to be given the highest priority are those relevant to the power distribution, the flow distribution, the fuel enrichment, the fuel-cladding gap conductance and the fuel thermal conductivity.

A Monte Carlo Code for the Evaluation of the Thermal Design of a Nuclear Reactor Core

A new calculation code, the Hot Spot Probabilistic Evaluation Code (HOSPEC), is presented for evaluating hot-spot factors in a fast reactor. This code calculates the probability distribution of temperature in the whole core by means of a Monte Carlo method. Each Monte Carlo trial involves a complete thermal conduction analysis, thereby reducing errors due to assumptions in analytic procedures currently in use. With this code it is possible to determine the probability that fuel pellets, fuel pins, or subassemblies will exceed the limiting temperature, as well as determine the number of such ''hot spots'' that will develop. A quantitative comparison is made of the results obtained from this code with those from other analyses of a prototype fast reactor. The comparison has indicated, among other points, the following findings: 1. For zero hot spots, a conventional analytic evaluation code SHOSPA gives conservative results, i.e., approximately 20$sup 0$C at the fuel center, at a 3 sigma confidence level. 2. It is of crucial importance to take into account the temperature dependence of the properties of the materials. Neglecting such dependence leads to a much more conservative temperature prediction, e.g., approximately 50$sup 0$C at the fuel center.

The Zebra multi-chamber scanning system

The Multi-Chamber Scanning System developed for the zero-energy fast reactor Zebra comprises an integral array of 150 fission chambers providing a rapid three-dimensional power-mapping facility. The paper describes the basic concept, the design and calibration of the fission chambers, the electronics and computer interface, and overall performance in a mock-up of the Prototype Fast Reactor.

Correlation of Transient-Test Data with Conventional Mechanical Properties Data

AbstractThermal transient-test failure data for proto-typic fast test reactor fuel cladding were correlated with failure data from conventional mechanical properties tests. The correlation was developed by the application of a life fraction rule to stress rupture, tensile and tube burst, as well as transient test data and by constructing a Larson-Miller plot which encompasses the stress and temperature range for all four types of tests. The high degree of correlation between the different types of test data indicates that this approach will be useful in the prediction of failures under hypothetical transient conditions. It is shown that the Dorn parameter can be used instead of the Larson-Miller parameter with equal success for the various types of mechanical properties tests.

IEE Power Division: Chairman's address. The prototype fast reactor Dounreay—some interesting aspects of its design and construction

The address deals with some of the more interesting aspects of the design and construction of the prototype fast reactor at Dounreay. These include the bringing together of the primary-vessel components, all of which are supported from the reactor roof, the construction of which involved the use of flotation. A description is included of the rotating shield and its function related to the charge procedures of the reactor. The use of the irradiated-cave facilities is described and some detail is given of some of the special machines within this complex. The details of the electrical installation have been dealt with in a previous address, and little of that is repeated.

Design and implementation of the prototype fast reactor simulator: Nucl. Eng. Int. Vol 18, No 211 (December 1973) pp 952–956

Design and implementation of the prototype fast reactor simulator: Nucl. Eng. Int. Vol 18, No 211 (December 1973) pp 952–956

The prototype fast reactor at dounreay

Fast-breeder reactors like the Dounreay prototype are going to play an important part in providing future needs. The Dounreay prototype aims to prove its operational reliability and so be the pioneer for larger more powerful nuclear power stations.

LMFBR Neutron Monitoring Systems - A Review of the State-of-the-Art

The French PHENIX, the British PFR and the West German SNR-300 Prototype Fast Reactor neutron monitoring system concepts are reviewed and compared to U. S. systems. Emphasis is placed on the type of neutron detectors and preamplifier electronics being used. The French Series CFUC and the British Type P7 high temperature (~1050°F) fission counters are described and indicated performance discussed based upon published results. High temperature compensated ionization chambers manufactured in Europe and the U. S. are briefly covered. The European low level flux monitoring preamplifier philosophies are related and compared to present and planned U. S. systems. Major differences are noted and discussed. Recent test results from wide range systems mated with fission counters operating at 725°F in the EBR-II Nuclear Instrument Test Facility (NITF) are presented. The NITF facility is briefly described.

Cover-gas bubbles in recirculating sodium primary coolant

The cover gas used over sodium in a fast reactor is more soluble in hot sodium than in cold. Consequently the cover gas will have a tendency to dissolve in the hot sodium leaving the core, and come out of solution in the form of gas bubbles in the cooler part of the circuit. This paper calculates the equilibrium number of gas bubbles, on the assumption of no entrainment at the free surface. Consideration of the mass transfer processes at the free surfaces and in the intermediate heat exchanger is used to give the rate at which gas bubbles are formed. The equilibrium number of bubbles depends on how readily they rise to the free surface and escape, which is discussed. The size of the bubbles used in these calculations is found using a theory previously developed by the author. As an example the theory is applied to the Prototype Fast Reactor. At full power the radius of the bubbles is estimated to be 50 to 100 μm, and the number of bubbles entering the core between 10 4 and 2 × 10 5 m −3. The effect on boiling and detection is discussed.

Fast reactor power

The first fast-reactor nuclear power station in the world, as distinct from experimental assemblies however big these may be, is now in the final construction stages at Dounreay and should be generating power early in 1973. It is the UKAEA's prototype fast reactor (PER) designed to generate 250 MW at first and 300 MW later.

Dounreay prototype fast reactor

The 1.5 MW experimental fast reactor at Dounreay has been operating since 1959; the UKAEA is now building a 250 MW prototype fast reactor there, which is due for completion in 1972, and should act as a stepping stone to the design and construction of large commercial fast reactors.