Fast Power Reactors Research Articles

Fundamental study of earthquake isolation systems suitable for buildings of industrial facilities. 1st report Performance of earthquake isolation systems using hysteretic dampers.

For industrial facilities, such as computer centres, semiconductor manufacturing factories and fast breeder reactor power stations, earthquake isolation systems used for buildings should not only attenuate the seismic response of the buildings but also reduce the response of the internal equipment. From this point of view, isolation systems using various types of damper have been evaluated through both experimental tests of a 0.3 scale model of a building supported by rubber bearings and theoretical analysis, to find the most suitable systems for industrial facilities. The performance of two types of isolation system using friction dampers and lead shear dampers respectively was examined. The results showed that both systems using these hysteretic dampers could attenuate the response accelerations of the building and limit the excessive deflexion of rubber bearings. However neither system, particularly the system using friction dampers, was less effective in reducing the response acceleration of the internal equipment.

Fast reactor progress—slow but sure

The article sketches the history of fast reactor power station development showing how the technical development has been influenced by changes in the expectations of growth of electrical power consumption. All fast reactor development programmes have been affected by the present economic situation and the ways in which individual countries have reacted is examined. The capital cost of fast reactors is a key factor governing their rate of introduction. Areas of possible design improvements which might lead to cost reductions are outlined. Finally there is a brief review of some of the characteristics which contribute to the overall high safety of the fast reactor system and the areas in which future work may further enhance safety are identified. It is concluded that the capital cost of the reactor and the future supplies of uranium are the main factors affecting the rate of introduction of fast reactor power stations.

4560533 Fast reactor power plant design having heat pipe heat exchanger : Paul R Huebotter, George A McLennan assigned to The United States of America as represented by the United States Department of Energy

4560533 Fast reactor power plant design having heat pipe heat exchanger : Paul R Huebotter, George A McLennan assigned to The United States of America as represented by the United States Department of Energy

Earthquake isolation technology for industrial facilities

In Japan earthquake isolation technology has become increasingly necessary with the growing dominance of so-called "high technology" industries and is now being used to provide effective aseismic protection for precision equipment which is of major importance in such countries. The present applications are for computer systems and semiconductor manufacturing equipment. For computers, particularly those of banks, earthquake isolation of the floors is widely used. For semiconductor manufacturing equipment, earthquake isolation has begun to be used for the components; the isolation of the floors will follow. For both applications, earthquake isolation has not yet been used for entire buildings, but is expected in the near future. Promising applications for the future include bio-technology facilities, highly automated factories and nuclear plants, especially fast breeder reactor power stations.

Physics and engineering aspects of fast reactor safety

A vast amount of research and development work has been done in recent years to resolve the issues of relevance to the safety of the Liquid Metal cooled Fast Breeder Reactor (lmfbr). Based on the results of this research as well as on the experience gained from the operation of test and prototypelmfbr’s, a certain consensus is emerging on the safety requirements of a modern sodium-cooled large fast power reactor. The paper reviews the fundamental physics and engineering aspects oflmfbr safety with reference to the Fast Breeder Test Reactor (fbtr) now being commissioned at Kalpakkam, and the proposed larger Prototype Fast Breeder Reactor (pfbr). The elements contributing to the inherent safety of fast reactors are recapitulated followed by description of the philosophy of the plant protection system and the use of engineered safeguards to enhance the safety. Finally, the principles used for the containment of radioactivity are discussed.

The U.S. fast reactor materials and structures program

Materials and structures problems are central to many critical issues concerning the economic competitiveness, reliable performance, and safety of liquid metal fast breeder reactor (LMFBR) power plants.

Worths and Interactions of Simulated Fast Breeder Reactor Control Rods in ZEBRA and SNEAK Assemblies

Measurements of the worths of simulated control rods for fast power reactors have been made in the ZEBRA and SNEAK critical assemblies by the modified source multiplication method (MSMM). The assemblies used were the conventional and unconventional core arrangements from the BIZET program and a compacted version of a conventional core. The control rods were mainly natural B/sub 4/C, with some study of 40% /sup 10/B-enriched B/sub 4/C and of Eu/sub 2/O/sub 3/. Correction factors for the MSMM were obtained from eigenvalue and source-mode diffusion theory calculations in XY geometry. The measured rod worths and interactions are compared with calculated values from methods and data similar to those used by the different participants in the BIZET program to predict the corresponding parameters in fast power reactors. In general, acceptable agreement is found.

Paths for the development of fast power reactors with a high breeding factor

Paths for the development of fast power reactors with a high breeding factor

Double Heterogeneity Effect of Fuel Pin and Subassembly in a Fast Power Reactor

The double heterogeneity effect due to the fuel pin and the subassembly is estimated for neutronics parameters of a prototype fast power reactor. Both of the heterogeneity effects caused by the flux fine structure and the resonance shielding are taken into account. The model of the hexagonal unit subassembly consists of the smeared fuel, the wrapper tube, and the outer sodium regions, where the average cross sections of the smeared fuel region are obtained by a unit pin cell calculation in cylindrical geometry. It is shown that the equivalent smeared model agrees well with the refined cluster model. A formula to define the background cross section of the fuel isotopes is derived based on the rational approximation, and its accuracy is examined. This formula can take into account the contribution from the wrapper tube to the background cross sections of each ring of fuel pin array. The heterogeneity effect of the whole reactor model is calculated based on two-dimensional diffusion theory and perturbation theory. The double heterogeneity effect is found to be 0.5%Δk for keff, the positive sodium-void worth is reduced by 26%, and the negative Doppler reactivity increases by 7% for a prototype fast breeder reactor. These results are considerably larger than the estimates made by earlier workers. These effects are significant from the viewpoint of target accuracy goals in fast power reactor design.

ANEXDI—A Two-Dimensional Coupled Thermohydrodynamic and Point Kinetics Computer Model for Analysis of Extended Disassembly Accidents of Fast Power Reactors

A computer code, ANEXDI (analysis of extended disassembly), has been prepared for scoping studies of hydrodynamic interactions in typical core disruptive accidents in a fast power reactor. A two-ph...

Results from an International Intercomparison of Fundamental Mode Benchmark Calculations for Steam Ingress into Gas-Cooled Fast Reactor Cores

Steam ingress into a gas-cooled fast reactor (GCFR) core may lead to reactivity effects that are undesirable from the point of view of reactor safety. Unfortunately, the amount of reactivity increase caused by a certain steam concentration is usually subject to considerable uncertainty, as has become evident by occasional comparisons between various laboratories for specific examples. Therefore, some time ago, a series of intentionally simple benchmarks were proposed in order to study in a systematic way the calculational uncertainty of the steam ingress reactivity arising essentially from differences in the nuclear data basis used at various laboratories. The analysis of corresponding results provided by laboratories in France, Germany, Japan, Switzerland, and the United States reveals that there still exist appreciable deviations in the predicted steam ingress reactivity effect. Due to the extensive cancellation of positive and negative contributions to this reactivity effect, the resulting net value is extremely sensitive to deviations in the nuclear data and calculational methods.Typical discrepancies for the calculated steam ingress reactivity observed within the framework of an international intercomparison are described, leading to the conclusion that further improvements in the nuclear data basis are desirable and the development and application of fairly refined calculational methods is mandatory to be able to predict the corresponding effect with sufficient reliability for related power reactor designs. In addition, measurements of equivalent reactivity effects should be continued in different critical assemblies to provide a broader experimental basis for the verification of the calculational tools. If further analytical work could be pursued, the Argonne National Laboratory experiment on the GCFR Phase II Steam Entry Effect might be the appropriate object to be studied and analyzed in detail, e.g., by a similar intercomparison effort, especially if the discrepancies existing at present in nuclear data bases could be removed or diminished to a tolerable level. Reasonable progress in these areas would increase the confidence attributed to calculations of the reactivity effect of the assumed entry of hydrogeneous material into the core of a fast power reactor.

Atomic Power Safety: The Case of the Power Reactor Development Company Fast Breeder, 1955-1956

Controversy over the safety of atomic power, dramatized by demonstrations at reactor sites; by countless articles, editorials, and speeches; by the popular film The China Syndrome; and, above all, by the accident at Three Mile Island, has made atomic energy a national issue. Polemical arguments are many, but few writers have devoted much attention to the early history of America's atomic policies or placed those policies in the context of their times. Current debate stems from the 1950s, when the government first opened atomic power to development by private enterprise. Under the 1954 Atomic Energy Act, the United States Atomic Energy Commission (AEC) acquired a dual role: promoting private-sector electric power generation by atomic reactors, and licensing and regulating those plants to protect the public health and safety. When this duality was established, it appeared necessary because of limited technical knowledge and qualified personnel. But issuance by the AEC of a construction permit for the Power Reactor Development Company's fast breeder power reactor soon embroiled the agency in a controversy that underscored the difficulties of promoting privately owned power reactors while at the same time assuring adequate health and safety protection to the public. The episode focused public attention on the agency's licensing process for atomic power plants, particularly on the manner in which the AEC resolved safety questions. Equally important were the ramifications for the future development of the agency's regulatory system.1

On the need for changing the ENDF/B convention for the representation of cross-sections in the unresolved resonance region of fertile and fissile nuclei

A more fundamental scientific basis for the prediction of the Doppler effect in fast reactor systems demands that the usual ENDF/B and KEDAK conventions for the representation of cross-sections in the unresolved resonance region should be changed. This conclusion is based upon the experience acquired in evaluations of statistical mean resonance parameter sets, made at RRC Kalpakkam, for use in neutronic calculations of fast reactors in the recent past for 235U, 239Pu, 232Th, 233U and 238U. This paper gives a brief account of several sensitivity studies and neutronic calculations made with these mean resonance data sets. The existence of non-uniqueness of mean resonance data sets has been clearly established and an uncertainty due to the choice of the mean resonance data set is found to be associated with the theoretical calculations of neutronic parameters of reactor systems. This uncertainty, identified at Kalpakkam, has not received attention in the past and is found to be significant in the case of the calculation of the Doppler coefficient in fast reactors. It should have been taken into account in the interpretation of the Doppler effect experiments performed in fast critical facilities for both fissile and fertile samples. The present method of statistical representation of cross-sections leads to a higher statistical uncertainty in the prediction of the Doppler reactivity effect under coolant voided conditions in fast power reactors than in the normal case. Also the effect of inclusion of intermediate structure in the fission cross-section for 239Pu on the temperature derivative of self-shielding factors is significant and cannot be modelled satisfactorily at present. If the high resolution cross-section measurements, as suggested by de Saussure and Perez, are made available for the main fissile and fertile isotopes, thinning and direct Doppler broadening of the cross-section data by the preprocessing methods developed by Cullen and his coworkers appear to be a promising approach. Until such time the evaluated nuclear data files should attempt to contain self-shielding factors directly evaluated with the support of self-indication and transmission measurements instead of the present convention of having mean resonance data sets in the unresolved resonance region.

Development of the scientific-technological bases of fast power reactors

In the next decade work on fast reactors will be concentrated around the BN-1600 and BN-800, based on and with consideration of the experience in operating the BN-350, BN-600, and BOR-60. The weight of the experience of developing fast reactors with sodium cooling confirms that this scientific-technical trend is based on well developed and substantiated technical decisions. The practical steps for the improvement of the characteristics and modernization of nuclear power stations of this type are clearly visible.

Fuel Pin and Subassembly Heterogeneity Effect on Neutronics Properties of Fast Power Reactor

Fuel Pin and Subassembly Heterogeneity Effect on Neutronics Properties of Fast Power Reactor

Fuel pin and subassembly heterogeneity effect on neutronics properties of fast power reactor.

Fuel pin and subassembly heterogeneity effect on neutronics properties of fast power reactor.

Comparison of finite differences and finite elements in the case of a large fast power reactor

A large number of test calculations in two-dimensional hexagonal geometry for different configurations of SUPER-PHENIX 1 type Fast Reactor have been performed to compare finite difference theory vs finite element theory performances. At present, no definitive advantages were found for the application of the finite element method to two-dimensional hexagonal design calculations.

Estimate of the errors in calculating the criticality and breeding coefficients of fast power reactors due to inaccuracy of the neutron data

It should be noted that the previous estimates of the achievable level of accuracy referred roughly to 1970, and the discrepancy in the estimates of the accuracy of the calculated predictions are associated first and foremost with the difference inthe estimates of the indeterminacies of the initial group cross sections. The spectrum of the sensitivity of keff and CB towards different forms of the neutron cross sections in the present paper and, e.g., in [4-6], as a comparison shows are identical. Thus, changes in the estimates of the accuracy of the calculated predictions for kef f and CB of a fast breeder-reactor are dependent upon the consideration of the extensive data obtained during the last decade in differential neutron-physics experiments. Moreover, the predicted values of the achievable accuracy of the neutron cross sections in many cases are of a qualitative nature and are based on approximate estimates of the accuracy of the cross sections according to certain "recommendations ," based on agreements between experts [12]. They do not take into account all the sources of errors in the differential measurements, and also the effect of a specific method of estimating the cross sections, on which the correlation properties of the errors in the group cross sections may depend. The estimate of the accuracy of the calculated predictions of kef f and CB obtained in this present paper is based on the latest data of differential neutron-physics experiments. The feasibility of this accuracy has emerged because of the development of a system of group cross sections BNAB-78 [13], based on which the system of cross sections BNAB-MIKRO [14] was proposed, compiled from new estimates of the results of differential neutron data measurements. In parallel with the estimates of the most probable values of these data, their errors were also determined, which were represented in the form of a eovariational matrix W of the group cross sections, i.e., in a form suitable for practical application. It is important to emphasize that the covariational matrix describes the accuracy of the group cross sections BNAB-MIKRO: the errors of the experiments themselves are taken into account in it, and it is on these errors that this system of cross sections has been compiled. The BNAB-78 system differs from the BNAB-MIKRO system by taking account of the results of macroexperiments . Its use stipulates a higher accuracy of the calculated results [14].

Procedure for calculation of fuel depletion to determine the physical characteristics of a fast power reactor in the steady-state mode

Procedure for calculation of fuel depletion to determine the physical characteristics of a fast power reactor in the steady-state mode

Flow Behavior of Volume-Heated Boiling Pools: Implications with Respect to Transition Phase Accident Conditions

Observations of two-phase flow fields in single-component volume-heated boiling pools were made. Photographic observations, together with pool-average void fraction measurements, indicate that the churn-turbulent flow regime is stable for superficial vapor velocities up to nearly five times the Kutateladze dispersal limit. Within this range of conditions, a churn-turbulent drift flux model provides a reasonable prediction of the pool-average void fraction data. An extrapolation of the data to transition phase accident conditions suggests that intense boilup could occur where the pool-average void fraction would be >0.6 for steel vaporization rates equivalent to power levels >1% of nominal liquid-metal fast breeder reactor power density. The extended stability of bubbly flow to unusually large vapor fluxes and void fractions, observed in some experiments, is a major unresolved issue.