Status of gas-cooled fast breeder reactor programs

Most gas-cooled fast breeder reactor (GCFR) programs in Europe and the US are now coordinated and focused on a 300 MW(e) GCFR demonstration plant program. Except for venting and artificial surface roughening, GCFR fuel is similar to liquid metal fast breeder reactor (LMFBR) fuel and operates under nearly identical conditions. The primary helium system is integrated within a PCRV like all large gas-cooled thermal reactors, with three main loops and three auxiliary loops. Design and safety studies and various experiments, including heat transfer, irradiation, and critical experiments, indicate that most feasibility questions have been answered and a demonstration plant could be in operation within 12 years. This could be followed in the mid-1990s by a large-size GCFR with a doubling time of about 10 years fueled by (UO₂---PuO₂) and producing either ²³³U in thorium blankets as fuel for advanced converters or plutonium in depleted uranium blankets.     

Safety features of the gas-cooled fast breeder reactor

The safety features of the gas-cooled fast breeder reactor (GCFR) are described in the context of the 300-MW(e) demonstration plant design. They are of two general types, inherent and design-related. The inherent features are principally associated with the helium coolant and the nuclear coefficients. Design-related features influencing safety include shutdown systems, residual heat removal systems, method of core support, and the prestressed concrete reactor vessel (PCRV). This paper discusses the safety-related aspects of each of these. Recently completed residual heat removal system reliability studies are also discussed. The probability of residual heat removal system failure in the GCFR is found to be lower than that described for light water reactors. The safety characteristics of larger plants are examined, and increases in size are found to improve GCFR safety margins.     

The gas-cooled fast breeder reactor critical experiments program

This paper presents results of measurements and calculations of physics parameters in the first gas-cooled fast breeder reactor (GCFR) critical assemblies in the US, a program of experiments conducted on the ZPR-9 facility at Argonne National Laboratory. Through a progressive three-phase series of assemblies, the major features unique to GCFR physics due to the gaseous coolant, and the resulting hard neutron spectrum and greater leakage, were investigated. Phases I and II were simple-geometry, uniform-core assemblies providing tests of nuclear data and GCFR design methods for fast reactors with large void fractions. The Phase III core simulates a GCFR design with three enrichment zones. This report primarily concerns the results obtained in Phase II.In addition to the usual central indices, reaction rate mappings, etc. these initial studies have provided the first experimental data on reactivity coefficients relevant to GCFR safety, such as worths of fuel, control, and cladding materials, Doppler effect, and coolant (helium) depressurization worth. Effects of steam ingress into coolant channels (due to a hypothesized steam generator leak) were simulated using polyethylene. The physics information obtained is providing a valuable base for verification of GCFR design and safety analyses.     

Development of components for the gas-cooled fast breeder reactor program

The gas-cooled fast breeder reactor (GCFR) component development program is based on an extension of high temperature gas-cooled reactor (HTGR) component technology; therefore, the GCFR development program is addressed primarily to components which differ in design and requirements from HTGR components. The principal differences in primary system components are due to the increase in helium coolant pressure level, which benefits system size and efficiency in the GCFR, and differences in the reactor internals and fuel handling systems due to the use of the compact metal-clad core.The purpose of this paper is to present an overview of the principal component design differences between the GCFR and HTGR and the consequent influences of these differences on GCFR component development programs. Development program plans are discussed and include those for the prestressed concrete reactor vessel (PCRV), the main helium circulator and its supporting systems, the steam generators, the reactor thermal shielding, and the fuel handling system. Facility requirements to support these development programs are also discussed. Studies to date show that GCFR component development continues to appear to be incremental in nature, and the required tests are adaptations of related HTGR test programs.     

Roughening characteristics and choices for the gas-cooled fast breeder reactor

The thermohydraulic performance of several types of rough surfaces proposed for use in the gas-cooled fast breeder reactor has been investigated experimentally at the Swiss Federal Institute for Reactor Research. Based on the tests, the most suitable roughness design has been defined. In addition to the thermohydraulic performance requirements, some other technological and operational criteria should be used for the final choice of roughness. There is not sufficient information on the different roughening methods to enable any decision to date, but when the new complex thermohydraulic performance criterion is considered, additional requirements become relatively more important.     

Fuel element design for the 300 MW(e) gas-cooled fast breeder reactor

The fuel element design for a 300 MW(e) gas cooled fast breeder reactor (GCFR) is presented. The design is the result of a program sponsored by Kernforschungsanlage, Julich (KFA) to develop and fabricate a full size fuel element model under extension of an agreement between General Atomic (GA), Kraftwerk Union (KWU), and KFA to exchange information from GCFR irradiation experiments. The resulting fuel element model design was achieved by joint participation between GA and KWU and relies on the experience and knowledge of the two companies. The model, which will be manufactured by KWU using prototypical materials and specifications, except for dummy fuel pellets, will establish manufacturing feasibility and identify areas for future cost reduction improvements. The evolved designs, particularly the fuel rods, are very similar to those employed in the liquid metal fast breeder reactor (LMFBR) programs. These similarities enable the GCFR to use the vast amount of data being generated for the LMFBR programs, with only an incremental development plan needed to verify certain unique features inherent to the use of helium as the primary coolant.     

Seismic response of gas-cooled fast breeder reactor core structural assembly via modal synthesis

An investigation has been conducted to determine theoretically the dynamic response of the GCFR core support structural assembly when subjected to boundary excitation from seismic disturbances. The system analyzed consists of a thick grid plate to which many core elements are vertically attached. The dynamic problem was solved by synthesizing component modes of two substructures and treating them as continuous subsystems. The investigation is of practical significance in the sense that the radial responses of the core elements in axisymmetric motions cause reactivity change of the core, and therefore an accurate assessment of the dynamic response of the system is important to the core and core support structure design. Numerical system modal data and time-history response results are presented.     

Response of gas-cooled fast breeder reactors to depressurization accidents

The gas-cooled fast breeder reactor (GCFR) under design by Gulf General Atomic is cooled with helium pressurized to 85 atm and has the reactor core, the steam generators and their associated steam turbine-driven helium circulators, and auxiliary core cooling loops all contained within a massive prestressed concrete reactor vessel (PCRV).The response of the GCFR to coolant depressurization accidents has been investigated and it has been shown that this class of accidents can be safely handled with considerable safety margin. Rapid depressurization is assumed to be caused by a seal failure in a large concrete plug closing one of the large PCRV cavities and the depressurization rate is controlled by a flow restrictor incorporated within the closure plug. Continued core cooling is provided by the main core cooling loops. The plant transient reponse following a depressurization accident has been calculated with a computer code developed at GGA. The results obtained indicate rather mild increases in peak clad temperature for a depressurization accident with the leak area defined by the flow restrictor.Additional cases investigating larger leak areas to explore safety margins indicate that the peak cladding temperature does not increase rapidly with increasing leak area. Secondary containment conditions in a depressurization accident have also been evaluated.     

Contributions to the neutronic analysis of a gas-cooled fast reactor

In this work the Monte Carlo codes MCNPX and TRIPOLI-4 were used to perform the criticality calculations of the fuel assembly and the core configuration of a gas-cooled fast reactor (GFR) concept, currently in development. The objective is to make contributions to the neutronic analysis of a gas-cooled fast reactor. In this study the fuel assembly is based on a hexagonal lattice of fuel-pins. The materials used are uranium and plutonium carbide as fuel, silicon carbide as cladding, and helium gas as coolant. Criticality calculations were done for a fuel assembly where the axial reflector thickness was varied in order to find the optimal thickness. In order to determine the best material to be used as a reflector, in the reactor core with neutrons of high energy spectrum, criticality calculations were done for three reflector materials: zirconium carbide, silicon carbide and natural uranium. It was found that the zirconium carbide provides the best neutron reflection. Criticality calculations using different active heights were done to determine the optimal height, and the reflector thickness was adjusted. Core criticality calculations were performed with different radius sizes to determine the active radial dimension of the core. A negative temperature coefficient of reactivity was verified for the fuel. The effect on reactivity produced by changes in the coolant density was also evaluated. We present the main neutronic characteristics of a preliminary fuel and core designs for the GFR concept. ENDF-VI cross-sections libraries were used in both the MCNPX and TRIPOLI-4 codes, and we verified that the obtained results are very similar.     

Aseismic study on the reactor vessel of a fast breeder reactor

This paper presents the results of a seismic study using an scale steel model and a scale plastic model which simulate the reactor vessel of a loop type Fast Breeder Reactor (FBR). The main purposes of this study are to confirm the structure/liquid interaction and the aseismic safety of the reactor vessel experimentally, and also to verify the validity of the seismic response analysis model of the prototype vessel.The characteristics of coupled vibration between the structure and liquid were clarified, and the approach of calculation model to aseismic design was worked out. And, the dip plate and other core internals were found to be effective in suppressing the liquid free surface oscillation.     

Shear-bending buckling analyses of fast breeder reactor main vessels

The purpose of this paper is to describe the computation results and the knowledge of the buckling analysis strategy. Since fast breeder reactor main vessels are thin shell structures, plastic shear-bending buckling is one of the most important problems. To clarify the buckling behaviour, we carried out many tests and numerical calculations. Based on the experience of those buckling analyses, available elements, mesh division, modelling of shape imperfections etc. are described. These results show that the numerical analysis can be a useful tool for evaluating buckling phenomena.     

Inservice inspection of the reactor block of sodium-cooled fast breeder reactors

By using sodium as coolant special boundary conditions result for the inservice inspection (ISI) of fast breeder reactors. For that reason in general it is not successful applying the methods and equipment proved for the 151 of light water reactors.This report presents inspection methods and equipment developed for the ISI of the reactor block of sodium cooled fast breeder reactors. The survey takes into account the state of the art as well as some R&D-work at home and abroad. Entering into particulars the methods and equipment used for leak monitoring, the inspection of the reactor vessel wall, the inspection' of reactor internals above and below the sodium level, monitoring of structure home noise and the measurement of the gap between the reactor vessel and the guard vessel are described.     

Investigations on the integrity of fast breeder reactor piping and components

Results of investigations made with respect to the integrity of LMFBR (Liquid Metal Fast Breeder Reactor) piping and components are presented. The classification of sodium systems as Moderate Energy Fluid Systems is shown to be an important principal element to determine the failure mechanisms which are relevant. Based upon the selection of materials, design features and high-quality engineering standards the evaluation of the crack growth morphology of surface flaws contribute to the ensurance of the structural integrity. The crack shape development for bending stress distribution over the wall-thickness, which is a typical loading of FBR structures is discussed. It has been shown that even for such unfavourable loading conditions the through crack lengths are bounded. There is a considerable distance from critical crack configurations calculated by tearing modulus concept. Results from a large scale elbow test at operating temperature are reported. They contribute to the crack shape development under fatigue loading with bending type stress distribution over the wall thickness and are in good accordance with calculations. Acceptance criteria for flaws in structures are proposed showing that the structural integrity for coolant boundary and components of FBR can be assessed with a high degree of reliability.     

Irradiation performance of sodium-bonded control rod for the fast breeder reactor

The lifetime of control rods is limited by the absorber (B₄C pellets)–cladding mechanical interaction (ACMI). Therefore, sodium (Na)-bonded control rods were developed for long-life control rods. Na-bonded control rods have been irradiated in the experimental fast breeder reactor, JOYO MK-III, and the diametrical changes of the Na-bonded absorber pins after the irradiation were measured in detail.

In this paper, these detailed measurements were compared with the results obtained in helium (He)-bonded control rods with and without the shroud tube in a wide burn-up range. From the comparison, it was found that the diametric changes were smaller in the Na-bonded absorber pins than in the He-bonded ones. It was concluded that the Na-bonded absorber pins are very effective for achieving long-life control rods.     

铅冷快增殖堆物理和热工水力研究

对25MW电功率铅冷快增殖堆堆芯进行了物理和热工水力概算，并将计算结果与相同功率的钠冷快增殖堆的结果进行了分析比较。从初步概算的结果来看，铅冷快增殖堆是一种安全可行的快增殖堆堆型。     

Random thermal fatigue in fast breeder reactor: a narrow-band spectrum

Thermal fatigue crack growth in a fast breeder reactor is theoretically investigated with the aid of probabilistic fracture mechanics (PFM) under the conditions that (i) the temperature variation is a narrow-band stationary process and (ii) the crack grows owing only to the peak stress variation. First, a statistical property of residual life of the component with single crack is derived in an analytical form with the aid of an extended Markov approximation method, which is an efficient mathematical technique in PFM. Next, discussion is carried out on the generalization of the primitive model to the case with plural cracks, where a stress relaxation factor is introduced to express a stress intensity factor of each crack. Finally, a numerical example is shown to examine the quantitative behavior of the component's residual life, and sensitivity analysis is performed with respect to some model parameters.     

Gas-cooled fast breeder reactor designs with advanced fuel and cladding

This paper examines the potential impact of some alternative cladding and fuel materials being considered for the liquid metal fast breeder reactor (LMFBR) on the performance and design of large commercial gas-cooled fast breeder reactors (GCFRs). Mixed carbide fuel and Inconel 718 cladding material were examined. Another cladding alternative considered was silicon carbide (SiC), which presents some interesting possibilities in high-temperature performance. Design concepts based on the above fuel and claddings were examined and compared with a reference oxide/316 stainless steel design based on a commercial 4000 MW(th) [1500 MW(e)] system. Substantial benefits can be derived from a high-temperature cladding material such as Inconel 718 or ¹⁶Pe; core volume and steam generator heat transfer area could be reduced by 20% or more, and significant reductions in core inventory and doubling time are possible. Carbide fuels would reduce the number of fuel rods by 50% because of higher linear power, and doubling time would be lowered.     

Gas-cooled fast breeder reactor fuel bundle irradiations in the BR2 helium loop

In the BR2 helium loop at Mol, Belgium, a 12-pin test fuel element of gas-cooled fast breeder reactor (GCFR) design and materials will be irradiated at a 500 W/cm maximum pin rating and a 700°C maximum cladding temperature to a target burnup of 60 MWd/kg (extension to 100 MWd/kg is intended). The design of the test element and the loop is described in detail. To fabricate the test element, parts of the GCFR fuel development had to be anticipated. Preliminary out-of-pile testing was successfully performed, and irradiation is scheduled to start in early 1977 and will be completed between mid-1978 and mid-1979, depending on the final burnup objective. GCFR operating conditions will be completely simulated except for the full size of the fuel element and the fast neutron flux. In combination with out-of-pile performance testing of full-size dummy elements and fast flux experience from the liquid metal fast breeder reactor program, the helium loop irradiation is regarded as an adequate basis for the design of a fuel element for a GCFR demonstration plant serving as the final test bed.