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.     

Response of liquid metal fast breeder reactor containment to a hypothetical core meltdown accident

This paper presents an analysis of the response of the containment building of a 2500-MWt liquid metal fast breeder reactor to a hypothetical reactor core meltdown. Although not mechanistically justifiable, this type of event is chosen for analysis as a basis for risk assessments. Containment space atmosphere compositions, temperatures, pressures, and structural temperatures are calculated, based on decay energy release and chemical reactions associated with the incident. The CACECO containment analysis code, which was used to make the calculations, is described in detail.Results of the study show that by utilizing the passive heat absorption capability of structures normally present in containment design, reactor plant containment integrity can be maintained for more than a day, even for extreme hypothetical events.     

Current perspective on fast reactor containment

The concept of “containment” is to provide a series of physical barriers between the radioactive products of the fission process and the public. All nuclear reactors have several such barriers and LMFBRs have more than most. These barriers are, successively:

1. fuel, which retains fission products;

2. fuel cladding, which encloses the fuel;

3. sodium coolant, which absorbs fission products released through fuel caldding;

4. primary coolant boundary, which has energy absorption and leakage control capabilities;

5. containment building, hereafter referred to as containment, which provides the final engineered barrier for control of radioactive releases;

6. exclusion distance, which provides space for natural attenuation of radioactive releases before reaching the public.

These barriers, along with the design approaches and features which protect their integrity under normal and accident conditions, assure that the public is adequately protected from the potential hazards of radioactivity residing in the core. It is only in the case of hypothesized core disruptive accidents (HCDAs) that these successive barriers can be sufficiently threatened as to pose a significant threat to the public. These HCDAs involve an extremely low probability sequence of successive failures resulting in core cooling imbalances which lead to fuel overheating. Under such conditions, the fuel and cladding barriers can be lost and energy sources can be generated which threaten the primary coolant boundary and containment. This paper addresses current perspectives on containment of HCDAs with emphasis on the approach and programs in the US.     

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.     

A new physics design of control safety rods for prototype fast breeder reactor

The absorber rods of 500 MWe prototype fast breeder reactor (PFBR), which is under construction at Kalpakkam, have been designed to provide sufficient shutdown margin during normal and accidental conditions for ensuring the safe shut down. There are nine control and safety rods (CSR) and 3 diverse safety rods (DSR). Absorber material used is initially 65% enriched B₄C. Based on the reported experiments in PHENIX reactor and design of absorber rods in SUPERPHENIX, the design of CSR is modified by introducing 20 cm length natural B₄C at the top and bottom of absorber column and maintaining the remaining portion with 65% enriched B₄C. This design ensures sufficient shutdown margin (SDM) during normal operation and also during the one stuck rod condition. For comparison of the above two designs, a CSR of 57% of enrichment was considered which gives the same worth as the revised CSR design with natural B₄C sections in top and bottom. There is significant savings in the initial inventory of enriched B₄C for CSR. The annual requirement of enriched boron also reduces. This new CSR can last for about 5 cycles, based on its clad life. But, it is planned to be replaced after every 3 cycles (1 cycle equals 180 efpd) of operation due to radiation damage effects in hexcan D9 steel. Use of ferritic steel for hexcan can extend the life of CSR to 5 cycles.     

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.     

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.     

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.     

Hydraulic design of sparger arrangement for long guide tubes of a fast breeder reactor

In liquid metal cooled fast reactors, the core is submerged in sodium pool by ∼5 m below sodium free surface. This necessitates the control and shutdown of reactor be achieved by long overhanging mechanisms housed inside a control plug. These mechanisms are protected by porous guide tubes with a sparger type arrangement for the sodium flow through them. Comprehensive knowledge of flow distribution of sodium through these guide tubes is essential to assess the risks of flow induced vibration of thin thermowell tubes that pass close to these shroud tubes and entrainment of cover gas due to high free surface velocities. Three dimensional hydraulic analysis of single isolated shroud tube and integrated assembly of shroud tubes have been carried out using CFD tools to acquire this knowledge. The predictions of the CFD models have been validated against experimental predictions. These studies have provided important information regarding critical design parameters. Size of holes in the shroud tube, location of holes in the control plug shell and arrangement for breaking sodium jets emanating from shroud tubes have been optimized to reduce free surface velocity.     

Significant structural design features of commercial-sized liquid-metal fast breeder reactor (LMFBR) plants

From a structural design point of view, there are many significant differences between liquid-metal fast breeder reactor (LMFBR) plants and light water reactor (LWR) plants especially in the containment area, and the area containing the nuclear steam supply system (NSSS). This paper highlights the structural design features of a commercial-sized LMFBR plant.While the external loads are the same, the internal loads, due to characteristic differences in the LMFBR and LWR NSSS systems, have significant structural implications. Primarily, they lead to the low pressures and elevated temperatures in the LMFBRs in contrast to the high pressures and lower temperatures in the LWRs. Also, there are differences in the design basis accident scenarios. These and other differences are addressed.     

Nuclear accident simulation in a 1/6 scale model of the SNR-300 fast breeder reactor

In the frame of the licensing of the SNR-300 fast breeder reactor an HCDA has been investigated both by theoretical and experimental means. The latest phase of this programme was the simulation of a HCDA in a very well instrumented 1/6 scale model of the SNR-300 vessel. The Bethe-Tait accident was simulated by a slow burning charge. In order to check the reliability of the theoretical means allowing for an evaluation of these accidents, the German authorities wanted to receive, before the shot, a theoretical precalculation report showing the main permanent deformations and responses of the pressure and strain gauges versus time.This paper compares and comments on the most meaningful results of this experiment with the SURBOUM-II precalculated estimations.     

Three dimensional aspects of fast reactor containment studies

In assessing the strength of the primary containment of a pool type fast reactor with respect to the dynamic loading from core disruptive accidents (HCDA's), proper account must be taken of the 3-dimensional geometry of the components within the primary containment vessel. This paper reports on a series of experiments and the associated analysis carried out at AEE Winfrith to investigate this aspect of the containment loading. The experiments suggest that asymmetries in the containment loading, induced by symmetric or asymmetric rings of intermediate heat exchangers and sodium pumps, will be small. Calculations performed with the 2-dimensional axisymmetric code SEURBNUK provide a satisfactory estimate of the loads on the IHX's/pumps and their effect on the containment loading.     

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.     

Reactivity insertion mechanisms in the gas-cooled fast breeder reactor

This report summarizes an analysis of reactivity insertion mechanisms in the gas-cooled fast breeder reactor (GCFR). Inherent reactivity feedback mechanisms are identified and their effects on reactor start-up, during normal operation, and on anticipated and postulated transients are analyzed. Potential sources of accidental reactivity insertions and the resulting transients are investigated, including potential reactivity effects due to cladding and fuel melting. All nuclear calculations are based on the ENDF-B, Version 3, cross-section file. It is concluded from these analyses that the GCFR is an inherently stable reactor during start-up and normal operation. Potential accidental reactivity insertions are mild, and in each case the reactor can be controlled with a substantial margin for fuel melting or cladding damage. In low-probability accident sequences which lead to core melting, there are potential fuel motion mechanisms which can mitigate reactivity effects and accident consequences.     

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.     

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。     

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.