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.     

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.     

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.     

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.     

Gas-cooled fast breeder core element fabrication technology

The status of gas-cooled fast breeder reactor (GCFR) core element fabrication technology and testing is described. Special GCFR requirements arise from use of a high-pressure helium coolant. Ribbing of the cladding is employed to improve heat transfer, and venting and pressure equalization are utilized to avoid creep collapse of the cladding. Fabrication development, including fabrication of a full-scale core element model, has revealed no difficult fabrication or inspection problems, and testing of components has not indicated any feasibility problems.     

The hexagonal bundle heat transfer and fluid flow experiment agathe hex

Problems of heat transfer and fluid flow in gas-cooled reactor fuel elements have been studied at the Swiss Federal Institute for Reactor Research (EIR) for 14 years. Since 1967, the activities have been directed toward gas-cooled fast breeder reactors (GCFRs). The aim of analytical and experimental studies has been to develop analytical models and comprehensive computer codes for the prediction of temperature and pressure distributions in GCFR fuel element configurations. The models developed at EIR are based on the results of specific experiments. Full-scale experiments in actual geometry are being carried out to verify the computer codes for a wide range of parameters. This paper describes the heat transfer loop and the test sections designed to verify GCFR thermohydraulic design codes.     

Fission products and secondary gamma-ray spectra in a plutonium-fueled fast reactor

The fission products' gamma-ray and gamma-ray energy source spectra for a gas-cooled fast breeder reactor (GCFR) are calculated for different times after shutdown by modifying the RIBD computer code. The secondary gamma-ray energy source spectrum in the core of a GCFR, from fission, inelastic scattering, and capture reactions, is calculated using a typical GCFR neutron spectrum. The computer code LAPHANO is used to generate the multigroup (n, xγ) neutron-coupled gamma-ray transfer matrix. The weak dependence of capture and inelastic gamma ray source spectrum on the neutron flux spectrum has been noted. The fission products and secondary gamma-ray source spectra obtained can be used to calculate heat generation and refueling shielding requirements, etc.     

Materials and design concepts of gas-cooled reactor systems

The family of gas-cooled reactors being developed in the United States by Gulf General Atomic consists of the steam-raising and direct cycle versions of the high temperature gas-cooled reactor (HTGR) for electric power generation, the hydrogen-producing HTGR for chemical process applications, and the gas-cooled fast reactor (GCFR), a high gain breeder. The aim of this paper is to describe the underlying design concepts that are common to all of these reactors and relate these design concepts to the choice of both structural and fuel materials for the wide variety of environmental conditions encountered throughout the world. Interwoven with this discussion are typical examples of the interaction of design activities and materials selection required to give a reactor system of maximum safety and reliability, favourable environmental features, and minimum cost.     

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.     

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.     

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.     

高温气冷堆核级石墨相关问题研究

本文通过对石墨在高温气冷堆中的运行环境进行了分析,研究了在石墨堆内构件设计中的关键问题和在高温气冷堆单个模块及其未来发展中核级石墨的需求。从原料、成型及中子辐照等角度分析了核级石墨国产化研究方向。根据核级石墨目前的研发形势,进行了风险问题分析。     

An analysis of GCFR depressurization transients without scram

Four transients imposed on a gas cooled fast breeder reactor (GCFBR or GCFR) plant are analyzed. The transients are variations of a design basis depressurization accident (DBDA) [1], in which a rupture is postulated in the inlet plenum, the reactor is scramed and the circulator output is drastically reduced. Variations considered are, (1) in all cases the reactor is not scrammed; and (2) the rupture size is varied from 0.6 ft² to zero. In the limit of zero (no rupture) the transient imposed on the system is due to the behavior of the circulator and steam generator during a DBDA.     

Prediction calculations for the first criticality of the HTR-PM using the PANGU code

The high-temperature reactor pebble-bed mod-ule (HTR-PM) is a modular high-temperature gas-cooled reactor demonstration power plant. Its first criticality exper...     

Analysis of stresses in GCFR cladding under normal operating conditions

A modified version of the LIFE-III code, LIFE-GCFR, and classical stress analysis techniques have been employed to calculate the stresses in GCFR cladding under normal reactor operating conditions. Several types of loadings on the cladding which occur during normal operation have been considered. These include fuel-cladding mechanical interaction, thermal stresses induced by radial and axial temperature gradients, and stresses induced by swelling gradients. The combined and individual effects of these loadings as well as the effect of creep on cladding stresses have been assessed. Results obtained from this study have provided input to the experimental GCFR cladding development work at Argonne National Laboratory.     

Similarity of high-temperature gas-cooled reactor technologies and designs in Russia and USA

Historical information concerning the development of high-temperature gas-cooled reactors in the USA and Russia is presented. The reactor facilities MHTGR (USA), VG-400 (Russia), VGM (Russia), GT-MGR (Russia, USA), and at the Fort St. Vrain nuclear power plant (USA) are described. The US programs for developing innovative high-temperature nuclear reactor technologies are examined. It is shown that the Russian and US technological developments for the fuel, reactor system, energy conversion system, and fission-product transport are similar.     

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.     

Fast flux irradiation tests performed at high temperature

The first gas-cooled fast breeder reactor (GCFR) fast flux irradiation experiment [F-1(X094)] consists of seven fuel rods clad in 20% cold-worked 316 stainless steel. The rods are individually encapsuled, with sodium filling the gaps within the capsule walls. The rods are fueled with (15% Pu, 85% U)O₂ and have depleted UO₂ lower and upper axial blankets and charcoal to trap volatile fission products. The cladding i.d. temperature range covered by these rods is 570–760°C (1055–1400°F).The in-reactor performance of the fuel rods in the F-1 high-temperature experiment, which achieved a burnup of 121 MWd/kg (13.0 at.%) on the lead rod, is described. All rods in the experiment have remained intact. The results of interim examinations [at 25 and 50 MWd/kg (2.7 and 5.4 at.%)] of fuel and fission product behavior and transport and comparisons of observed results with LIFE-III code predictions are described.The F-3 experiment, which consists of ten encapsulated GCFR fuel rods with surface-roughened (ribbed) cladding, shares a nineteen capsule subassembly with Argonne National Laboratory. Temperatures are controlled over the range 675°C (1250°F) to 750°C (1380°F). Irradiation is in the core region of the EBR-II and thus permits achievement of a higher fluence-to-burnup ratio than that obtained in the F-1 experiment.Preliminary results of a planned interim examination at an exposure of 46 MWd/kg (4.9 at.%) burnup and a fluence of 5.2 × 10²² n/cm² show that cladding failures occurred in nine of the ten rods. Preliminary indications are that the failures are due to defects in the sodium bond between the fuel rod and the capsule.The tests completed and currently under way have been scoping in nature, and irradiation in EBR-II of GCFR prototypical fuel (pressure equalized) rods with ribbed cladding is required to provide the information needed for reactor design on effects of exposure to high fluence and burnup and on design reliability for a statistically significant number of rods. The design and the operating conditions for the F-5 experiment being prepared for this purpose are described.     

Design aspects of the Chinese modular high-temperature gas-cooled reactor HTR-PM

The modular high-temperature gas-cooled reactor (MHTGR) has distinct advantages in terms of inherent safety, economics potential, high efficiency, potential usage for hydrogen production, etc. The Chinese design of the MHTGR, named as high-temperature gas-cooled reactor-pebble bed module (HTR-PM), based on the technology and experience of the HTR-10, is currently in the conceptual phase. The HTR-PM demonstration plant is planned to be finished by 2012. The main philosophy of the HTR-PM project can be pinned down as: (1) safety, (2) standardization, (3) economy, and (4) proven technology. The work in the categories of marketing, organization, project and technology is done in predefined order. The biggest challenge for the HTR-PM is to ensure its economical viability while maintaining its inherent safety. A design of a 450 MWth annular pebble bed core connected with steam turbine is aimed for and presented in this paper.     

HTGR plant dynamics analysis under accident conditions

Important features of high temperature gas-cooled reactor (HTGR) systems related to plant dynamics and accident analysis are discussed. Because of the basic simplicity of the HTGR system, it is possible to analyze the full reactor plant (core, helium circulators, steam generators and reheaters, feedwater controls, turbine controls, and plant protective action) in a single computer code. Representative dynamics analysis is presented for the Fort St. Vrain Power Station.