Fuel for pebble-bed HTRs

More than half a million spherical fuel elements with high-enriched uranium have been manufactured in Germany. Containing high-enriched uranium and thorium oxide fuels, many elements have been successfully tested in MTRs and in AVR under high temperatures.In 1979 the reference fuel cycle was changed to low-enriched uranium fuel. Design specifications both for coated particles and the fuel element were reviewed in the light of previous experience with high-enriched fuel. The LEU qualification programme is well under progress to be completed by 1985/86.     

Design and analysis of multicavity prestressed concrete reactor vessels

A critical state-of-the-art review of multicavity prestressed concrete reactor vessel (PCRV) design and analysis practice is presented. Included are discussions of basic design concepts, the behavior of liners and penetrations, and the various tests required and/or employed to demonstrate acceptance of new vessel geometries and innovations. Brief reviews are given of the influences of design codes such as ACI/ASME Section III, Division 2, and BS4975; analysis methods including elastic, inelastic, and time-dependent techniques; the constituve equations that are essential to the satisfactory use of these techniques; and semi-empirical methods for calculating ultimate strengths of multicavity vessels.Tests conducted on liner plates, liner anchorage systems, and cooling tubes are reviewed together with the methods of analysis used in the design of anchorage systems. The adequacy and economy of present liner systems are considered and possible modifications in design are suggested.Design code requirements and methods of analysis for penetrations are discussed. The various types of closure designs that have been proposed and in some cases employed are evaluated on the basis of overall PCRV design philosophy. Several methods of prestressing PCRVs are considered with respect to relative advantages and disadvantages; existing overall vessel in-service inspection requirements are evaluated.     

Embedment instrumentation for prestressed concrete pressure vessels

An investigation is being conducted to evaluate the performance of various types of concrete and embedment instrumentation in order to determine present capability for monitoring of prestressed concrete reactor vessels for their projected 20–30 yr operating life-times. The initial phase of the investigation consisted of a technology assessment and an experimental evaluation of commercial concrete embedment strain gages to determine basic gage characteristics and performance under both simulated PCRV and extreme environments. It was concluded that (1) gage selection should be based on specific applications, (2) gage calibration factors should be determined for each application, (3) improved materials and sealing techniques are needed, and (4) other promising measurement techniques should be evaluated.     

Development of prestressed concrete pressure vessels for HTR

In 1970 an HTR development programme was started. Whilst the PCPV required to house the HTR will be seen as a natural development from those already designed and under construction for AGR, it was considered important to obtain experimental verification of a number of certain aspects related essentially to the HTR vessel. In particular these included the increased wall thickness in relation to cap thickness and internal height, diameter of boiler penetration and the small ligament in the standpipe zone. Furthermore, about 50% of the cross-sectional area of the pressurised zone is occupied by the standpipe penetrations.This paper describes part of the experimental works undertaken to develop the design of the PCPV. The particular investigations described are:

1. (1) Ultimate load tests on 1/40th scale concrete vessels pressurised in the short term to failure. The behaviour of the models is compared with that predicted by an ultimate load method of analysis. The method which was applied to the Hartlepool and Heysham vessels is based on finding the collapse mechanism associated with the minimum potential energy in the vessel.

2. (2) Short and long term tests on 1/26th and 1/8th scale models of the top cap of the vessel to examine the behaviour under the action of prestress loads, internal pressure and sustained temperature. An assessment of the ultimate strength of the cap, based on simplified design methods is presented. Furthermore the components of the total shear force are identified, e.g. shear resistance of the compression zone and shear transfer of the aggregate interlock within the tension zone. Estimation of the magnitude of these components is also given.

This work has shown that the design concept developed for the AGR vessels can be satisfactorily developed to produce a containment vessel suitable for HTR.     

The behaviour of concrete in prestressed concrete pressure vessels

In the design of prestressed concrete pressure vessels, long term concrete property data are required by the designer such that realistic estimates can be made of the vessels' 30-year stresses and deformations under the various operating conditions to which it will be subject. To achieve this aim, the shrinkage, short and long term deformation under load and thermal expansion behaviour of the vessel concrete has to be determined under conditions simulating those likely in the structure. In this paper, therefore, concrete properties are examined in relation to vessel design. Results obtained from the test programmes carried out for the Wylfa and Hartlepool nuclear power stations are presented in relation to our understanding of each property obtained from a detailed literature analysis.

The effect of temperature on three concrete properties of major importance in vessel design, e.g. compressive strength, thermal expansion and long term deformation under load (creep), is discussed at operational temperature up to 70°C. Consideration is also given, in the light of experimental data, on the effect of higher temperatures on these properties.     

Industrial aspects of revisited HTRs

The ‘new’ High Temperature Reactors (HTRs) are based on the direct gas turbine cycle which brings high inherent safety together with high efficiency. This combination makes them attractive cost-wise for countries with small and medium size grids that cannot afford larger units. For others, their fuel flexibility gives them extra attraction as burners of Plutonium (from weapons or LWR reprocessing) or to decrease minor actinide build-up by use of thorium. Extrapolation of existing technologies, mostly materials, are needed to bring the concept to life. The EU framework programme and collaboration with the Japanese and Chinese developments should help realise this wish. Two industrial groupings have launched the PBMR and GT-MHR concepts. Promotion of such forward looking concepts will also help attract talented young people to nuclear industry.     

Analysis of reinforced concrete containment vessels considering concrete cracking

Cracking of concrete influences the stress analysis of concrete containment vessels. If cracking is ignored, the resulting shell analysis can be unconservative in some cases and extremely conservative in others. A cracked concrete shell is a structurally orthotropic one. That is, it does not have the same properties in membrane action and bending action. Closed form equations are presented for cracked concrete shells using the split rigidity concept. The equations cover symmetrically loaded cylindrical shells, effects of concentrated forces and moments on spherical shells, and effects of openings and concentrated forces and moments on cylindrical shells. In addition, methods are discussed that can be applied to cracked concrete shells by using finite element techniques.     

Safe analysis for cooling pipes for prestressed concrete reactor vessels

A step-by-step analysis is given for the direct computation of two-dimensional heat flow to and safe pitching of the cooling pipes. Two models of the cooling system have been selected and calculations have been carried out for an existing vessel. On one model this analysis is compared with the three-dimensional finite element analysis for obtaining insulation conductances for various cooling pipe pitches.     

Study on the influence of steel-fibre-reinforced concrete on the behaviour of anchorage for steel liners

On the interior wall of concrete vessels with high requirements to their leak-tightness, usually steel liners are provided and attached by anchors. The carrying capacity of the anchors is defined by load-displacement curves with regard to concrete quality, anchor type and loading. Using normal concrete and steel-fibre-reinforced concrete in an extensive test programme the main parameters, such as temperature from 20 to 250 °C, short- and long-time loading, without and with prestressing, have been investigated. Two different types of curve result. The influence of this on the carrying behaviour of different liner anchor systems with headed studs and cooling pipes has been worked out by calculations. Using steel-fibre-reinforced concrete the anchor displacements can be reduced by about 30%.     

Advantages of going modular in HTRs

A multitude of problems that are encountered in large HTR power plans, constructively as well as concerning plant safety, can be related to the mere physical size of a large reactor core.In limiting the thermal power of an HTR-module to approximately 200 MW an inherent limitation of the fuel element temperature below critical values can be guaranteed for all possible core heat up accidents. Consequently, a significant failure rate of coated particles can be excluded and, hence, out of physical reasons, no intolerable fission product release from the core will ever have to be considered.The HTR-module is so qualified and very well suited for all possible plant sides which have to be taken into consideration for medium sized plants for the production of process steam and electricity.The cost investigations show considerable cost advantages for modular HTRs. For German conditions it was found that even a four-modular plant (800 MW/thermal) is competitive with a fossile-fueled plant of the same size, the specific plant costs were evaluated to be DM 4700/kW (electric). Moreover the investigations show that the increase of the power of the modular unit yields only small cost advantages, therefore in a modularized power plant it even would be possible to reduce the power of a modular unit below 200 MW without having to cope with severe economic penalties, if the distance from technological or safety limits is felt to be too small.     

Movements in prestressed concrete reactor vessels

Prestressed concrete reactor vessels (PCRVs) are constructed by placing concrete using lift and bay arrangement. The vessels are built up in a number of individual bays. The size and shape of these bays and the order in which they are cast has a significant effect on the size and sense of construction movements. Lack of information about factors such as age of any bay before prestress, the time of year of casting and the time interval between adjacent vertical and horizontal bays would lead to costly design improvements and could lead to unforeseen technical problems both under operational and ultimate conditions. An attempt has been made to establish a philosophy and a rational method of assessing fairly accurately these movements. Various movements and their causes and effects are discussed in detail. A three-dimensional finite element analysis is developed to predict these movements and to assess the behaviour of the vessel parameters under operational conditions. A computer program, ISOPAR, is developed and is tested against experimental and measured results.     

Design and analysis of concrete reactor vessels: New developments, problems and trends

This lecture reviews new developments in analysis and design of prestressed concrete reactor vessels (PCRV). After a brief assessment of the current status and experience, the advantages, disadvantages, and especially the safety features of PCRV, are discussed. Attention is then focused on the design of penetrations and openings, and on the design for high-temperature resistance — areas in which further developments are needed. Various possible designs for high-temperature exposure of concrete in a hypothetical accident are analyzed. Considered are not only PCRVs for gas-cooled reactors (GCR), but also guard vessels for liquid metal fast breeder reactors (LMFBR), for which designs mitigating the adverse effects of molten sodium, molten steel, and core melt are surveyed. Realistic analysis of these problems requires further development in the knowledge of material behavior and its mathematical modeling. Recent advances in the modeling of high-temperature response of concrete, including pore water transfer, pore pressure, creep and shrinkage are outlined. This is followed by a discussion of new developments in the analysis of cracking of concrete, where the need of switching from stress criteria to energy criteria for fracture is emphasized. The lecture concludes with a brief discussion of long-time behavior, the effect of aging, and probabilistic analysis of creep.     

Development and testing of analytical models for the pebble bed type HTRs

The pebble bed type gas cooled high temperature reactor (HTR) appears to be a good candidate for the next generation nuclear reactor technology. These reactors have unique characteristics in terms of the randomness in geometry, and require special techniques to analyze their systems. This study includes activities concerning the testing of computational tools and the qualification of models. Indeed, it is essential that the validated analytical tools be available to the research community. From this viewpoint codes like MCNP, ORIGEN and RELAP5, which have been used in nuclear industry for many years, are selected to identify and develop new capabilities needed to support HTR analysis. The geometrical model of the full reactor is obtained by using lattice and universe facilities provided by MCNP. The coupled MCNP-ORIGEN code is used to estimate the burnup and the refuelling scheme. Results obtained from Monte Carlo analysis are interfaced with RELAP5 to analyze the thermal hydraulics and safety characteristics of the reactor. New models and methodologies are developed for several past and present experimental and prototypical facilities that were based on HTR pebble bed concepts. The calculated results are compared with available experimental data and theoretical evaluations showing very good agreement. The ultimate goal of the validation of the computer codes for pebble bed HTR applications is to acquire and reinforce the capability of these general purpose computer codes for performing HTR core design and optimization studies.     

Analysis and application of prestressed concrete reactor vessels for LMFBR containment

An analytical model of a prestressed concrete reactor vessel (PCRV) for LMFBR and the associated finite element computer code, involving an explicit time integration procedure, is described. The model is axisymmetric and includes simulations of the tensile cracking of concrete, the reinforcement, and a prestressing capability. The tensile cracking of concrete and the steel reinforcement are both modeled as continuously distributed within the finite element. The stresses in the reinforcement and concrete are computed separately and combined to give an overall stress state of the composite material. The reinformcement is assumed to be elastic, perfectly-plastic; the concrete is taken to be elastic, with tensile and compressive stress limits. Cracking of concrete is based on the criterion of maximum principal stress; a crack is assumed to form normal to the direction of the maximum principal stress. Attention is also given to the fact that cracks do not form instantaneously, but develop gradually. Thus, after crack initiation the normal stress is reduced to zero gradually as a function of time. Residual shear resistance of cracks due to aggregate interlock is also taken into account. An existing crack is permitted to close. Prestressing of the PCRV is modeled by special structural members which represent an averaged prestressing layer equivalent to an axisymmetric shell. The internal prestressing members are superimposed over the reinforced concrete body of the PCRV; they are permitted to stretch and slide in a predetermined path, simulating the actual tendons.The validity of the code is examined by comparison with experimental data. Both static and dynamic data are compared with code predictions, and the agreement is satisfactory. A preliminary design has been developed for both pool and loop-type PCRVs. The code was applied to the analysis of these designs. This analysis reveals that the critical locations in such a design would be the head cover and the junction between the cover and the vessel wall and indicates the pattern of crack development. The results show that the development of a design adequate for current HCDA loads is quite feasible for pool-type or loop-type PCRVs.     

Contributions to the safety-analysis of pebble-bed HTRs

Results of calculations are presented, which are concerned with the short-time behaviour of pebble-bed HTRs with the so-called OTTO-(“Once Through Then Out) loading scheme. Two general cases are considered: a massflow reduction and an ingress of water. The results indicate that OTTO-HTRs display a high degree of inherent safety. As a special conclusion power densities of about 5 MW/m³ are recommended.     

Safety and reliability of prestressed concrete reactor vessels

A step-by-step analysis is given for the direct computation of two-dimensional heat flow to and safe pitching of the cooling pipes. Two models of the cooling system have been selected and calculations have been carried out for an existing vessel. On one model this analysis is compared with the three-dimensional finite element analysis for obtaining insulation conductances for various cooling pipe pitches.     

Structural criteria for extreme dynamic internal pressure loadings of vessels and closure heads

Structural criteria for analytically evaluating the pressure retaining capability of vessels and closure heads subjected to extreme internal transient dynamic loadings are presented. The criteria project against tensile plastic instability and local ductile rupture failure modes. To minimize the number of critical areas that may need more rigorous analytical methods, a screening criterion for limiting the membrane, bending and local stresses is defined. The stresses for this criterion are calculated from either simple and economical elastic dynamic or equivalent static methods. For the critical areas that remain, a strain-based criterion for strains derived from dynamic, inelastic methods is given. To assure that the criteria are properly applied, guidelines are outlined for controlling methods for deriving stresses and strains, for selecting appropriate material properties and for addressing specific dominating parameters that affect the validity of the analysis. The application of the criteria to a complex liquid metal fast breeder reactor vessel and closure head and the subsequent experimental verification of the results by several scale model experiments are summarized.     

Cavity liner,penetration liners and closures of a prestressed concrete pressure vessel for a gas-cooled power reactor

The design criteria of the PCRV cavity liner and the penetration liners and closures are discussed, including the requirements for anchoring the liners to the concrete and the closure design requirements. Materials of construction are identified including discussion of special impact strength requirements and neutron radiation effects. Construction consideration, including inspections, tests, and quality assurance employed during construction are identified. The application of these requirements to the Fort St. Vrain reactor is discussed.