Mapping Conjugate Gradient Algorithms for Neutron Diffusion Applications onto SIMD, MIMD, and Mixed-Mode Machines

The performance of conjugate gradient (CG) algorithms for the solution of the system of linear equations that results from the finite-differencing of the neutron diffusion equation was analyzed on SIMD, MIMD, and mixed-mode parallel machines. A block preconditioner based on the incomplete Cholesky factorization was used to accelerate the conjugate gradient search. The issues involved in mapping both the unpreconditioned and preconditioned conjugate gradient algorithms onto the mixed-mode PASM prototype, the SIMD MasPar MP-1, and the MIMD Intel Paragon XP/S are discussed. On PASM , the mixed-mode implementation outperformed either SIMD or MIMD alone. Theoretical performance predictions were analyzed and compared with the experimental results on the MasPar MP-1 and the Paragon XP/S. Other issues addressed include the impact on execution time of the number of processors used, the effect of the interprocessor communication network on performance, and the relationship of the number of processors to the quality of the preconditioning. Applications studies such as this are necessary in the development of software tools for mapping algorithms onto either a single parallel machine or a heterogeneous suite of parallel machines.     

Fracture behaviour of radiolytically oxidised reactor core graphites: a view

Abstract

This paper provides a view on the fracture behaviour of polygranular graphites, used to moderate gas cooled nuclear reactors. Graphite is often cited as a classic example of a brittle material because failure, in tension, is associated with small strains. However, attempts to characterise the fracture behaviour of graphite by linear elastic fracture mechanics methods have been largely unsuccessful. Observations of graphite fracture show that elastic strain energy may be dissipated by the formation of distributed microcracks, and their formation may be responsible for non-linearity in the rising load–displacement curve. Progressive softening behaviour may also be observed in some specimens after the peak load. This type of load–displacement behaviour is a characteristic of quasi-brittle materials. Radiolytic oxidation increases the proportion of porosity within reactor core graphite so that the microstructure becomes increasingly skeletal. Consideration is given to the fracture of radiolytically oxidised graphite to support an argument for quasi-brittle behaviour.     

Nanocrystalline diamond films for nanotechnology applications

Abstract

Nanocrystalline diamond films with rms roughnesses in the order of 50 nm have been prepared by microwave plasma enhanced chemical vapour deposition. Ar-CH₄ gas mixtures have been used in a coaxially bladed reactor, and Ar-H₂-CH4 gas mixtures have been used in a reactor with a dual mode microwave applicator. These films have been examined by ultraviolet and visible Raman spectroscopy. They have been subjected to nanotribological investigation, with a view to their incorporation into microelectromechanical systems (MEMS). Comparisons with other diamond and hard carbon films are made, and the advantages of nanocrystalline diamond films are discussed.     

Evaluation of hot strip mill backup roll cladding

Abstract

A new cladding material based on the Fe–Cr–Mo–V–C alloy system, suitable for submerged arc welding, has been designed for the refurbishment of forged and cast backup rolls used in the finishing stands of hot strip rolling mills. The work undertaken includes mechanical analysis, mechanical testing, and microstructural characterisation. The mechanical analysis indicated the nature and level of stresses operating near the surface of rolls; mechanical testing allowed material performance to be anticipated. An optimal post-weld heat treatment procedure, which maximises strength while minimising material strain hardening, was subsequently chosen. The microstructure of the candidate cladding material is a mixture of lower bainite and martensite, containing a very fine distribution of molybdenum carbides. In situresults have shown that welded rolls outperform traditional rolls, as the amount of steel rolled per millimetre of cladding material is 40% higher than with forged rolls and double that obtained with cast rolls.     

Fracture behaviour of radiolytically oxidised reactor core graphites: a view

Abstract

This paper provides a view on the fracture behaviour of polygranular graphites, used to moderate gas cooled nuclear reactors. Graphite is often cited as a classic example of a brittle material because failure, in tension, is associated with small strains. However, attempts to characterise the fracture behaviour of graphite by linear elastic fracture mechanics methods have been largely unsuccessful. Observations of graphite fracture show that elastic strain energy may be dissipated by the formation of distributed microcracks, and their formation may be responsible for non-linearity in the rising load–displacement curve. Progressive softening behaviour may also be observed in some specimens after the peak load. This type of load–displacement behaviour is a characteristic of quasi-brittle materials. Radiolytic oxidation increases the proportion of porosity within reactor core graphite so that the microstructure becomes increasingly skeletal. Consideration is given to the fracture of radiolytically oxidised graphite to support an argument for quasi-brittle behaviour.     

Repatriation of Vinča RA reactor spent fuel

Abstract

The RA research reactor is located at the Vin?a Institute of Nuclear Sciences near Belgrade, Serbia. The reactor is a 6·5 MW, tank-type, heavy water moderated and cooled reactor of Russian design which commenced operation in 1959. After being temporarily shut down in 1984 for refurbishment, a final shutdown decision was made in 2002. Operations are underway to safely remove and repatriate the spent nuclear fuel (SNF) to the Russian Federation (RF), as well as to improve waste management throughout the Vin?a site and prepare a plan for reactor decommissioning. As a major activity within the Vin?a Institute Nuclear Decommissioning (VIND) Programme, the repatriation of over 8000 SNF elements containing 2·5 tons of uranium metal will significantly reduce nuclear proliferation and environmental safety risks confronting the current facility. Poor water quality in the SNF storage basins and degraded fuel integrity significantly challenge efforts to repackage and transport the SNF. This paper will focus on the activities related to SNF repackaging and shipment, report on progress, detail significant challenges and provide an overview of the fully integrated VIND project.     

Experience of surveillance schemes adopted for magnox steel reactor pressure vessels

Abstract

Surveillance or monitoring schemes are recognised to be an important part of any strategy to demonstrate that reactor pressure vessels used in civil nuclear power stations are operated within a safe regime. In the paper the authors describe the experience obtained from the surveillance schemes adopted for the UK's magnox nuclear power stations that were constructed with C–Mn steel reactor pressure vessels. These power stations were constructed in the late 1950s and 1960s and the last ceased generating in 2006. During the lifetime of the fleet with steel pressure vessels, there were developments in testing, observed changes in properties and understanding of radiation damage process that challenged the safety cases to support the operation of the stations. At the time the reactors were designed the concept of fracture toughness was only beginning to be investigated yet, during the lifetime of the stations, fracture toughness testing was successfully adopted as an input to fracture mechanics based assessment of the steel vessels. Over the operating life, a series of challenges emerged that were successfully addressed, including both hardening and non-hardening embrittlement, the latter due to impurity phosphorous segregation in weld metal and contributions from thermal nuclear embrittlement. These challenges led to the adoption of sophisticated statistical techniques to assess changes in embrittlement properties of the most critical construction material – submerged arc weld metal. A large scale sampling and testing programme of submerged arc weld metal removed from a decommissioned reactor pressure vessel validated the assessment process. As a result of successfully addressing these, and other challenges when the last two steel pressure vessel stations closed in December 2006, they had achieved lifetimes of nearly 40 years.     

Rapporteur summary of the PATRAM 2010 Symposium

Abstract

A new feature at the PATRAM 2010 symposium was the introduction of 'rapporteurs' to provide summaries of the oral presentations and panel discussions. At the beginning of each morning plenary session, from Tuesday through Friday, a different rapporteur presented brief highlights from the sessions and panels taking place on the previous day. These reports are summarised below.     

Characterisation of as cast model Mn–Si–Cr–Ni steels for reactor pressure vessel applications

Abstract

Initial results are reported from a study aimed to investigate the role and influence of the elements Cr, Ni, Mn and Si on the radiation stability of reactor pressure vessel steels. Twelve as cast model ferritic steels with basic composition typical of those used in Russian WWER-1000 and Western PWR reactor pressure vessel materials were subjected to Charpy impact, magnetic Barkhausen noise (MBN), Vickers hardness tests and SEM examination. Higher Cr content in model steels was found generally to give increased RMS values independent of Mn and Si contents. The ductile–brittle transition temperatures (DBTT) and hardness values of the model steels were found to be independent of composition. Two steels, with low concentration of Ni and high concentration of Cr or vice versa , showed high transition temperatures (?16 and ?42°C respectively). An additional heat treatment to improve the properties is being considered for these compositions. The correlation between DBTT and MBN results has potential for rapid determination of the effect of composition and irradiation on the steel properties. The next stage of the assessment will investigate the effect of irradiation of the model steels to accumulated neutron fluences of ～10¹⁹ cm^?2.     

Energy drivers for materials research and development

Abstract

An overview is given of the significant materials aging issues for nuclear power reactors. Despite the fact that many of these issues were not anticipated at the design stage, the development of active management techniques, backed by improved understanding and modelling of the underlying processes, has permitted many nuclear plants to run safely and efficiently well beyond their design lifetime. The mechanisms of aging are reviewed and management strategies for aging plant assessed. In conclusion, the requirements and options for future nuclear plant are considered.