The prestressed cast iron reactor pressure vessel (PCIPV)

The concept of the prestressed cast iron reactor pressure vessel (PCIPV) emerges from the utilization of cast iron in the design of radiation and thermal shields. The principles of construction are explained using a model which is at present being assembled. Salient differences between the proposed vessel concept and a prestressed concrete reactor pressure vessel (PCPV) are discussed.     

Research and development towards a prestressed concrete reactor pressure vessel for light-water reactors

The escalation of power unit for nuclear power stations with light-water reactors has led to difficulties in manufacturing and to safety problems for thick-walled steel pressure vessels. The unit output was increased in the last ten years step by step from 300 to 600 and 1200 MW(e). Escalation to more than 2000 MW(e) can be expected. In light of this escalation of unit generation capacity the PCPV has an economic chance of being introduced for light water reactors. In 1968, Krupp began development of a PCPV for boiling-water reactors (BWRs). In 1972, a reference design was completed. In 1973, work on a PCPV for light-water reactors was concentrated on a prototype design with particular reference to safety and economic aspects. This paper presents the reference design finished in 1972. The main points covered are: the overall design; the structural design of particular parts such as the liner and insulation, and the penetration tubes; structural analysis of the PCPV and liner; safety aspects; tests made with the liner, insulations, and concrete; and development potential of the PCPV for light-water reactors.     

The optimum temperature for regenerative water-heating systems at atomic power stations with water-cooled and water-moderated power reactors (WWPR)

The base point method is used to determine the optimum temperature t _r ^opt for a regenerative water-heating system which corresponds to a minimum calculated cost c_e (kopeks/kwh) for electrical energy. It is shown that for an atomic power station having a water-cooled and water-moderated power reactor (WWPR) the value of t_r ^opt is near the maximum defined by the steam conditions after the first turbine stage.     

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.     

Vessels for sodium-cooled fast nuclear reactors: General considerations and critical examination of some vessel structures

In the first part of this paper some general characteristics of vessels for sodium-cooled fast nuclear reactors are discussed, emphasizing their differences with the vessels of thermal nuclear reactors.     

Reconstitution techniques qualification and evaluation to study ageing phenomena of nuclear pressure vessel materials (RESQUE)

The RESQUE project aims at optimising and normalising reconstitution techniques and is now in its final phase. The project belongs to the AGE-cluster, that also involves the REFEREE project being used as an input to RESQUE. At FISA '97 the reference data on non-reconstituted specimens were presented together with a set of recommendations on temperature measurements (WP1, WP2). Now, the results on the quality and limiting conditions of the reconstitution weld seam are discussed. The combination of this information leads to a set of recommendations for optimised reconstitution parameters that allow to qualify reconstitution equipment and methodology (WP3). The recommendations on the minimum insert length for impact and three-point bend fracture toughness testing have been established (WP4). Recommendations on dimensional tolerance deviations were put forward (WP5) and series of tests have been performed on selected reconstituted irradiated specimens (WP6). All work packages have been summarised. The overall information is being recapitulated in a ‘Proposal for Code of Practice for Reconstitution of Irradiated C_V-type Specimens’ (WP7).     

Thermal-hydraulic characteristics of pressurized water reactors during commercial operation: IV. Saturation pressure, circulation ratio and carry-under in a PWR steam generator (Tricastin 1 nuclear power plant)

Saturation pressure, downcomer velocity distribution (angular), local circulation ratio (angular distribution), overall circulation ratio, and carryunder are measured as a function of load in the steam generator No. 20 of the PWR Tricastin 1. Carryunder is found to be negligible; It reaches a maximum value of about 0.8% at full load. The circulation ratio decreases with load; it decreases from 22 at 11% of full load to 4.4 at full load. At full load the measured circulation ratio, 4.4, compares with a design value of 4.1. Variation of the circulation ratio and saturation pressure with load reported here for Tricastin 1 are in good agreement with results reported earlier for the PWR Bugey 4.     

Hybrid soliton nuclear reactors: A model and simulation (encapsulated long living accelerator driven system)

We intend to explore the potential of Hybrid Soliton Reactors (Réacteur Hybride à Soliton, RHYS) for producing energy. In our case an encapsulated long living fission reactor is driven by a proton accelerator, who produces neutrons on a target. In a first part we give the mathematical approach of such a sub-critical reactor, as an extension of the “Soliton Reactor” which was recently proposed by different authors, Edward Teller, L.P. Feoktistov, and others (H. Sekimoto under the name “Candle reactor”). In a second part we give results of simulations and explore the possibilities to control such a system.     

Severity, causes and frequencies of pressure vessel thermal shock at U.S. pressurized water reactors (1963–1981)

This article summarizes the results of a study whose purpose was to identify and analyze events at U.S. pressurized water reactors (PWRs) which either resulted in or could be considered as precursors to pressure vessel thermal shock (PVTS). The data base includes 16,000 Licensee Event Reports on 49 PWRs, covering 329 reactor-years from 1963 through 1981. The 99 identified events were assigned to one of five categories of severity. On this basis, 34 of the events were considered significant to PVTS. The events are categorized by cause and by vendor of the plant at which the event occurred. Some probabilistic risk assessment data are also developed.     

A probabilistic seismic risk assessment procedure for nuclear power plants: (I) Methodology

A new procedure for probabilistic seismic risk assessment of nuclear power plants (NPPs) is proposed. This procedure modifies the current procedures using tools developed recently for performance-based earthquake engineering of buildings. The proposed procedure uses (a) response-based fragility curves to represent the capacity of structural and nonstructural components of NPPs, (b) nonlinear response-history analysis to characterize the demands on those components, and (c) Monte Carlo simulations to determine the damage state of the components. The use of response-rather than ground-motion-based fragility curves enables the curves to be independent of seismic hazard and closely related to component capacity. The use of Monte Carlo procedure enables the correlation in the responses of components to be directly included in the risk assessment. An example of the methodology is presented in a companion paper to demonstrate its use and provide the technical basis for aspects of the methodology.