NPP Reactor Research Articles

Development of Software for Calculation of Erosion-Corrosion Wear of NPP Reactor Piping

Based on the analysis of laboratory test results and operation inspection reports on steam piping bend sections, computer programs have been created for predicting erosive-corrosive wear of materials of NPP facilities. The proposed software makes it possible to take into account the influence of various variables (piping design and layout, water chemistry characteristics, chemical composition of piping metal, operating parameters) on the erosion-corrosion wear rate of NPP facilities during long-term operation.

Assessment of In-Service Degradation of Mechanical Properties of NPP Reactor and Piping Materials

An experimental investigation has been carried out to clarify the influence of operating factors on static and cyclic strength of NPP reactor vessels and piping. The authors present the results of analysis of crack growth resistance characteristics of the base material and welded joints in WWER-1000 reactors (steel 15Kh2NMFA) in fatigue failure. The mechanisms of thermal aging of steels of NPP piping in long-term operation (up to 10,000 hours) have been studied.

Micromechanical Coupled Study of Crack Growth Initiation Criterion in Pressure Vessel Steel

We present results of the combined design-theoretical investigation of the mechanism of crack growth at the onset of ductile fracture of NPP reactor pressure vessels. Micromechanical approach to the prediction of ductile fracture has been applied, according to which the volume fraction of voids in the deformed material is determined by the finite-element method. On the basis of CT-specimen tests and known damage parameters, determined for smooth spherical specimens, we propose a micromechanical criterion of crack growth initiation for ductile fracture.

Radiation Embrittlement Understanding for PLIM Activities at EC-JRC-IE

Radiation embrittlement and aging mechanisms for NPP reactor pressure vessels and vessel internals have been studied within NPP Plant Life Management activities for the evaluation, prediction, and monitoring of the critical components' service life. The main achievements of the SAFELIFE project, integrating various networks on PLIM issues, are given. Results of neutron embrittlement of model alloys are presented, and surveillance and research data on WWER reactor pressure vessel and other steels are analyzed. Projects for the development of destructive and non-destructive testing of irradiated materials are outlined.

High Precision Measurements of Spectrum Averaged Total Neutron Cross Section for Chromium

The intense neutron beams formed by composite neutron filters at reactor horizontal channels give the fluxes of about 106 – 108 neutron/cm2s at the fixed neutron energies and it makes it possible to measure the neutron cross sections with accuracy better than 0.1%.The need of such measurements is formulated in the document The NEA High Priority Nuclear Data Request List, May 1998 (for Cr in the energy range 0.1 keV – 5 MeV and for Cr-52 in 10 eV – 20 MeV). Cr is one of the important structural materials for NPP and research reactors just as for fusion technology. These precise cross section data may be used also in the practice of work for evaluated nuclear data libraries. Now the discrepancies in total cross section for Cr in evaluated nuclear data libraries are in the range 3 – 50% for energies up to 200 keV.At Kyiv Research Reactor now we have the set of neutron filters for the fixed energies in the range 0 – 600 keV, here we present the transmission measurements at the energies 24.3 and 58.8 keV for Cr and Cr-52 (99.3% enrichment) samples.The results of measurements are presented together with the analysis of the known previous experimental data and the evaluated nuclear data from ENDF/B-VI, JEF-2.2, JENDL-3.2, BROND-2 and CENDL-2 libraries.

Effect of Irradiation on the Stability and Natural Vibrations of Containment Shells

An approximate approach is proposed to solve the problem on the natural vibrations and stability of an NPP reactor containment. Radiation alters the mechanical properties of such shells and causes volumetric expansion of their material. The problems are solved within the framework of the Timoshenko kinematic hypotheses by reducing shells inhomogeneous throughout the thickness to homogeneous shells with reduced tension–compression, shear, and flexural rigidities

Vibration monitoring of the Atucha-I NPP reactor vessel: experience and future improvements

Keeping nuclear plant components in prime operating condition is essential for overall plant safety and availability. One way to assess the good operating condition of nuclear components is an on-line monitoring system. A vibration monitoring system is intended to monitor a component or system to ensure that changes in the mechanical integrity are detected early, even during operation. This paper describes the experience gained during the development of a vibration monitoring system for the Atucha-I Nuclear Power Plant reactor vessel that started up in 1990. High sensitivity velocity sensors were mounted on the closure head of the reactor vessel to monitor the vibrations. Some of the results obtained for different reactor operating conditions are presented in the form of frequency spectra or RMS values for different frequency bands. A description of the system, its evolution and future improvements are also included.

Possibilities of Co-60 production at Leningrad NPP reactors

Possibilities of Co-60 production at Leningrad NPP reactors

CANDU® NPP reactor regulating system modelling using CATHENA

A CATHENA model is developed for the reactor regulating system and tested independently. Then, a complete CATHENA model is created by combining this model with the reference CATHENA model which has been developed to analyze a loss-of-coolant accident (LOCA) for Wolsong 2 generating station. This model is intended to set up plant control routines in CATHENA to provide a trip coverage analysis capability. The purpose of this work is to model the reactor regulating system using the CATHENA control model system. It includes the demand power routine, the light water zone control absorbers, mechanical control absorbers and adjusters. The CATHENA model was tested for steady state at 103% full power. A postulated accident transient (small LOCA) was also tested. The results show that the control routines in CATHENA were set up properly.

Primary Circuit Metal State With Regard to NPP Lifetime Extension

This article describes the assessment of the suitability for operation of soviet reactors. The VVER-440 (10 units), VVER-1000 (16 units) and RBMK-1000 (15 units) are the main types of soviet NPP reactors. Reactor design lifetime is 30 years. Primary circuit main equipment and piping have accordingly the same lifetime. The problem of lifetime extension arises in the case of design lifetime expiration or in the case of deviations from the reactor nominal characteristics. At present, only two reactors have expired their design lifetime and are decommissioned (1 unit of Beloyarsc NPP and 1 unit of Novovoronezh NPP). Deviations from the reactor nominal characteristics may appear for different reasons. The majority of them are as follows: Level of equipment and pipeline manufacture and assembly technology which at present does not provide for an absolutely defect-free metal state; level of flaw detection monitoring method development considering that defect detection is an accidental, probabilistic process; deviations from nominal operation conditions which, as a rule, are of an accidental nature; design deficiencies related as a rule to in essential equipment components: Deviation in their operation leads to an essential equipment component operational condition change; highly conservative level of quality assessment standards of operational flawmore » detection control stages are practically the same. Defects due to structural deficiencies occur rather rarely, but in the case of such defect occurrence at one unit, the same defects are to be expected at the other ones.« less

Response of a NPP reactor building under seismic action with regard to different soil properties

The object of this investigation is the response of a reactor building on seismic action with systematic variation of the soil stiffness. A thin-walled orthotropic containment shell on varying heavy and rigid foundations is regarded as calculation model. The soil stiffness is simulated by means of spring elements for horizontal translation and for rocking motions of the building. By the response spectra method the loads of the containment shell are calculated for a horizontal seismic excitation. The investigation is aimed at determining the influence of differentiated soil stiffnesses on the containment action effects and at recognizing the causes for the occurring effects. The results are thoroughly represented by selected quantities of the building's response, the effects from the soil-structure interaction are discussed and the causes of the effects clearly explained. A possibility is provided for determining critical soil stiffnesses which cause a significant intensification effect. The results of the investigations show that both the soil stiffness and structural configuration of the reactor building, particularly in case of the substructure being heavy and rigid, exert a decisive influence on the loading of the superstructure.