Verification of NETFLOW code using plant data of sodium cooled reactor and facility

The plant simulation code NETFLOW on PC applicable to the liquid-metal cooled reactors has been developed on the basis of the models developed for single-phase and two-phase light water flow systems. The functions of this code have been verified by individual tests for light water flow systems and a sodium flow system. In order to apply this code to a sodium cooled fast reactor, several extra functions were verified using the plant data obtained using 50 MW steam generators and the Monju fast breeder reactor. Finally, the turbine trip transient of the Monju was simulated and the result was compared with the measured plant data. Good agreements were obtained in these verifications. As a result of the present study, the code can be applied as an education tool for students.     

Development of natural circulation analysis methods for a sodium cooled fast reactor

A natural circulation evaluation methodology has been developed to insure safety of a sodium cooled fast reactor (SFR) of 1500 MWe adopting a natural circulation decay heat removal system (NC-DHRS). The methodology consists of a one-dimensional safety analysis which can be applied to safety evaluation for SFR licensing taking into account the temperature flattening effect due to buoyancy force in the core, and a three-dimensional fluid flow analysis which can evaluate thermal-hydraulics for local convection and thermal stratification in the primary system and DHRSs. The one-dimensional safety analysis method and the three-dimensional fluid flow analysis method have been validated using the test results of a water test apparatus and a sodium test loop for some typical transient events selected from the design basis events of the SFR. Finally, it has been confirmed that a good agreement between the test results and analysis results has been obtained, and reliability of each method has been demonstrated.     

Control system for the reactor at the krsko atomic power plant

Translated from Atomnaya Énergiya, Vol. 68, No. 5, pp. 343–346, May, 1990.     

Theoretical and experimental thermal-hydraulic analysis of a new sodium cooled fast breeder reactor core

This paper presents the experimental and theoretical results of the thermal-hydraulic design of a new fast breeder reactor core concept. The main feature of this concept is the omission of fuel element cans.The hydraulic function of these fuel element cans is substituted by a winding flow path through the radial blanket and a ring chamber without tubes.A computer code based on the quasi-continuum-theory and especially adapted to the features of the new core concept is developed for theoretical investigations. The pressure drop of the rod bundles is specified by a resistance tensor.The experimental investigations are realized in a test facility, where sodium is simulated by water. Pressures and velocities are measured.Theoretical and experimental results show good agreement. The aim of flattening of the coolant outlet temperature distribution can be reached with satisfying accuracy.     

Optimization of material and production to develop fluoroelastomer inflatable seals for sodium cooled fast breeder reactor

The feasibility of producing thin-walled fluoroelastomer profiles under continuous, atmospheric-pressure vulcanization conditions in air has been demonstrated by successful manufacture of ∼2 m diameter test inflatable seals for the 500 MWe, Prototype Fast Breeder Reactor (PFBR) using a 50/50 blend formulation of Viton^® GBL-200S/600S based on advanced polymer architecture (APA). A commercial cold feed screw extruder with 90 mm diameter screw was used along with continuous cure by microwave (2.45 GHz) and hot air heating (190 °C) at a line speed of 1 m/min to produce the seals. The blend formulation promises significant improvement in the performance and safety of the seals. This article depicts the relevant characteristics of the original inflatable seal compound that was used as reference to achieve the objectives through synchronized optimization of material and production technologies. The production trials are outlined and the blend formulation used with minor factory modifications to produce the test seals is reported. Progressive refinements of the original, Viton^® A-401C based compound to the blend formulation is presented along with an assessment of potential performance gains. Possible uses of the reported formulation and production technique for other large diameter, high temperature seals of PFBR are indicated along with the envisaged activities en-route the production of perfected reactor inflatable seals.     

Applications possibilities for fuzzy failure analysis, and diagnosis of reactor plant components and areas

Presented is an analysis on the relevance of fuzzy sets and systems theory to the reactor plant failure analysis and diagnosis. A fuzzy failure diagnostic Command Synthesis is elaborated upon and two inequalities for estimating the bounds on the components of the failure vector are presented. These are then combined to obtain a single inequality for failure diagnosis. Numerical examples are then given for realistic reactor plant situations and specific results are obtained for the failure diagnosis of a main coolant pump.     

Experimental stress analysis at reactor and plant components using hard- and software

Increasing the reliability of pressure vessels the methods of experimental stress analysis are world-wide used for additional data. Components and software programs extend and improve the possibilities of the experimental stress analysis continuously. On-line display of stress results during applications is now possible. Field portable modular equipment with numerous instrument configurations meet the plant and field environmental constraints. The following paper describes a transportable quasistatic multipoint measuring unit with data processing and its use during the first pressure test of a nuclear reactor containment and further applications.     

Measurement of the parameters of the VVÉR reactor at the Kalinin nuclear power plant

Scientific and Technical Center YaRB, Gosatomnadzora of Russia. Kalinin Nuclear Power Plant. Translated from Atomnaya énergiya, Vol. 77, No. 3, pp. 229–230, September, 1994.     

Development of an evaluation methodology for the natural circulation decay heat removal system in a sodium cooled fast reactor

A natural circulation evaluation methodology has been developed to ensure the safety of a sodium-cooled fast reactor (SFR) of 1500 MW adopting the natural circulation decay heat removal system (NC-DHRS). The methodology consists of a one-dimensional safety analysis which can evaluate the core hot spot temperature taking into account the temperature flattening effect in the core, a three-dimensional fluid flow analysis which can evaluate the thermal-hydraulics for local convections and thermal stratifications in the primary system and DHRS, and a statistical safety evaluation method for the hot spot temperature in the core. The safety analysis method and the three-dimensional analysis method have been validated using results of a 1/10 scaled water test simulating the primary system of the SFR and a sodium test simulating a part of the primary system and the DHRS with about a 1/7 scale, and the applicability of the safety analysis for the SFR has been confirmed by comparing with the three-dimensional analysis adopting the turbulence model. Finally, a statistical safety evaluation has been performed for the SFR using the safety analysis method.     

Investigation of the coolant temperature at the core entrance for the reactor in the Tianwan nuclear power plant (China)

Tests to determine the coolant temperature at the core entrance have been performed for the first time on power generating units with VVéR-1000 reactors in the No. 1 unit of the Tianwan nuclear power plant at the physical startup stage. Such tests are needed to increase the operational reliability of the shielding on the basis of the margin up to a crisis of heat transfer. If the results of the tests performed on the No. 1 unit are confirmed by planned tests on the No. 2 units of the nuclear power plant, then the possibility of performing such tests only on the first units will be validated, which will decrease the time required to put serially produced power generating units into operation. Experiments have validated the possibility of eliminating tests on mixing of the circuit flows in subsequent units with VVéR reactors to decrease the time and cost of putting a unit into operation provided that tests are performed to determine the coolant temperature. __________ Translated from Atomnaya énergiya, Vol. 103, No. 2, pp. 93–98, August, 2007.     

Validation of nuclear and radiation safety of a container for spent amb reactor fuel assemblies at the Beloyarskaya nuclear power plant

An approach is proposed for validating the nuclear and radiation safety of a container for spent fuel assemblies from AMB-100 and-200 reactors at the Beloyarskaya nuclear power plant. To validate the radiation safety, the characteristics of fuel assemblies and their classification according to the average fuel burnup in the casing, and the intensities of n and γ radiation in the casing are analyzed. Nuclear safety is validated on the basis of the concept of a “model” casing. This model makes it possible to obtain an upper estimate of the effective coefficient of neutron multiplication for all real casings with fuel assemblies. Calculations are used to determine the minimum necessary thickness of the vessel, bottom, and cover for 17-and 35-place casings. It is shown that no special neutron protection is needed. The container design to be developed meets the IAEA and OPBZ-83 safety standards. __________ Translated from Atomnaya énergiya, Vol. 100, No. 6, pp. 423–428, June, 2006.     

Process resulting in the destruction of the No. 4 unit reactor at the Chernobyl nuclear power plant

The results of an investigation of structural fragments of the core of the destroyed reactor in the No. 4 unit of the Chernobyl nuclear power plant are analyzed. It is shown on the basis of investigations of the fission product distribution over the cross-section of the graphite blocks and the determination of the physical properties of graphite that the temperature of the graphite blocks, including the reflector, at the moment they were ejected from the core exceeded 1000°C. The heat content of the fuel was estimated on the basis of an analysis of fragments of the dispersed uranium dioxide particles and an analysis of possible graphite dispersal mechanisms at the moment of the explosion. It is shown that energy sufficient for dispersing and partial vaporization of the fuel and for dispersing the graphite could have been introduced into the fuel during the accident process. Analysis confirms the possibility of a core destruction scenario with ejection from the shaft and ejection of part of the fuel in the form of vapor and dispersed particles into the atmosphere. __________ Translated from Atomnaya énergiya, Vol. 104, No. 6, pp. 319–328, June, 2008.     

A comparative study of the MATRA-LMR-FB calculation with the SABRE result for the flow blockage accident in the sodium cooled fast reactor

A sub-channel flow blockage may be initiated by an ingression of damaged fuel debris or foreign obstacles into a core subassembly for the sodium cooled fast reactor (SFR) due to the compact design of the fuel arrangement. Since local coolant temperature could go up high enough to reach a safety limit by the blockage disturbance in the subassembly, the MATRA-LMR-FB code was developed to analyze such blockage effect. An effort has been undergoing to enhance its reliability.In this study, a code-to-code comparison analysis with another code, SABRE4, was performed to supplement a qualification of the MATRA-LMR-FB. The two codes were applied to the analysis of partial sub-channel blockage accidents in a subassembly of the KALIMER-150, which is a conceptual design of a sodium-cooled fast reactor with an electric output of 150 MW. The analyses were carried out not only for radially different blockage positions but also for different blockage sizes in the subassembly.In result, the two code results were generally agreed both in magnitude and trend within a range. Therefore, it was concluded that the comparison results could support complementarily the applicability of the MATRA-LMR-FB to the partial flow blockage accident in the subassembly of the SFR.     

Real time reactor noise diagnostics for the Borssele (PWR) nuclear power plant

After the upgrade of Borssele NPP in 1997, core cycle 24, the power plant operated three years more with 91% availability. The authority of the power plant decided to enhance and upgrade the reactor trend monitoring and plant information recording system with higher frequencies than the plant data processing system (PPS) as well as installing a flexible and multiple-purpose reactor noise analysis system which may support the reactor maintenance group with on-line and off-line capabilities for several different signal processing applications. Two measuring and monitoring systems were built in 2001 and fully taken in implementation during the start-up of the new core 28. In this sense, the new system was used in power operation during the 29^th of September 2001. This paper will introduce the measuring system, the operational tasks, and the results obtained so far on the real-time core-barrel motions (CBM) and the two-primary coolant pump vibrations measured through the reactor noise analysis.     

压水堆核电站堆芯集中参数模型的微机仿真

阐述了PWR核电站堆芯的模型化问题,提适用于微机仿真的核电站堆芯的物理数学模型,将核电站堆芯分为三大块分别建立模型,中子动力学模块,反应性反馈模块,堆芯热力学模块,建立系统传递函数,运用MATLA仿真,得到良好结果。