Determining the three-dimensional energy output field in an RBMK-1500 fuel assembly

Bronitskii, L. L.; Postnikov, V. V.; Прошин, Владимир; Vorontsov, B. A.; Chernyshev, V. G.

doi:10.1007/bf01138223

At Energy

1991

DOI: 10.1007/bf01138223

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Determining the three-dimensional energy output field in an RBMK-1500 fuel assembly

L. L. Bronitskii¹,

V. V. Postnikov²,

Владимир Прошин³

et al.

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Cited by 3 publications

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“…In the SKALA in-reactor monitoring system, the critical power of fuel assemblies was calculated using a relation that neglected the form of the axial distribution of the power release blies depends strongly on the form of the axial distribution of the power release, and for this reason for this reactor the measurement means were improved, detailed methods of calculating the axial power release distribution were derived and methods for calculating the critical power of fuel assemblies were improved [12,13]. Similar changes in the calculations of the critical power of fuel assemblies were also made in the algorithms used in the SKALA-mikro information and measurement system in RBMK-1000 to increase accuracy.…”

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confidence: 99%

“…During the time that RBMK-1500 operated, there were no programs for performing a three-dimensional neutronphysical calculation that were required to accurately reconstruct the axial distribution of the power release. For this reason, a semiempirical method of reconstruction based on a harmonic expansion of the axial distribution of the thermal neutron flux density was used for this purpose [12]. Subsequently, a computational and experimental procedure using a three-dimensional neutron-physical calculation started to be used in the SKALA-mikro informational and measurement system to reconstruct the axial distribution of the power release, but the need to continually use the semiempirical procedure to correct the results in the period between successive calculations remained.…”

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confidence: 99%

“…*** Includes scanning error. thermal neutron flux density is expanded, is also used to reconstruct the axial distribution of the neutron flux density in the semiempirical and computational-experimental methods [12].…”

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Development of RBMK in-reactor control

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The most important component of reliable and safe operation of a nuclear reactor is controlling its distributed and general parameters. Different in-reactor control systems used previously and currently in NPP with RBMK are examined. The means for controlling the power release in the reactor core are examined and the methods for reconstructing the power release in RBMK are reviewed briefly.In the development of RBMK, significant attention was devoted to devising means for monitoring the distributed parameters of the core, such as the neutron flux density, coolant flow rate in fuel channels and graphite temperature, and computing systems capable of real-time calculations of safety-important parameters which cannot be measured directly, specifically, the power, margin to crisis of heat transfer, and maximum lineal power of a fuel channel [1]. The urgency of developing detectors of a new type, computational means and information processing methods was determined for RBMK by the absence of means for heat-engineering monitoring of the power of a fuel channel with boiling coolant, unstable power release and refueling an operating reactor with a large perturbation of the distributed parameters of the core.Research on in-reactor detectors and methods of processing their indications, performed at the Beloyarskaya NPP, where at the first stage coaxial γ-sensitive ionization chambers intended for monitoring the fuel-channel power were tested, played a large role in the development work. Chambers of this type with a 6 m long sensitive part, central electrode and housing, which are separated from one another by spacing insulators made of aluminum oxide, failed after operating for 3-6 months because the resistance of the insulation decreased. Subsequently, they were replaced with triaxial chambers with a guard electrode in the coupling line and in the working volume and did fail because of a drop in the resistance of the insulation [2]. At the Beloyarskaya NPP, β-emission detectors with silver emitters successfully underwent tests and subsequently served for more than 10 years as the main means for monitoring the power release in RBMK [1].The power release in the channels of RBMK built in the first phase was measured with a system performing physical monitoring of the power release distribution and centralized monitoring system SKALA, which performed mathematical processing and displayed information for the operator in a form convenient for controlling the parameters of the core.The system performing physical monitoring of the power release distribution contains 130 radial monitoring detectors D.42 with a 7 m long sensitive part, uniformly arranged along the core in the central dry sleeves of the fuel assemblies 49 and 12 height monitoring assemblies 156, placed in the cooled channels of the CPS. Each assembly 156 has seven sections are uniformly distributed along the core height, each comprising a cable and sensor with a silver filament of total length 2.6 m formed into a 36-mm long spiral. A dry sleeve for periodic calibrati...

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Development of RBMK in-reactor control

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Computational modeling of the accident in the fourth power-generating unit of the chernobyl nuclear power plant

Podlazov¹,

Trekhov²,

Cherkashov³

et al. 1994

At Energy

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During the accident in the fourth power-generating unit of the Chernobyl nuclear power plant complicated spatially distributed processes (neutron-physical, thermohydrodynamic, chemical, and thermomechanical) were focused and became intertwined. This has made it difficult to model the accident numerically and it has made international collaboration in this field urgent.Experience gained in the joint work performed by ENEA and the Scientific-Research and Design Institute of Electrotechnical Machinery [1] showed that organization of such investigations requires a parallel solution of a set of complicated scientific-methodological problems with the participation of different types of specialists. In Russia several groups of investigators used different programs (codes) to perform difficult calculations of the first phase of the accident. The main results of these investigations were summarized at the Paris conference in April 1991 [2]. The conclusions drawn there about the accident can be briefly formulated as follows:The main factors influencing the development of the accident were the high positive steam coefficient of reactivity and deficiencies in the construction of the safety and control rod system which came to light during the irregular state of the reactor prior to the accident; spatial dynamic processes played an important role in the development of the accident; to reproduce a real accident process, the physical characteristics must be reproduced in detail throughout the entire volume of the reactor in the period prior to the accident; and the codes used for comprehensive investigations of the first phase are incomplete and they must be further updated and verified.Specialists in different countries performed a series of methodological investigations of the effect of different factors on the positive reactivity arising as a result of the insertion of the safety and control rods [3][4][5][6][7]. These works confirmed that the positive reactivity is highly sensitive to the state of the core prior to the accident and they substantiated the need for reproducing in detail the preliminary initial conditions during computational modeling of the first phase of the accident.The first stage of a combined comprehensive computational analysis of the Chernobyl accident were quasistatic estimates of the positive reactivity according to the DINA [8] and CITATION [9] computational codes. The results of the reconstruction of the three-dimensional neutron fields on the basis of information recorded approximately 2 min prior to the accident by the SKALA system were used as the initial information for constructing the preaccident state of the reactor.Formulation of the Problem. The main problems addressed in the investigations were as follows: formation of the initial state of the reactor, so as to agree correctly with the data recorded; to make, using these initial data, quasistatic estimates of the positive reactivity according to different three-dimensional programs (DINA and CITATION) on the basis of a polycell library ...

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Axial neutron flux density distribution in RBMK-1000: reconstruction using the Prizma-M program

et al. 2013

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A computational-experimental procedure for reconstructing the axial neutron flux density and powerrelease distributions using the Prizma-M program is described. An approach to reconstructing the axial neutron flux density and power-release distributions in the RBMK-1000 assembly channels by using extended detectors with a hafnium dioxide emitter to scan the core is presented. This approach made it possible to compare the reconstruction accuracy of the semi-empirical and computational-experimental procedures used in Prizma-M. It is shown that the reconstruction error for the axial neutron-flux density distribution does not exceed the attested error of the Prizma-M program.A method of reconstructing the axial neutron flux density and energy release distributions was adopted in 1985 in RBMK-1500 at the Ignalina NPP. Based on a semiempirical method, it uses a harmonic expansion of the flux density distribution without a neutron-physical calculation and signals from eight neutron detectors belonging to the control-and-protection system (CPS) and uniformly distributed over the height of each of 20 assemblies [1]. Ionization chambers placed in (n, γ) converter channels served as fast-response neutron flux density detectors. In 2002, this method was incorporated into RBMK-1000 in the newly introduced program Prizma-M of the SKALA-Mikro information-and-measurement system. A special feature of the Prizma-M program here was that 72 four-section in-reactor height-monitoring detectors were used. Each section with a hafnium dioxide emitter had a sensitive part with length equal to 1/4 the core height. The length of the communication line section located in the core was equal to 3/4 the core height. The top and bottom parts of the communication line were connected to the emitter, which made it possible to decrease the uncertainties related with the background signal of the communication line.For the first nine years, a semiempirical method was used in the Prizma-M program. Subsequently, on the basis of positive experience in reconstructing the radial-azimuthal distribution by means of a neutron-physical calculation, the computational-and-experimental procedure was adopted for reconstructing the axial distribution of the neutron flux density and energy release, retaining the semiempirical method in the auxiliary operations.In the computational-and-experimental method, just as in the semiempirical method, an expansion of the form (1) is used for the axial flux density distribution of thermal neutrons in the jth assembly channel. Here k is an index of a point along the core height, k = 1, 2, ..., 28; h k is the distance to the kth point from the core top, m; B j is a vector of the coefficients in the harmonic expansion of the axial neutron flux density distribution in the jth assembly channel; ϕ T (h k ) = [ϕ 1 (h k ), ϕ 2 (h k ),

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Determining the three-dimensional energy output field in an RBMK-1500 fuel assembly

Cited by 3 publications

References 4 publications

Development of RBMK in-reactor control

Development of RBMK in-reactor control

Computational modeling of the accident in the fourth power-generating unit of the chernobyl nuclear power plant

Axial neutron flux density distribution in RBMK-1000: reconstruction using the Prizma-M program

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