Programs and methods for testing in the MIR research reactor fuel elements of water-cooled reactors under conditions simulating transient and emergency regimes

Alekseev, A. V.; Burukin, A. V.; Izhutov, A. L.; Kalygin, V. V.; Kiseleva, Irina; Ovchinnikov, V. V.; Shulimov, V. N.

doi:10.1007/s10512-012-9614-6

Cited by 3 publications

(2 citation statements)

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“…Up to the present, experiments have been carried out in the reactor channels to study the behavior of VVER-1000 fuel elements at ramp and step-like power increases, cyclic power variations (Alekseev et al 2007), and in design-basis loss-of-coolant accidents (Alekseev et al 2012, Goryachev et al 2004). In addition, a series of reactor experiments was intended to determine the yield of fission products from fuel elements with artificially applied defects on their jackets (Burukin et al 2009).…”

Section: Introductionmentioning

confidence: 99%

Outcomes of the “steady-state crisis” experiment in the MIR reactor channel

Alekseev

Dreganov

Izhutov

et al. 2019

NUCET

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To license nuclear fuel for nuclear power plants, data on the behavior of fuel elements (FE) under design-basis accidents are required. These data are obtained during tests of fuel assemblies (FA) and single fuel elements in research reactor channels followed by post-test studies in protective chambers. A reactivity-initiated accident (RIA) with an unauthorized release of CPS rods from the reactor core leads to a pulsed channel power increase. This accident can proceed according to two scenarios: without a critical heat flux (CHF) on the fuel element jacket at the final stage and with a dry heat flux. To date, a series of experiments have been carried out according to the first scenario in the MIR reactor channel and the corresponding data on the behavior of fuel elements have been obtained. An urgent task for today is to prepare and conduct reactor experiments according to the second scenario. The main experimental parameter that determines the behavior and final state of the studied fuel elements is their temperature. No experimental data were found on the critical heat flux for the rod bundles in the low coolant mass flow rate region (experiments in the MIR reactor channel can be conducted in the range of 200–250 kg/(m2s)). The available data are in the extrapolation range. The “steady-state crisis” experiment was conducted to obtain data on the critical heat flux value within the specified coolant mass flow rate range in the MIR reactor channel. The test object was a jacket fuel assembly composed of three shortened VVER-1000 fuel rods with a length of 1230 mm (the fuel part length = 1000 mm) installed in a triangular grid at a pitch of 12.75 mm, which is a cell of the VVER-1000 core. This assembly configuration is used for in-pile tests to study the behavior of fuel elements under emergency conditions. The in-pile testing results are presented. The paper shows the possibility of detecting the start and development of a dry heat flux based on the readings of thermocouples located inside the FE kernel. As a result, the directly measured test parameters were used to determine the critical heat flux value.

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Section: Introductionmentioning

confidence: 99%

Outcomes of the “steady-state crisis” experiment in the MIR reactor channel

Alekseev

Dreganov

Izhutov

et al. 2019

NUCET

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“…To study the behavior of the VVER-1000 reactor fuel elements during normal operation and anticipated operational occurrences, there is a whole class of reactor experiments (Alekseev et al 2007a;Alekseev et al 2006a, b;Alekseev et al 2012), in which the item of interest shall be tested in a strictly defined temperature range. This is especially true in the event of a fuel element test in conditions of a loss-of-coolant accident (LOCA) (Goryachev et al 2004, Alekseev et al 2009, Salatov et al 2013, Goryachev et al 2004a, in which the cladding temperature varies in a temperature range of 700-900°C for different experiments with an accuracy of not more than 2-5%.…”

Section: Introductionmentioning

confidence: 99%

Measurement of the spent fuel rod cladding temperature during the in-pile testing at 500 – 900°C

Dreganov

Shulimov

Kiselyova

et al. 2018

NUCET

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This paper deals with the problem of measuring the VVER-1000 burnup fuel cladding temperature in a 500–900°C range in the process of experiments in a channel of the MIR research reactor to obtain data on the fuel element behavior under the influence of the parameters typical of the maximum design-basis loss-of-coolant accident (LOCA). Studying the burnup fuel cladding deformation pattern requires measurements of the cladding temperature with no (thermal, mechanical and other) impacts on the cladding in the maximum deformation region. For dynamic experiments in the MIR reactor channel with fuel testing in a vapor-gas environment, a cladding temperature measuring unit has been developed, in which the cladding is not subjected to external impacts in the maximum deformation region. In the process of being installed into the spacer grid, the thermoelectric transducer (TET) has its hot junction forced against the cladding making it possible to prevent the external impact on the cladding. The thermometric characteristic of the TET attachment, which is associated with the impact of the grid as such on its thermal condition, was studied using a laboratory facility. This technique was used in an in-pile experiment to study the fuel cladding deformation pattern.

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