EXPERIMENTAL CAPABILITY FOR INVESTIGATIONS OF THERMAL-HYDRAULIC PROCESSES AND CRITICAL HEAT FLUXES ON FULL-SCALE MODELS OF ROD ASSEMBLIES FOR SMALL MODULAR REACTORS

J.A. KUZINA, A.I. GONIN

State Scientific Center of the Russian Federation –

Leypunsky Institute for Physics and Power engineering IPPE

Obninsk, Russian Federation

Email: aigonin@ippe.com

Abstract

The report presents the capabilities and description of the experimental high-pressure stand SVD-2, available at the Physics and Energy Institute named after Leipunsky JSC "SSC RF-IPPE", designed to conduct studies of hydrodynamics and heat transfer, including critical heat fluxes, under stationary and dynamic conditions on various models of the fuel rod assemblies (FA), including full-scale models for small modular reactors with water coolant.

INTRODUCTION

The most developed technology in nuclear energy is light water-cooled reactors. This type of reactor is used at more than 80% of the world's nuclear power plants, and in most cases - on transport ships and installations. It is not surprising that almost half of all SMR projects are pressurized water reactors, which in terms of technological readiness and legitimization are most competitive if to compare with all other types of reactors. It means that in SMR segment water-cooled reactors are the most promising and closest to start their commercial operations. Of the 83 SMR projects in 18 countries, water-cooled projects number is 38 [1].

SVD-2 has unique capabilities for performing research in terms of the range of operating parameters and design solutions for test sections, including models of fuel assemblies (FA) that are as close as possible to the ones according project. Table 1 presents the range of main operating parameters available for conducting experiments on SVD-2. [2].

TABLE 1	OPERATING PARAMETERS SVD-2

PARAMETER	VALUE
Input Power, MWt	10
Pressure, MPa	25.5
Inlet Temperature of coolant, оС	450
Outlet Temperature of coolant, оС	500
Coolant flow rate, м3/h	70
`Measuring channels number	384

Thus, parametric capabilities make it possible to conduct studies in all operating modes of existing and designed light water reactors, including SMRs.

Pressure drops along the test sections, the efficiency of the coolant mixing by measuring the flow rate and temperature in different sub-channels of the cross section of the models, as well as critical heat fluxes are studied at the SVD-2.

We must understand that one of the most important purposes of the SVD-2 facility is the study of the critical heat fluxes. It is necessary to repeat about importance of such researches from the point of view of the safety of nuclear reactors. Drying of the fuel rod cladding with subsequent sharp and significant deterioration in heat transfer to steam instead of water leads to its possible destruction and the occurrence of emergencies. Experimentally justified assessments ​​of the burnout ratio before the heat transfer crisis are a critically important value responsible for the safe operation of nuclear power facilities.

On the other hand, the burnout ratio cause of heat transfer crisis greatly limits the thermal power and, consequently, the efficiency of the nuclear power facilities, as before mentioned, for safety reasons. Working to shift potentially the burnout ratio before a heat transfer crisis through design changes can give significant economic benefits. However, any improvements and, especially, the development of new designs must be justified carefully experimentally from the point of view of permissible values ​​of heat fluxes.

In this sense, SVD-2 has a number of obvious competitive advantages:

With increasing flow rate through the core, the values ​​of critical heat fluxes increase, which makes it possible to shift the burnout ratio. In addition, what about an increase in pressure, among other things, increases efficiency due to higher coolant temperatures, which means having experimental data at high values of the main operating parameters, is very relevant. As we can see from Table 1, at SVD-2 it is possible to achieve supercritical pressure values ​​of up to 25 MPa and provide coolant flow rates of up to 70 m3/hour. The fact is that the test loop of SVD-2 consists of two independent loops of high and medium pressure (HPL and MPL) (Fig. 1), which, depending on the goal, can work in concert, doubling the powers of heat removal and coolant flow rates getting of up to 70 m3/hour. In this case, the experiments are carried out in conditions when combining abilities of both loops. On the other hand, the connection diagram allows experiments to be carried out on each loop independently with its own working section, and then it is possible to achieve supercritical pressure values ​​of up to 25 MPa on the HPL loop.

P~25 MPa G= 70 м3/h

FIG. 1	Heat remove at SVD-2 diagram.

HPL – medium pressure loop; MPL – medium pressure loop; ICL – intermediate loop ; OCL – cooling circuit; 1-2 - intermediate loop pumps; 3-1– intermediate loop heat exchangers; 4 – water-cooling tower ; НЦ1 - circulation pumps; АТ-1 — heat exchangers; Н-1 - cooling circuit pumps; УР2 – experimental model.

Data obtained by conducting studies on the FA models that are as close as possible in their design to the project ones are the most representative for the justification of reactor designs. For these purposes, combined with high requirements for basic operating parameters, there is a specified need for significantly increased energy supply. Electric power of up to 10 MW, which the SVD-2 has, makes it possible to study critical heat fluxes on full-scale height models of FA consisting of up to 25 rods for PWR and of up to 37 rods - for VVER.

FA model consisting of 37 rods	FA model consisting of 25 rods.

FIG. 2	Types of FA models and spacer elements for them studied on SVD-2.

Under all the same conditions, the power supplied to test section at the SVD-2 is sufficient to carry out thermal-hydraulic studies in the post-dryout heat transfer zone.

And in the framework of today’s conference is especially important that experiments in all described above areas are being successfully carried out at the SVD-2 using generally full-scale models of FA for light water small modular reactors (LWSMR).

DESIGNE AND OPERATION ABILITIES OF SVD-2

The schematic diagram of the SVD-2 is shown in Fig. 3 [2]. The facility is designed to conduct studies under both stationary and dynamic conditions and includes three main circuits: a high-pressure test circuit, an intermediate cooling circuit and a water-cooling circuit with a cooling tower.

FIG. 3 SVD-2 flow diagram.

1-1...1-3 – intermediate loop pumps; 2 – feed pump; 3-1...3-3 – intermediate loop heat exchangers; P, P0 – pressurizers; Н1...Н4 – cooling circuit pumps; CP1...CP5 – circulation pumps; H1, H2 – electric heater-ups; EM2 – experimental model; THR – thyristor rectifier; TFR – transformer; HE1...HEЗ – heat exchangers; MF1 – mechanical filter; 11, 12 – distillate storage tanks; 13 – chemical water treatment pump; 14 – ion-exchange filter; 15 – feed pump; 18 – submersible pump.

Specifically to perform the experimental studies on models for SMR FA, where fuel rod imitators are often very thin-walled tubes with a wall thickness of 1 mm or less, a loop of gas balancing system was developed and installed to prevent the destruction of the fuel rod imitators under external coolant pressure (Fig. 4.)

FIG. 4 Gas balancing system schematic.

1 – circulating pumps; 2 – feed pumps; 3 – pressurizers; 4 – test section; 5 – moisture separator A2; 6 – equalizing tank A1; 7 – level meter (Y1); 8 – pressure difference cell (Y2); 9 – gas pressure gauge; 10 – water pressure gauge.

The fuel rod imitators are heated during experiments by directly passing an electric current through the walls of the imitators.

At the SVD-2, measurements of electric voltage U and electric current I on the test FA models are carried out; water pressure at the model inlet, Рin; and at the - output, Pout; water temperature at the inlet to the test section, Тin; and at the exit from it, Tout; water flow rate G; pressure drop along the height of the model ΔР; temperature of the experimental column surface Tk; temperature of pressure impulse lines Til; pressure Psh and temperature Tsh of the coolant on the water rate measuring orifice, water temperature in the subchannels of the test assembly; temperature of the inner surface of the walls of the imitators; temperature of the isothermal block, Tx. The measured parameters make it possible to take into account the influence of the hydrostatic column of pressure tap impulse lines on the result of reading pressure drop in height a test section, take into account heat losses and make corrections for the temperature dependence of the coolant properties and materials of model constructions not only in the main, but also in control experiments.

Systematization of experimental data occurs using a data acquisition and processing system (DAS) with a measuring channel (MC) accuracy class of 0.3%, consisting of a subsystem of measuring transducers (MT), a subsystem of measuring amplifiers (MA), a multichannel measuring system (MMS), running under the control of a personal computer (PC). The passage of the signal from the its registration moment followed by the processing with help DAS up to its transformation into a final physical value is shown in the example of measuring coolant flow rate in Fig. 5. Implemented speed of the DAS - up to 5 cycles over 384 channels per second [3].

Fig. 5. The block-diagram for coolant flow rate measurement.

1- constricting device (orifice); 2, 5 - “Sapphire-22DD” - type DP cell; 3, 6- C2-29 - type precise resistors; 4, 7- UT-70 - type measuring isolation amplifiers.

QUALITY ASSURANCE

The “Quality Assurance Program” (QAP) is developed, including taking into account the requirements of the IAEA, the American standard ASME NQA-1-2008 “Quality Assurance Requirements for the Use of Nuclear Facilities.” In accordance with the QAP, before starting the main research, the metrological characteristics of the Data Acquisition and Processing System (DAS) are determined, in order to avoid systematic errors, sensors are calibrated and experiments are carried out to determine heat losses from the external surface of the experimental column. While carrying out the main experiments, due to control events the stability and structural integrity state of the experimental column and FA model, the right validity of the DAS system working as well as thermal balance maintenance with an accuracy of ± 3% are under strict constant control.

Heat losses from the experimental column external surface were assessed both under conditions of power supply to the test section and under adiabatic conditions. The results of heat loss measurements were approximated by the relation:

Qloss = к∙(Tk – Tа), кВт,

where Qloss is heat loss from the column surface, Tk is the average temperature of the column external surface, Ta is the ambient temperature, k is empirical coefficient

Ignoring heat losses can lead to an incorrect assessment of the heat balance and to a significant overestimation of the outlet balanced steam quality.

Experiments to control repeatability showed that if critical power measuring under the same conditions several times one after another the difference in the measured values to be negligible.

Reproducibility control consisted of repeating experiments under the same conditions, after a significant period of 30 days or more. Moreover, during this period various types of events associated with carrying out repair and restoration work use to be happened. In most cases, the maximum discrepancy between the critical powers values measured at these different times did not exceed ± 5%.

In experiments, it is not allowed the deviation of the measured significant parameters to exceed from the values ​​​​set by the Quality Assurance Program by more than: inlet pressure ± 1.5 bar; mass velocity (flow rate) ± 3%; water temperature at the inlet to the FA model ± 1.5 o C.

Measurement errors are estimated on base of the theory of errors in single measurements. [4] Let us consider the accepted algorithm for such assessment using the case of coolant flow rate reading according to the scheme presented in Fig. 5.

The value of the coolant flow rate is determined by the formula:

Gi = Gmeas i KT Kg,

where	Gmeasi = KG + BG;

KG и BG – flow rate calibration of constricting device (orifice) coefficients;

Pi – pressure drop on orifice;

KT and Kg – are the factors accounting the water properties and orifice geometry dependence on temperature.

Pi = KP

where

KP – calibration coefficient, kPa/V

signal reading in the DAS, V

The error of measurement of flow rate can be estimated as:

где G = 0.5 % of reading – calibration error.

dGmeasi - ошибка измерения расхода, которая определяется в первую очередь ошибкой измерения перепада давления на расходомерной шайбе ΔPi:

where (∆P) = ± 0.25 % of reading – pressure drop (ΔP) transducer (DP cell) calibration error;

А = ± 0.15 % of full scale – the UT-70 – type amplifier error;

S = ± 0.1 % - of full scale - the DAS error;

R = ± 0.15 %; - the Ro resistor error;

- the amplifier output voltage corresponding to ΔPi;

Um - normalizing voltage of amplifier and DAS.

Thus, the error is estimated taking into account the accuracy of all elements included in the chain of registration, conversion and signal transmission. The principle of measurement, the accuracy of calibration, the accuracy of each element included in the measurement channel (MC), as well as the zone of ​​the instrument scale in which the measured value falls, all of them are taken into account.

As a result, the relative error in measuring parameters in experiments is: pressure ± 0.5%; flow rate ± 2.0%; pressure drop ± 1.0%; electric power ± 1.0%; temperature ± 1.0%.

EXPERIMENTS ON SMR MODELS.

To date, a number of experimental studies of coolant flow mixing, pressure drop and critical heat fluxes in a wide range of operating parameters on full-scale models of SMR FA models have been carried out at the SVD-2 within the framework of international contracts. In particular, the influence of the heat release shape profiling was studied both in the cross section and along the height of SMR FA models (Fig. 6).

Areas of cross section with increased heat generation

Cosine profile of heat generation in height

FIG. 6 Kinds of heat release profiles in cross-section and along height of SMR fuel assemblies’ models i n experiments conducted at SVD-2.

Heat generation profiling was performed due to different thicknesses of the walls of the rods imitators, which changed rods electrical resistance or their segments. For longitudinal profiling, rods were produced, consisting of a number of segments with different wall thicknesses in accordance with a given heat generation profile.

References

1. Vladimir ARTISIUK, Hadid SUBKI, Dohee HAHN, «IAEA SMR Agency-wide Initiative and SMR Platform», International Workshop on Infrastructure Development Issues for Small Modular Reactors & Microreactors Deployment, including Policy, Strategy, and Key Aspects, 27 November – 1 December 2023 Vienna, Austria.

O.F. Kukharchuk, A.P. Sorokin, A.A. Trufanov, V.L. Chernonog "Thermophysical facility base for nuclear energy FSUE «SSC RF - IPPE»” Leypunsky Institute for Physics and Power engineering IPPE, Obninsk, Russian Federation, 2014.

N.V. Dovedchik, V.L. Chernukhin, “High-pressure thermophysical facilities of the TFK-2 building. Performing measurements during experimental studies", internal report, Leypunsky Institute for Physics and Power engineering IPPE, Obninsk, Russian Federation, 2011.

R.I. Mofat “On the theory of errors in single measurements”, Theoretical Foundations, v. 104, No. 2, Moscow, 1982.