1111 CA9600926 VOID-REACTIVITY MEASUREMENTS IN THE ZED-2 REACTOR USING THE SUBSTITUTION METHOD

M.B. Zeller Physicist Reactor Physics Branch AECL Research Laboratories Chalk River, , KOJ 1J0

G.P. McPhee Facility Supervisor, Reactor ZED-2 Reactor Physics Branch AECL Research

R.S. Davis Physicist Reactor Physics Branch AECL Research

A. Celli Section Head, Experimental Physics Section Reactor Physics Branch AECL Research

1. ABSTRACT

Experiments are planned in the ZED-2 reactor to measure the reactivity change resulting from the loss of coolant in channels containing coolant and fuel at conditions that simulate those found in an operating CANDU* reactor. For reasons of economics, substitution measurements will be employed for the study. The purpose of the present work is to establish the accuracy of the substitution method.

The substitution method has been used to measure critical bucklings for lattices consisting of 28-element -dioxide assemblies, 7-element uranium-dioxide assemblies, and 19- element uranium-metal assemblies. Two test-fuel coolants were studied: and air. The test channels were arranged hexagonally at a 31-cm spacing, and were driven by channels containing 28-element uranium-dioxide bundles.

The bucklings are compared to previous results derived from flux-map measurements in uniform lattices. The comparison shows that the substitution-derived buckling-change-on- voiding results for all three test-fuel types are accurate, within the reproducibility error associated with the flux-map method. These results demonstrate the suitability of using the substitution method to study void reactivity in CANDU-type channels.

* CANDU: CANada Deuterium Uranium. Registered Trademark. 2. INTRODUCTION

Many of the physics data for CANDU reactors are measured in the ZED-2 reactor at Chalk River Laboratories. In the past, ZED-2 experiments have been performed with a sufficient number of fuel assemblies to construct uniform critical lattices. Recently, there has been an increased requirement to perform measurements using assemblies with special characteristics, and there is an economic incentive to use fewer test assemblies for physics experiments.

The current focus of the ZED-2 program is to study void reactivity in CANDU. The goal is to obtain experimental data appropriate to conditions in an operating power reactor. These data are to be used as a check against the reactor-physics codes and associated data bases used in the study of loss-of-coolant accidents.

Examples of test assemblies used for void reactivity studies include: fuel channels modified to allow for the continuous variation of coolant density, channels in which stratified flow conditions are simulated, assemblies adapted to allow for void-reactivity measurements at CANDU-coolant temperatures and pressures, and mixed-oxide (MOX) fuel containing (and possibly simulated fission products) that resembles CANDU fuel at mid- burnup.

A non-uniform critical lattice, containing only a few special assemblies, can be used to study the reactivity effects caused by channel voiding by employing the substitution method. The method involves the systematic substitution of the test assemblies into a reference lattice, and observing the resulting change in lattice critical size as the substitutions proceed. Void reactivity is studied by performing the substitutions using both heavy water and air (void) coolant in the test assemblies.

To establish that substitution void-reactivity experiments can be used as a viable alternative to full-core flux-map experiments, substitution studies have been performed in the ZED-2 reactor using a number of different test assemblies. This paper describes the basic approach used in performing both a typical flux-map experiment and a substitution experiment. Substitution-derived buckling-change-on-voiding results for various test assemblies are presented, and compared to results obtained from full-core flux-map studies. The comparisons are used to demonstrate the applicability of the substitution method for measuring void reactivity for CANDU-type channels.

3. EXPERIMENTAL SET-UP AND PROCEDURE

3.1 General Description of the ZED-2 Reactor

The ZED-2 is a tank-type critical facility used primarily to study the physics of heavy-water-moderated lattices. It consists of a cylindrical aluminum calandria, approximately 3.3 m in diameter by 3.3 m deep. Fuel assemblies are suspended from movable beams that are positioned across the top of the reactor. The reactor is made critical by pumping moderator into the calandria, and is controlled by adjusting the moderator level. Typical moderator critical levels (calandria floor to moderator surface) range from 200 to 250 cm. ZED-2 is a flexible facility, allowing experimental assemblies to be easily inserted, removed, and rearranged. The maximum allowed reactor power is 200 watts. This corresponds to a neutron flux of about 109 n cm"2 s'\ which is sufficient for performing foil activation experiments, but low enough that the fuel does not become significantly activated. Accessibility to the core is therefore possible soon after reactor shutdown.

Figure 1 shows a plan view of the standard ZED-2 lattice. The fuel charge is 275 natural- uranium-dioxide 28-element bundles that are contained in pressure/calandria-tube assemblies. The channels are arranged in a hexagonal lattice at a 31-cm spacing (equivalent to the square pitch of CANDU pressurized heavy-water reactors). The lattice is surrounded, both radially and below, by heavy-water and graphite reflector.

Figure 2 shows a plan view, and a vertical cross section, of a standard assembly. The bundle geometry is Pickering-type, and each assembly contains a five-bundle string, with the bottom of the fuel suspended 15 cm above the reactor floor. An endplug opening at the bottom of each assembly allows moderator into the "channels", to provide "coolant" for the fuel. When the endplugs are inserted, moderator is prevented from entering the channels, so that the fuel remains voided. The 28-element assemblies are described in reference 1.

3.2 Measuring Lattice Critical Size by Flux Map

If a sufficient amount of fuel is available to construct a uniform critical lattice, it is possible to measure lattice buckling directly by performing a flux-map experiment. The procedure is to position detector foils interstitially across the lattice (see Figure 1), and then operate the reactor to activate the foils.

Pure-metal copper foils are typically used for flux-map experiments and the reaction Cu63(n,7)Cu64 is employed to activate the foils. The cross-section curve for this reaction is one-upon-v and the activation data are used to map out the global flux shape, both radially and axially through the lattice.

Figures 3 and 4 show plots of (relative) activation data obtained from an experiment to measure the critical buckling for the standard ZED-2 lattice. For this experiment, the bottom endplugs were removed from the channels so that the fuel was cooled with heavy water.

The radial buckling is determined by fitting the activation data radially to a Bessel function. The fitted curve is used to extrapolate to the point of zero flux, which defines the extrapolated radius of the lattice, R,,,.

The axial buckling is determined by fitting the activation data axially to a cosine function. The cosine curve is used to extrapolate to the zero-flux points, above and below the lattice. The distance separating the upper and lower extrapolated boundaries defines the extrapolated height of the lattice, Hev

The lattice critical buckling, B2, is obtained from the extrapolation boundaries using the following expression:

B2 = (2.405/RJ2 + 3.3 Measuring Lattice Critical Size Using the Substitution Method

Often, flux maps cannot be used to measure the critical buckling for a test lattice, either because of cost or because of the critical size of the lattice. An alternative method has been developed which uses smaller quantities of fuel than are required to make a uniform critical assembly, yet which allows the critical buckling to be determined. This "substitution" method typically uses seven channels (35 bundles) of the test fuel.

The critical size of a test-fuel lattice is determined by systematically substituting test-fuel assemblies into a reference lattice, and observing the change in lattice critical size as the substitutions proceed.

Figure 5 shows a plan view of a substitution lattice. The seven centre channels make up the substitution region. These are surrounded by "driver" channels containing reference fuel with known properties (measured previously by flux map). Also shown are the various configurations of test and reference fuel in the substitution region used for a typical substitution experiment.

A pure lattice containing only reference fuel is measured initially and the reference-lattice critical size is determined. The centre channel in the lattice is then replaced with a channel containing test fuel, and the critical size of the one-rod substitution lattice is measured. Measurements are then performed using three-, five-, and seven-rod substitution configurations.

Figure 6 shows a plot of moderator critical-height data versus the number of substitution channels obtained from a substitution experiment to measure the critical buckling for the standard ZED-2 lattice. The critical height for zero substitution rods corresponds to the reference-lattice measurement. The reference fuel for this experiment is air-cooled 28- element uranium-dioxide bundles, in channels arranged at a 31-cm spacing. The test fuel is heavy-water-cooled 28-element assemblies. The substitutions thus involve replacing air- cooled channels with heavy-water-cooled channels.

Some information concerning the fuel can be derived immediately, without resorting to a detailed analysis of the critical-height data. For example, the fact that the critical height increases with the number of substitution channels indicates that the flooded fuel is less reactive than the voided fuel, implying a positive void coefficient. However, precise information concerning the critical buckling for the flooded lattice, and the buckling-change- on-voiding, requires detailed analysis, using a core code.

3.4 Analysis of Substitution Experiments

The core code used for the substitution analyses described in this work is the 3-D diffusion code CONIFERS [2], using cell parameters provided by the lattice code WTMS-AECL [3]. CONIFERS has a number of properties that make it particularly suitable for substitution analysis: geometrical flexibility, flexibility of reactivity scales, and the ability to use any number of energy groups. The present study used four energy groups and modelled the lattice using homogenized cells, employing the finite-difference method. The first step is to perform a calculation that models the reference lattice. Boundary conditions are set, and neutron-production rates in the reference fuel are adjusted to make the model consistent with experiment. The requirement is that the calculated flux shape in the reference lattice agrees with that measured (by flux map), and that neutron balance is achieved in the calculation (i.e., k-effective calculated for the reference lattice is unity). The calculation yields a parameter that is used to renormalize the neutron yield in all cells containing reference fuel in subsequent substitution-lattice calculations.

The substitution lattices are modelled using the measured moderator critical heights to define the lattice critical size for each calculation. The reference-fuel neutron yield is renormalized according to the reference-lattice calculation, and the neutron yield in all cells containing test fuel is renormalized so that neutron balance is achieved for each substitution-lattice calculation.

The renormalized neutron yield for the test fuel, derived from each substitution-lattice calculation, is affected by neutron interaction between adjacent reference- and test-fuel assemblies. First-order perturbation theory can be used to define a contamination parameter that specifies the amount of interaction. The contamination parameter is equal to the average number of reference-fuel assemblies that are adjacent to each test-fuel assembly.

The renormalized neutron yields from each substitution-lattice calculation are plotted versus the contamination parameter, and a fitted curve is used to extrapolate to the point of zero contamination. This point corresponds to the renormalized neutron yield appropriate for a uniform-lattice calculation. Weighting factors, equal to the square of the number of substitution channels in each lattice, are used for the fitting procedure. This weighting is to account for the increased accuracy of the substitution data as the number of substitution channels increases.

At the end of the procedure, CONIFERS can be used to calculate the global flux shape in a (hypothetical) uniform test-fuel lattice. A Bessel- and cosine-function fit to the calculated flux shape defines the critical buckling for the test fuel.

3.5 Results of the Analysis

Figures 7 and 8 show plots of C/E (calculation-upon-experiment) ratios, comparing measured and calculated copper-capture rates in the reference lattice (voided 28-element assemblies) used for the above experiment. The ratios in Figure 7 compare the calculated radial flux profile to the experimental data, and Figure 8 compares the calculated axial flux profile to the experiment. The typical discrepancy between calculation and experiment is under 1 percent (and never greater than 2 percent), indicating that leakage is being modelled correctly in the calculation.

A flux-map experiment was performed in the seven-rod substitution lattice (seven flooded 28- element assemblies driven by voided 28-element assemblies) to check that the flux shape calculated by CONIFERS is consistent with that measured. Figures 9 and 10 show plots of CONIFERS-calculated C/E ratios for the radial and axial copper-capture rates, respectively. The good agreement between calculation and experiment shows that the code is correctly calculating leakage in the lattice, and is reasonably successful in calculating the interaction between the flooded and voided assemblies.

The flux shape predicted for the uniform lattice of flooded 28-element assemblies, derived from the substitution analysis, is compared to the measured copper-capture data obtained from the flux-map buckling measurement (described in section 3.2). C/E ratios for the radial and axial copper-capture data are plotted in Figures 11 and 12. The good agreement between the prediction (based on the substitution analysis) and the copper-capture rates (measured in the standard lattice) demonstrates the applicability of the substitution method.

Figure 13 shows a plot of the buckling-change-on-voiding for the 28-element fuel, derived from the substitution of flooded assemblies into a voided reference lattice. Also plotted are results from a similar experiment involving the substitution of voided 28-element uranium- dioxide assemblies into a flooded reference lattice.

The buckling data are plotted versus the contamination parameter appropriate to each substitution configuration. The buckling data correspond to the renormalized neutron yields, obtained from each substitution-lattice calculation. The data are least-squares fitted and the fitted curve is used to extrapolate to the point of zero contamination, corresponding to a uniform lattice of the test fuel.

The agreement between the substitution-derived buckling change and the flux -map result is excellent. The uncertainty for the flux-map result is based on the reproducibility error for flux-map measurements. Both substitution-derived results agree with the flux-map result, within this reproducibility error.

4. SUBSTITUTION MEASUREMENTS USING 7-ELEMENT URANIUM-DIOXIDE ASSEMBLIES AND 19-ELEMENT URANIUM-METAL ASSEMBLIES

4.1 Outline

Similar substitution measurements were performed using 7-element uranium-dioxide assemblies and 19-element uranium-metal assemblies as the test fuel. Figure 14 shows plan views of the test assemblies, which are described in references 4 and 5.

The reference lattice for these measurements comprised 28-element uranium-dioxide assemblies, arranged at a 31-cm spacing. Experiments were performed using both heavy- water and air coolant in the reference lattice.

Two sets of substitution experiments were performed. Initially, the test assemblies were cooled with heavy water. The measurements were then repeated with air coolant to study the reactivity effect of voiding the test channels. 4.2 Results and Conclusion

The results of the substitution analyses are plotted in Figures 15 and 16, and compared to results obtained from full-core flux-map experiments [4,5]. In all cases, the buckling - change-on-voiding derived from the substitution experiments agrees with results obtained by flux map, within the reproducibility error for the flux-map method. The major conclusion of the present study is that the substitution method can be used to measure buckling change due to channel voiding with an accuracy equivalent to that achievable using flux maps.

5. ACKNOWLEDGMENT

The authors gratefully acknowledge the financial support of the CANDU Owners Group, Working Party 25.

6. REFERENCES

[1] K.J. Serdula, "Lattice Measurements with 28-Element Natural UO2 Assemblies. Part 1: Bucklings for a Range of Spacings with Three Coolants", AECL Report, AECL-2606, 1966 July.

[2] R.S. Davis, "CONIFERS: A Neutronics Program for Reactors with Channels, NOS/VE version", AECL Report, AECL-8567, AECL Research, in preparation.

[3] J.V. Donnelly, "WIMS-AECL. A User's Manual for the Chalk River Version of WIMS", AECL Report, AECL-8955, 1986 January.

[4] G.A. Beer and D.W. Hone, "Lattice Measurements with 7-Element UO2 Clusters, in ZED-2. Part 1: Bucklings over a Range of Spacings with Three Coolants", AECL Report, AECL-1505, 1962 January.

[5] K.J. Serdula and R.E. Green, "Lattice Measurements with 19-Element Natural Uranium Metal Assemblies. Part 1: Bucklings for a Range of Spacings with D2O and He Coolants", AECL Report, AECL-2516, 1965 October. 795432101

i FUEL CHANNELS CONTAINING 2«-ELEMENT BUNDLES STRINOERS FOR POSITIONING COPPER FOILS

Figure 1 - Plan view of the standard ZED-2 lattice

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0.98 20 40 60 80 DISTANCE FROM LATTICE CENTRE (cm) Figure 7 - CONIFERS-derived C/E ratios for radial copper-capture distributions measured in the reference lattice (voided 28-element UO2)

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0.98 25 50 75 100 125 150 175 200 DISTANCE ABOVE REACTOR FLOOR (cm) Figure 8 - CONIFERS-derived C/E ratios for axial copper-capture distributions measured in the reference lattice (voided 28-element UO2) 7-ROD SUBSTITUTION LATTICE RADIAL FLUX SHAPE

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0.98 25 50 75 100 125 150 175 200 DISTANCE ABOVE REACTOR FLOOR (cm) Figure 10 - CONIFERS-derived C/E ratios for axial copper-capture distributions measured in the seven-rod substitution lattice (flooded and voided 28-element UO2) RADIAL FLUX PROFILE ACROSS A UNIFORM LATTICE OF TEST FUEL 1.02

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0.98 0 25 50 75 100 125150175200225 DISTANCE ABOVE REACTOR FLOOR (cm) Figure 12 - C/E ratios comparing substitution-derived CONIFERS calculations to axial copper- capture distributions measured in the standard lattice (flooded 28-element UO2) 28-ELEMENT URANIUM-DIOXIDE FUEL

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Figure 16 - Comparison of substitution-derived buckling-change-on-voiding, using the 19-element assemblies, to a measurement obtained by flux map