Results of Experiments under the Physical Start-Up Program of the IVG.1M Reactor

Abstract

In 2022, the physical start-up stage of the IVG.1M research reactor was successfully completed, initiated by reducing the fuel enrichment of ²³⁵U. This phase included the loading of nuclear fuel into the reactor core and conducting experiments to determine the neutron-physical characteristics of the reactor. Prior to the physical start-up, preliminary calculations were performed using the computational code MCNP6 and a full-scale model of the IVG.1M reactor with low-enriched uranium fuel (LEU). During the start-up series, the reactivity worth curves of the reactor control and protection system’s operating and compensating elements were determined. Additionally, experiments were performed to measure the reactivity effects of technological channel draining and to obtain activation reaction rates in the central experimental channel using nickel and gold activation indicators. The results of determining the neutron-physical characteristics of the IVG.1M reactor have confirmed the operability of the reactor core with LEU fuel.

1. Introduction

In relation to initiatives to reduce the proliferation of nuclear weapons, since 1978, many countries have decided to convert their research reactors to low-enriched uranium (LEU) fuel [1]. It was found that a significant number of research reactors, originally designed with technologies that required highly enriched uranium (HEU) fuel, can effectively serve their research purposes with decreasing uranium enrichment in the fuel. Consequently, since 1978, 71 reactors that were initially fueled with HEU have been successfully converted to LEU [2,3].

NASA is currently engaged in developing cutting-edge designs for LEU-fueled nuclear thermal propulsion systems, with the goal of employing them for propelling human missions to Mars [4].

Kazakhstan also supported the concept of minimizing the use of highly enriched uranium and has successfully completed the conversion of both the VVR-K (2016) and the IVG.1M research reactors (2023) to LEU fuel [5].

The IVG.1M research reactor, developed by specialists from N.A. Dollezhal Scientific Research and Design Institute of Energy Technologies (NIKIET) (https://www.nikiet.ru/, accessed on 26 August 2023), was constructed at the former Semipalatinsk nuclear test site near Kurchatov, Kazakhstan, between 1966 and 1969. It was originally designed for testing fuel assemblies and cores of high-temperature gas-cooled reactors, including prototype reactors for nuclear rocket engines. In 1990, the reactor underwent modifications that involved replacing the gas-cooled technological channels with water-cooled ones, resulting in the reorientation of the reactor towards a wider range of applications. Reactor irradiation experiments, such as the study of the radiation resistance of optical fibers [6] and the investigation of the spectral-luminescent characteristics of gases [7,8,9], research on the interaction between the structural materials of the international thermonuclear experimental reactor (ITER) and hydrogen in reactor irradiation conditions, etc., have become possible.

Since 2010, specialists from the National Nuclear Center, in collaboration with American and Russian partners, have conducted a series of studies to explore the feasibility of converting the IVG.1M reactor to LEU fuel as part of international non-proliferation initiatives. The main conversion criteria were to maintain the initial maximal thermal neutron flux density (3.5 × 10¹⁴ n/cm² × s⁻¹) and to avoid structural modifications.

The feasibility of the IVG.1M reactor’s conversion was confirmed through computational and analytical studies [10,11,12]. In a similar way, in [13], for the analysis of proposed LEU-fueled reactor cores, precise Monte Carlo computational codes were used.

Subsequently, successful experimental tests were conducted on the IVG.1M LEU fuel [14,15], followed by fuel manufacturing. The enrichment of ²³⁵U in the fuel was set at 19.75% (previously it was 90%), and the core of the fuel element was modified from a uranium–zirconium alloy to a zirconium matrix with inserted uranium filaments [16].

Argonne National Laboratory (ANL) supported the National Nuclear Center in the evaluation of the feasibility of converting the IVG.1M reactor to operate with LEU fuel, selecting a suitable LEU fuel design and developing a new version of the safety analysis report. ANL also assisted NNC in obtaining approval for the conversion procedure through the assessment of the national regulatory body.

During the conversion process, the need for upgrading and modernization became evident. As a result, the reactor underwent systematic modernizations, which included the replacement of coolant systems for the primary coolant of the reactor, the upgrading of the power supply system, and the modernization of the radiation monitoring system.

In 2022, a physical start-up of the reactor was conducted, involving the loading of the core with LEU fuel and a series of experimental studies to determine the neutron-physical characteristics of the reactor. These studies are necessary as actual reactor parameters can differ from the calculated and design values, and understanding these parameters is crucial for nuclear safety and the subsequent operation of the reactor. For example, in [17], the experimental uncertainties in the neutron-physical characteristics of several VVER-1000 units with similar initial fuel loading are estimated. For the CEFR fast reactor, the uncertainty of the reactivity measured in the start-up tests was investigated through reactivity measurements and McCARD Monte Carlo calculations [18].

This paper primarily focuses on the key experimental results of the physical start-up program, specifically the evaluation of the control and protection system reactivity worth and the determination of neutron-physical characteristics of the IVG.1M reactor with LEU-fuel.

Section 2 briefly describes the main structural characteristics of the IVG.1M research reactor. The features of achieving the first critical state of the reactor with LEU fuel are given in Section 3. Section 4 includes experimental and calculation results of the determination of the reactivity worth curve of the control drum system and the reactivity worth of individual control drums. The reactivity worth measurements of the reactivity compensation rod system are presented in Section 5. Additional experiments to determine the reactivity effects of removing the physical mock-ups of fuel assemblies and draining the technological channels are compared with preliminary calculation assessments and presented in Section 6. Section 7 and Section 8 contain the results of measurements of reaction rates in the central experimental channel and power distribution in LEU fuel using activation and fuel detectors. Measurements of gamma radiation fields and coolant activity during physical start-up are presented in Section 9 and Section 10, respectively. Section 11 contains a discussion of the results obtained from the complex of experiments. The conclusions are presented in Section 12.

2. The IVG.1M Research Reactor

The IVG.1M research reactor (Figure 1) is heterogeneous thermal neutron reactor using a light water moderator and coolant. The effective core diameter is 548 mm, and the core height is 800 mm. The fuel assemblies are located in 30 water-cooled technological channels placed in three ring rows. The reactor control and protection system (CPS) consists of 10 control drums situated in the lateral beryllium reflector. The control drums combine the functions of emergency protection, automatic and manual control.

The control drums are cylindrical structures made of beryllium, with absorbing rods mounted on the outer surface at an angle of 112°. The regulation is achieved by rotating the control drums using stepper motors. After taking one step, each control drum in the IVG.1M reactor rotates by 0.3 degrees and remains in that position until all the other control drums have also completed one step. This sequential movement ensures the synchronized regulation of the reactor. The total stroke of each drum is 6000 steps, covering a range of 0° to 180°. In the fully extracted position (0-step position), the absorption sector of the regulating drum is aligned with the core axis, facing the core [19].

3. First Critical State

The reactor with LEU fuel achieved the first critical state by turning the control drums from 1600 steps, corresponding to the position of activated emergency protection. The control drums were rotated at a rate of 10 steps per second (0.04 β_eff/s). Readings from impulse neutron power control channels (CCNP1, CCNP2) were recorded every 100 steps, and countdown curves were plotted (Figure 2).

The critical state of the reactor with a full load of LEU fuel was achieved on 5 May 2022, when the control drums were turned to 91° (3050 steps).

4. The Measurements of the Control Drum Reactivity Worth

The reactivity worth curve of the control drums was determined by implementing various modifications to the reactor core’s configuration, allowing for the achievement of different critical states. These modifications involved the insertion and removal of reactivity-compensating rods, the installation and unloading of physical mock-ups of fuel assemblies, and the utilization of additional absorbers.

For these studies, the initial core configuration comprised 29 standard technological channels and one measurement channel containing a physical mock-up of fuel assembly from the first row. During the experiments, the reactivity compensation rods were removed from the core.

The reactivity worth curve was determined by adjusting the control drums from the critical position to introduce positive or negative reactivity. For the IVG.1M reactor, the critical position refers to the position of the control drums when the reactor is in a critical state. The introduced reactivity was measured using a CVR-11 reactimeter, which is capable of measuring ionization chamber currents ranging from 10⁻¹¹ A to 10⁻³ A. The reactimeter provides reactivity measurements in the range of −25 β_eff to +0.7 β_eff with an error of ±5%.

The challenge was to measure the section of the reactivity worth curve that extended beyond the standard critical position (ranging from 3050 to 6000 steps). The standard critical position refers to the reactor being loaded with 30 regular technological channels and the reactivity compensation rods fully withdrawn from the core. To reduce the excessive reactivity margin, a steel neutron absorber was placed in the central experimental channel, and a second additional steel neutron absorber was installed instead of the physical mock-up of fuel assembly. With these modifications, the critical state was achieved when the control drum system reached 5720 steps. The reactivity margin was determined to be 0.035 β_eff.

The section of the control drums’ reactivity worth within the 5300–5950 step range was determined by introducing positive reactivity into the reactor. After measuring this section, the additional neutron absorber was unloaded. Subsequently, measurements of the control drums’ reactivity worth in the 3000–6000 step range were conducted by introducing negative reactivity through the synchronous rotation of the control drums by 300, 600, and 1000 steps from the critical state.

Measurements of the section from 0 to 3000 steps were made with the core configuration reverted to its initial state (29 technological channels and one physical mock-up in the measurement channel).

The reactivity worth of the control drum system was determined by performing drops of the control drums from various critical positions. The reactivity worth was defined as the sum of two values:

-

The reactor subcriticality for a given position of the control drum system, obtained by measuring the introduced negative reactivity when the drums were dropped from this position;

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The reactivity margin for the same position of the drums, determined from the known section of the reactivity worth curve.

Figure 3 shows the obtained reactivity worth curve of the control drum system in comparison with preliminary neutronic calculations performed with MCNP6 [20] using the ENDF/BVII.0 data library [21] and a full-scale model of the IVG.1M reactor with LEU fuel [22,23]. It should be noted that the calculated values were obtained under the condition of fully withdrawn reactivity compensation rods.

The operating position of the control drums at 3050 steps corresponds to a reactivity margin of 4.8 β_eff. The reactivity worth of the control drum system is 11.2 ± 0.3 β_eff. The calculated value of β_eff is 0.00719.

In addition to the experiments conducted to determine the reactivity worth of the control drum system, additional measurements were performed to determine the reactivity worth of individual control drums (No. 6–10). The reactivity worth of the control drum is understood as the change in the reactivity margin with a full turn of the control drum to the position 0° (the absorbing layer of the drums is facing the core). These measurements involved dropping the control drums from positions ranging from 5040 to 5390 steps, with the reactivity compensation rods fully removed from the reactor. The reactivity worth values of control drums No. 6, 7, 9, and 10 were 1.34 β_eff, 1.43 β_eff, 1.27 β_eff, and 1.07 β_eff, respectively.

5. Determination of the Reactivity Worth of the Reactivity Compensation Rod System

A stationary system of reactivity compensation rods is a system of beryllium displacers whose position in the core affects the ratio of absorption and the moderation of thermal neutrons in the central part of the reactor. Withdrawing the reactivity compensation rods from the reactor core creates a void that is filled with water. When the reactivity compensation rods are fully withdrawn, they correspond to a position of 0 mm, and at their maximum insertion position, the rods correspond to 800 mm.

The reactivity worth curve of the reactivity compensation rods was determined by incrementally inserting the rods into the core at 100 mm intervals. The experimental data points for the curve were obtained by comparison of the critical states using the reactivity worth curve of the control drums.

It is important to mention that during the investigation of the reactivity worth curve of the reactivity compensation rods, a neutron absorber was installed in the central experimental channel. The reactivity worth of the reactivity compensation rods for this specific core configuration was determined to be 3.2 β_eff. However, after the removal of the neutron absorber, the reactivity worth was reevaluated by comparing the critical positions of the control drums corresponding to the extreme positions of the rods in the core. The revised reactivity worth of the reactivity compensation rods was found to be 3.5 ± 0.1 β_eff.

To validate the reactivity worth curve of the reactivity compensation rods in a core configuration without a neutron absorber, a criticality verification was performed with the reactivity compensation rods set at a position of 470 mm.

The reactivity worth curve obtained for a reactivity compensation rod system is presented in Figure 4, along with the results of the preliminary neutron-physical simulations. In the reactor model, the control drums were rotated to the 180° position, with the reflection layer facing the core axis.

6. Statistical Weights of Technological Channels and the Reactivity Effects of Their Drainage

During the physical start-up, experiments were conducted to determine the power distribution in the fuel assemblies of the first, second, and third rows of technological channels using their physical mock-ups. These experiments are described in more detail in [24]. During this stage, the core loading consisted of 29 regular technological channels and a measurement channel with a physical mock-up of the fuel assembly from the investigated row.

The statistical weight of the physical mock-up was determined from the reactivity worth curve of the control drums by comparing the reactivity margin with and without the physical mock-up loaded in the core.

Similarly, the reactivity effect of draining the first, second, and third row measuring channels was determined. The critical states of the reactor before and after modifying the core were compared using the reactivity worth curve of the control drum system.

The results of determining these effects for the first, second, and third row channels are presented in

Table 1. Reactivity effects.

Change in the Core	Reactivity Effect, βeff
	Experiment	Calculation
Removing the physical mock-up from the first row channel	−3.2 ± 0.1	−2.91 ± 0.02
Removing the physical mock-up from the second row channel	−3.0 ± 0.2	−2.92 ± 0.02
Removing the physical mock-up from the third row channel	−2.2 ± 0.2	−1.86 ± 0.02
Draining water from the first row channel	−0.41 ± 0.05	−0.40 ± 0.02
Draining water from the second row channel	−0.43 ± 0.05	−0.48 ± 0.02
Draining water from the third row channel	−0.42 ± 0.05	−0.49 ± 0.02

.

7. Measurement of the Activation Reaction Rate in the Central Experimental Channel

To determine the activation reaction rate, ⁵⁸Ni and ¹⁹⁷Au activation indicators were placed in the central experimental channel. The indicators were irradiated during a 1 kW start-up with a duration of 1032 s (effective irradiation time). During these experiments, the core consisted of 30 regular technological channels, and the reactivity compensation rods were positioned at 0 mm.

Gamma spectrometric measurements were performed using a scintillation gamma detector and InSpector-2000 analyzer. The mass of the indicators was preliminarily measured on analytical scales (VLR-200). The results of determining the reaction rate are given in

Table 2. Results of activation reaction rates, react/(s$\cdot$nucl.).

Reaction	LEU Fuel Core (Reactor Power of 1 kW)		HEU Fuel Core (Reactor Power of 10 kW)
	Calculated	Measured	Measured
58Ni(n,p)58Co	2.04 × 10−16	1.93 × 10−16	1.84 × 10−15
197Au(n,γ)198Au	2.00 × 10−12	2.12 × 10−12	2.11 × 10−11

in comparison with calculated data and similar studies carried out on the core with HEU fuel. The error in the determination of reaction rates in activation detectors is estimated to be 8%.

The reaction rate serves as an input value for subsequent calculations of the neutron field characteristics.

8. Measurement of Power Distribution in the LEU Fuel of the IVG.1M Reactor

The power distribution over the volume of fuel assembly of water-cooled technological channels (WCTCs) was studied using the activation method and physical mock-ups of fuel assemblies. The physical mock-ups were designed to imitate regular fuel assemblies, and allowed for the extraction and disassembly of the mock-up and the installation of indicators in the form of fuel rods and activation foils or wires.

The power distribution within one row of WCTCs can be considered almost uniform when the core is fully loaded with thirty WCTCs, provided there are no disturbances in the location of the WCTCs. Therefore, during physical studies, it is sufficient to investigate the energy release in one channel from each of the three WCTC rows.

Each physical mock-up of the WCTC fuel assembly was equipped with:

-

Fuel fragments calibrated for uranium-235 content (100 mm high) to study the relative power distribution over the fuel assembly cross section;

-

Copper wire located along the length of the physical mock-up to determine the relative power distribution throughout the height of the fuel assembly.

Based on the results of gamma-spectrometric measurements of fuel and activation indicators, the power distributions over the height and radius of the fuel assemblies were obtained. A detailed description of these results is provided in [24].

The main parameters characterizing the non-uniformity of energy release are the power peaking factors. They are defined as the ratio of the maximum value of thermal neutron flux density (or intensity of gamma radiation) to the average value at the specific coordinate (height, radius, etc.). The results of determining the power peaking factors are presented in

Table 3. Power peaking factors.

	Power Peaking Factor, rel. u.
	First WCTC Row	Second WCTC Row	Third WCTC Row
along height of fuel assembly	1.58	1.58	1.27
along radius of fuel assembly	1.52	1.55	1.37

.

9. Measurement of Gamma Radiation Fields during the Physical Start-Up of the Reactor

Radiation studies were carried out to measure various parameters during reactor operation, including the dose characteristics of external radiation fields during reactor operation and the level of coolant activation (Section 11).

The research involved 15 reactor start-ups with power levels ranging from 2 W to 10 kW. To ensure radiation safety, several precautions were taken:

-

Removal of personnel from specific areas during the reactor operation;

-

The availability of stationary and mobile radiation monitoring systems;

-

Relatively low levels of the reactor operating power during experiments to determine the reactivity worth;

-

The use of personal protective equipment for personnel involved in preparatory work and physical research.

Gamma radiation fields were measured using Thermo Fisher Scientific automated radiation monitoring system in different locations, and the results of measurements are presented in

Table 4. Results of measuring the radiation fields at a reactor power of 10 kW.

Check Point	Distance from the Reactor, m	Dose Rate, µSv/h
Sub-reactor room	5	0.13–0.21
Above the reactor	5	0.28–20.4
Adjoining room	10	0.12–0.19
Ceiling	35	0.10–0.20

.

10. Study of Coolant Activity

The objective of these tests was to determine the specific activity of the coolant (water) poured into the measuring channel, and to calculate the relative yield of the ²³⁵U fission products (FP) into the water in the measuring channel. The tests were carried out using the GC1518 semiconductor gamma detector and InSpector 2000 gamma spectrometer.

Water from the measuring channel was sampled using a syringe with an extension tube, and the volume of the water sample in the measuring tank was 0.5 L. The water sample was measured four times. The exposure time before the first and second measurements was approximately 20 min, and for the third and fourth measurements, it was around 1 h. The acquisition time for the first two measurements was 200 s, and for the second two measurements, it was 400 s.

Table 5. Results of the coolant activity measurements.

Isotope-AP	A(t), Bq/L		A0(t), Bq/L		A0(t), Bq/L
	t1 = 30 min	t2 = 39 min	t1 = 30 min	t2 = 39 min
Na-24	9.9 × 102	1.0 × 10³	1.0 × 10³	1.1 × 10³	1.0 × 10³
Isotope-FP	A(t), Bq/L		A0(t), Bq/L		R/B
	t1 = 30 min	t2 = 39 min	t1 = 30 min	t2 = 39 min
Kr-85	5.6 × 102	4.8 × 102	6.1 × 102	5.4 × 102	5.1 × 10−6
Kr-87	2.7 × 10³	2.6 × 10³	3.5 × 10³	3.8 × 10³	5.0 × 10−6
Kr-88	2.5 × 10³	2.3 × 10³	2.8 × 10³	2.6 × 10³	5.7 × 10−6
Rb-89	1.2 × 104	7.9 × 10³	4.8 × 10³	4.7 × 104	6.3 × 10−6
Sr-92	3.8 × 10³	4.2 × 10³	4.3 × 10³	5.0 × 10³	5.4 × 10−6
Y-94	2.7 × 10³	2.0 × 10³	8.2 × 10³	8.7 × 10³	1.1 × 10−6
Tc-104	2.2 × 10³	2.0 × 10³	6.7 × 10³	8.8 × 10³	3.3 × 10−6
Te-131	3.5 × 10³	3.4 × 10³	8.2 × 10³	1.0 × 104	3.4 × 10−6
I-133	8.8 × 102	6.1 × 102	9.0 × 102	6.2 × 102	6.0 × 10−6
I-134	7.7 × 10³	7.4 × 10³	1.1 × 104	1.2 × 104	4.0 × 10−6
Te-134	5.3 × 10³	4.7 × 10³	8.7 × 10³	9.1 × 10³	2.3 × 10−6
I-135	1.6 × 10³	2.0 × 10³	1.7 × 10³	2.1 × 10³	5.0 × 10−6
Cs-138	1.4 × 104	1.3 × 104	2.8 × 104	3.0 × 104	5.9 × 10−6
Xe-138	7.5 × 10³	4.3 × 10³	3.3 × 104	3.0 × 104	3.0 × 10−6
La-142	3.9 × 10³	2.9 × 10³	4.9 × 10³	3.9 × 10³	2.9 × 10−6
Average					4.3 × 10−6

presents the results of determining the specific activity of the water sample at the start of the measurement A(t), and at the time of the reactor shutdown A₀, as well as the average value of the coefficient R/B [25] based on the results of measurements.

11. Discussion of Result

During the experiments, the reactivity worth of the control and compensating elements of the CPS were determined for the IVG.1M reactor after conversion. When the reactor was loaded with 30 regular technological channels containing LEU fuel, the reactivity margin was found to be 4.8 β_eff. In comparison, for the reactor with fresh HEU fuel, the reactivity margin was 3.2 β_eff. The resulting margin is considered sufficient for conducting experiments with ampoule devices with a reactivity worth of 1–2 β_eff, and therefore, there is no need to introduce additional positive reactivity up to 3.5 β_eff by inserting reactivity compensation rods.

After the conversion, the reactivity worth of the control drum system was measured to be 11.2 β_eff, whereas before the conversion it was 10.6 β_eff. This indicates that the reactivity worth curve of the control drums and the fundamental requirements for reactor control remain largely unchanged. These requirements include a single rotation of the control drum system during power build-up and when the critical state is achieved.

A deviation of up to 20% was observed in the measured reactivity effects during the removal of the physical mock-up of the fuel assembly and the drainage of water from the channel compared to their predicted values. These deviations are likely attributable to the specificities of the experimental procedure used to determine these effects.

Studies conducted with a gold activation indicator demonstrated that the thermal neutron production parameters in the IVG.1M reactor did not deteriorate after the conversion.

The energy release fields in the physical mock-ups of fuel assemblies were investigated, and measurements of gamma radiation fields were conducted during the physical start-up of the reactor, as well as the study of coolant activity. During the start-ups, in all rooms except the reactor rooms, the control level of the equivalent dose rate of external radiation did not exceed 12 µSv/h. The total values of the specific activity A₀ of the registered fission and activation products in the water sampled from the measuring channel for the reactor start-ups during the study of physical mock-ups of fuel assemblies ranged from 6×10⁵ Bq/L to 8×10⁵ Bq/L.

The conducted studies have confirmed that the neutron-physical characteristics of the reactor allow for the realization of power start-up modes.

Further research on LEU fuel for the IVG.1M reactor will focus on calculating changes in isotopic composition during the reactor campaign [23] and estimating fuel burnup [13,26]. At present, there have been a few advancements in the field of the safe storage of spent nuclear fuel. Notably, calculations have been conducted for the neutron components of both fresh and irradiated fuel from the IVG.1M reactor [27].

12. Conclusions

In 2022, the first stage of commissioning the IVG.1M reactor with LEU fuel was successfully completed, which included the physical start-up of the reactor. During the physical start-up, approximately 40 critical states were investigated, and the conducted physical studies demonstrated that the operating characteristics of the LEU core remained largely unchanged.

The reactivity worth curves of the CPS operating and compensating components, as well as the statistical weights of the technological channels and the reactivity effects of their drainage, were determined. The results obtained for the reactivity effects are in satisfactory agreement with the preliminary results of neutron-physical modeling.

An increase in the reactor’s reactivity margin has been confirmed, with the maximum reactivity margin reaching approximately 8.3 β_eff. This increase extends the operating time of LEU fuel and the range of applications that can be carried out on the IVG.1M reactor.

The operability of the IVG.1M reactor’s LEU core was confirmed through a series of start-ups at nominal power levels as part of the power start-up program. As a result, the reactor successfully completed the final check of operability for all systems under normal conditions.