KAZAKHSTAN’S NUCLEAR GOVERNANCE AS A FOREIGN POLICY ASSET

Kazakhstan’s nonproliferation initiatives are recognized worldwide. Kazakhstan is a party to almost all major nuclear treaties, a key driver in the creation of a Central Asian nuclear-weapon-free zone, initiated the Universal Declaration on Building a World Free of Nuclear Weapons, established a low-enriched uranium bank under the auspices of the IAEA in Ust-Kamenogorsk to be used for peaceful purposes in the event of a disruption in the supply of fuel for nuclear power plants.While the foreign policy acts of Kazakhstan in non-proliferation are well known and internationally appreciated, the transfer of Kazakh experience in the governance of the nuclear sector is lesser known asset. For example, the experience of the Committee for Atomic and Energy Supervision and Control and KAZATOMPROM in uranium mining and transport was shared with countries from the Southern African Development Community, under an EU project, implemented by the International Science and Technology Center. This example reveals the great potential this themes have for the further input of Kazakhstan in international development cooperation.

Study of migration of radioactive elements in clay layers during the burial of radioactive waste

In 2015, Kazakhstan and the International Atomic Energy Agency (IAEA) signed an agreement to host a low-enriched uranium bank in Ust-Kamenogorsk. In 2019, several batches of enriched uranium were delivered to Kazakhstan and the bank began operations at the Ulba Metallurgical Plant. When transporting and disposing of radioactive elements, there is a need to reduce this possibility by limiting the transfer of uranium from underground storage to underground water. Therefore, in this article, a study was conducted on the migration of radioactive elements in clay layers during the disposal of radioactive uranium waste. There are now many underground repositories (for some types of radioactive waste). These systems are based on different underground container structures for different geological formations. For underground repositories located in geological environments where enriched uranium can migrate, other system components must reduce this possibility by preventing or limiting uranium mobility. This work investigates the process of convective transport of radioactive elements, in a moist soil layer through the installation of an additional natural clay barrier layer, the migration of radioactive elements during safe disposal, the effect of diffusion and convection through the solid waste layer.

Radioactive Fission Waste from Molybdenum-99 Production and Proliferation Risks

Molybdenum-99 (99Mo) is a parent radioisotope of Technetium-99m (99mTc) widely used in nuclear diagnostics. The production of this radioisotope by PT. INUKI generated radioactive fission waste (RFW) that theoretically contains239Pu and235U, posing a nuclear proliferation risk. This paper discusses the determination of radionuclides inventory in the RFW and the proposed strategy for its management. The radionuclides inventory in the RFW was calculated using ORIGEN 2.1 code. The input parameters were obtained from one batch of 99Mo production using high enriched uranium in PT. INUKI. The result showed that the RFW contained activation products, actinides, and fission products, including239Pu and235U. This result was then used for consideration of the management of the RFW. The concentration of 235U was reduced by a down-blending method. The proposed strategy to further manage the down-blended RFW was converting it to U3O8 solid form, placed in a canister, and eventually stored in the interim storage for high-level waste located in The Radioactive Waste Technology Center.

Investigative study of the radiation damage on fuel clad of miniature neutron source reactor using computational tools

Core conversion requires some evaluation of the reactor safety. Changes to the reactivity worth, shutdown margin, power density and material properties are crucial to the proper functioning of the reactor. The focus of this article is to study the neutron flux distribution in the reactor core and radiation damage on candidate clads. The Ghana Research Reactor-1 (GHARR-1) operates at maximum power of 30 kW in order to attain a flux of 1.0× 1012 n·cm–2·s for the high enriched uranium core. Using the GHARR-1 core geometry, considering 348 fuel pins, the multiplication factor (Keff) is calculated at enrichments of 10%, 12.5%, 16%, 20%, 30% and 90.2%. The spectrum of neutron flux generated in the 26 group is also calculated at the specified enrichments. The ion/particle interactions with the targets (clad) were studied in the Stopping and Range of Ion in Matter code to establish the best clad material based on recorded defects and vacancies generated. From the calculations and simulations, the best choice from the candidate clads based on the assessment is SiC. The calculation of the fuel campaign length gives 7.5 years. The defects sustained by the prospective clad showed low susceptibility to swelling and other forms of deformation.

Weight Gain and Hydrogen Absorption in Supercritical Water At 500°C of Chromium-Coated Zirconium-Based Alloys: Transverse vs Longitudinal Direction

Canadian Nuclear Laboratories (CNL) has an on-going Research & Development programme to support the development of a scaled-down 300 MWe version of the Canadian Super-Critical Water Reactor (SCWR) concept. The 300 MWe and 170-channel reactor core concept uses low enriched uranium fuel and features a maximum cladding temperature of 500°C. Our goal is to test surfacemodified zirconium alloys for use as fuel cladding. Zirconium alloys are attractive as they offer low neutron cross section thereby allowing the use of low enriched fuel. In this paper, we report on the results of general corrosion experiments used to evaluate chromiumcoated zirconium-based alloys in the two chemistries (630 ug/kg O2 in both deaerated and lithiated supercritical water). These experiments were conducted in a refreshed autoclave at 500°C and 23.5 MPa. After exposure, the weight gain and the hydrogen absorption were examined. At adequate coating thickness, longitudinal and transverse coupons show similar corrosion behaviour with improved corrosion resistance compared to uncoated coupons. The measured concentrations of hydrogen absorption are higher for the transverse coupons. Alkaline treatment resulted in higher weight gains than was found in pure oxygenated supercritical water.

Conceptual design of critical assembly using Low enriched uranium fuel and moderated light wate

Critical assembly is a very important facility to serve for fundamental reactor physics research, application of neutron source, training and education. In nuclear engineering, critical assembly is a facility for carrying out measurement of reactor physics parameters, creating benchmark problem, validation of neutron physics calculation tool in computer codes and nuclear data. Basing on concept using commercial Nuclear Power Plant (NPP) fuels such as PWR (AP-1000) and VVER-1000 fuel rods with limited 2 meter in length and fully controlled by water level, the conceptual design of the critical assembly has been carried out in neutronic, thermal hydraulics and safety analysis. Ten benchmark critical core configurations of critical assembly are established and investigated to show safety during normal opeartion and accident conditions. Design calculation results show that NPP fuels are fully adequate for critical assembly operating under nominal power 100W and having average neutron flux about 3×109 neutron/cm2.s.

On the determination of neutron multiplication by the Rossi-alpha method

Introduction. This work contains the results of determining the prompt neutron multiplication factor in the subcritical state of a one-core BFS facility, obtained by the neutron coincidence method, for which the influence of the error in the βeff in determining the multiplication factor turned out to be insignificant. The core of the facility consisted of rods filled with pellets of metallic depleted uranium, 37% enriched uranium dioxide and 95% enriched plutonium, sodium, stainless steel and Al2O3. Stainless steel served as a reflector. Methods. In contrast to the inverse kinetics equation solving (IKES) method, which is convenient for determining reactor subcritical states, the neutron coincidence method practically does not depend on the error in the value of the effective fraction of delayed neutrons βeff. If in the IKES method the reactivity value is obtained in fractions of βeff, i.e., from the measurement of delayed neutrons, the neutron coincidence method is based on the direct measurement of the value (1 – kσp)2, where is the effective multiplication factor by prompt neutrons. The total multiplication factor is defined as keff = kσp + βeff. If, for example, keff ≈ 0.9 (which is typical for determining the fuel burnup campaign), then it is the error in determining kσp that is the main one in comparison with the error in βeff. Thus, a 10% error in βeff of 0.003–0.004 (typical for plutonium breeders) will make a contribution to the error 1 – keff equal to 1 – kσp + βeff ≈ 0.00035, i.e., approximately 0.35%, but not 10%, as in the IKES method. Rossi-alpha measurements were carried out using two 3He counters and a time analyzer. The measurement channel width Δt was 1.0 μs. From these measurements, the value of the prompt neutron multiplication factor was obtained. In this case, the space-isotope correlation factor for the medium with a source was calculated using the following values: Φ(x) – solutions of the inhomogeneous equation for the neutron flux and Φ+(x) – solutions of the ajoint inhomogeneous equation. Results. The authors also present a comparison of the results of the Rossi-alpha experiment and measurements of the BFS-73 subcritical facility by the standard IKES method in determining the multiplication factor value. The data of the IKES method differ insignificantly from the results of the Rossi-alpha method over the entire range of changes in the subcriticality with an increase in the subcriticality of the BFS-73 one-core facility. Conclusion. It was impossible to apply the neutron coincidence method to fast reactors; however, the method turned out to be quite workable on their models created at the BFS facility, which was successfully demonstrated in this study.

Study on transmutation efficiency of the VVER-1000 fuel assembly with different minor actinide compositions

The feasibility of transmutation of minor actinides recycled from the spent nuclear fuel in the VVER-1000 LEU (low enriched uranium) fuel assembly as burnable poison was examined in our previous study. However, only the minor actinide vector of the VVER-440 spent fuel was considered. In this paper, various vectors of minor actinides recycled from the spent fuel of VVER-440, PWR-1000, and VVER-1000 reactors were therefore employed in the analysis in order to investigate the minor actinide transmutation efficiency of the VVER-1000 fuel assembly with different minor actinide compositions. The comparative analysis was conducted for the two models of minor actinide loading in the LEU fuel assembly: homogeneous mixing in the UGD (Uranium-Gadolinium) pins and coating a thin layer to the UGD pins. The parameters to be analysed and compared include the reactivity of the LEU fuel assembly versus burnup and the transmutation of minor actinide nuclides when loading different minor actinide vectors into the LEU fuel assembly.