Experimental Investigation of Iodine Removal in a Submerged Venturi Scrubber

Severe nuclear accidents can cause over-pressurization and serious damage to the containment of a nuclear power plant, which can result in the release of radioactivity into the environment. Filtered containment venting systems are a nuclear safety system that is designed to control over-pressurization and prevent radioactive fission products from spreading into the environment in the case of a severe accident. Iodine is one of the most harmful products among this list of fissionable products, as it can cause thyroid cancer. The removal of iodine is very important in order to ensure the safety of people and the environment. Thus, an indigenous lab scale setup of this system was developed at PIEAS to conduct research on iodine removal. It is comprised of a compressor for replicating high-pressure accident scenarios, a heater to keep iodine in a vapor form, a dosing pump for the injection of iodine, and a venturi scrubber, submerged in the scrubbing column, containing a solution of 0.2% sodium thiosulphate and 0.5% sodium hydroxide. Inlet and outlet samples were trapped in 0.1 M KOH solution and analyzed via UV-VIS spectroscopy. Operating parameters play an important role in the working of a venturi scrubber. The throat velocity was varied to determine its influence on the removal efficiency of iodine. An increase in removal efficiency was observed with an increase in throat velocity. A removal efficiency of >99% was achieved, which fulfilled the requirements for FCVS.

Features of methods for monitoring the fuel cladding tightness in lead-cooled fast breeder reactors

To timely detect failed fuel elements, a reactor plant should be equipped with a fuel cladding tightness monitoring system (FCTMS). In reactors using a heavy liquid-metal coolant (HLMC), the most efficient way to monitor the fuel cladding tightness is by detecting gaseous fission products (GFP). The article describes the basic principles of constructing a FCTMS in liquid-metal-cooled reactors based on the detection of fission products and delayed neutrons. It is noted that in a reactor plant using a HLMC the fuel cladding tightness is the most efficiently monitored by detecting GFPs. The authors analyze various aspects of the behavior of fission products in a liquid-metal-cooled reactor, such as the movement of GFPs in dissolved and bubble form along the circuit, the sorption of volatile FPs in the lead coolant (LC) and on the surfaces of structural elements, degassing of the GFPs dissolved in the LC, and filtration of cover gas from aerosol particles of different nature. In addition, a general description is given of the conditions for the transfer of GFPs in a LC environment of the reactor being developed. Finally, a mathematical model is presented that makes it possible to determine the calculated activity of reference radionuclides in each reactor unit at any time after the fuel element tightness failure. Based on this model, methods for monitoring the fuel cladding tightness by the gas activity in the gas volumes of the reactor plant will be proposed.

Severe accident simulation for VVER-1000 reactor using ASTEC-V2.1.1.3

The station black-out (SBO) is one of the main accident sequences to be considered in the field of severe accident research. To evaluate a nuclear power plant’s behavior in the context of this accident, the integral ASTEC-V2.1.1.3 code “Accident Source Term Evaluation Code” covers sequences of SBO accidents that may lead to a severe accident. The aim of this work is to discuss the modelling principles for the core melting and in-vessel melt relocation phenomena of the VVER-1000 reactor. The scenario of SBO is simulated by ASTEC code using its basic modules. Then, the simulation is performed again by the same code after adding and activating the modules; ISODOP, DOSE, CORIUM, and RCSMESH to simulate the ex-vessel melt. The results of the two simulations are compared. As a result of SBO, the active safety systems are not available and have not been able to perform their safety functions that maintain the safety requirements to ensure a secure operation of the nuclear power plant. As a result, the safety requirements will be violated causing the core to heat-up. Moreover potential core degradation will occur. The present study focuses on the reactor pressure vessel failure and relocation of corium into the containment. It also discusses the transfer of Fission Products (FPs) from the reactor to the containment, the time for core heat-up, hydrogen production and the amount of corium at the lower plenum reactor pressure vessel is determined.

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.

Concept of a fast breeder reactor to transmute MAs and LLFPs

The long-term issues of nuclear power systems are the effective use of uranium resources and the reduction of radioactive waste. Important radioactive wastes are minor actinides (MAs: 237Np, 241Am, 243Am, etc.) and long-lived fission products (LLFPs: 129I, 99Tc, 79Se, etc.). The purpose of this study was to show a concept that can simultaneously achieve the breeding of fissile materials and the transmutation of MAs and LLFPs in one fast reactor. Transmutation was carried out by loading innovative Duplex-type MA fuel in the core region and LLFP-containing moderator in the first layer of the radial blanket. Breeding was achieved in the core and axial blanket. As a result, it was clarified that in this fast breeder reactor, a breeding ratio of approximately 1.1 was obtained, and MAs and LLFPs achieved a support ratio of 1 or more. The transmutation rate was 10.3%/y for 237Np, 14.1%/y for 241Am, 9.9%/y for 243Am, 1.6%/y for 129I, 0.75%/y for 99Tc, and 4%/y for 79Se. By simultaneously breeding fissile materials and transmuting MAs and LLFPs in one fast reactor, it will be possible to solve the long-term issues of the nuclear power reactor system, such as securing nuclear fuel resources and reducing radioactive waste.

Actinides induced irradiation damage and swelling effect in irradiated Zircaloy-4 after 30 years of storage

After several years in the reactor core, irradiated nuclear fuel is handled and subsequently stored for a few years under water next to the core, to achieve thermal cooling and decay of very short-lived radionuclides. Thereafter, it might be sent to dry-cask interim storage before final disposal in a deep geological repository. Here, the spent nuclear fuel (SNF) is subject to a series of physicochemical phenomena which are of concern for the integrity of the nuclear fuel cladding. After moving the SNF from wet to dry storage, the temperature increases, then slowly decreases, leading the hydrogen in solid solution in the cladding to precipitate radially with consequent hydride growth and cladding embrittlement (Kim, 2020). Another phenomenon affecting the physical properties of the cladding during interim dry storage is the irradiation damage produced in the inner surface of the cladding by the alpha decay of the actinides present at the periphery of the pellet, particularly when the burnup at discharge is high. SNF pellets with high average burnup present larger fuel volumes at the end of their useful life due to accumulation of insoluble solid fission products and noble gases, which leads to disappearance of the as-fabricated pellet–clad gap. Further swelling is expected as a consequence of actinide decay and the accumulation of helium. This leads to larger cladding hoop stress and larger alpha decay damage. The present work first investigates the variation in diameter caused by pellet swelling in an irradiated Zircaloy-4 cladding after chemical digestion of the uranium oxide (UOx) pellet. Second, the irradiation damage produced during the 30 years elapsed since the end of irradiation in terms of displacements per atom (dpa) is studied by means of the FLUKA Monte Carlo code. The irradiation damage produced by the decay of actinides in the inner surface of the cladding extends for less than 3 % in depth. The considered cladded UOx pellet was extracted from a pressurized water reactor (PWR) fuel rod consisting of five segments, with an average burnup at discharge of 50.4 GWd (tHM)−1.

Assessment of Partitioning and Transmutation of High-Level Waste and Hypothetical Implementation Scenarios in Germany

In the context of the search for a deep geological repository for high-level radioactive waste from nuclear energy a preliminary waste treatment is repeatedly called into play by partitioning and transmutation (P&T). Proponents of this approach promise that with P&T, the requirements for and the risks posed by a – then still necessary – repository could be significantly reduced. However, such technological promises have to be prospectively, promptly and publicly reasonably verifiable. Partitioning is reprocessing in which, in addition to separating uranium and plutonium from the fission products, other material streams (for example, the minor actinides) are extracted. In transmutation, radionuclides – especially through nuclear fission – are converted into other nuclides. Thus, conversion of the parent nuclides into nuclides with shorter half-lives, lower radiotoxicity, or into stable nuclides could be achieved. For the assessment of P&T, essential aspects are the current degree of maturity of necessary technologies, the requirements for research and development, technological development risks, the basic feasibility and objective, risks of a hypothetical operation of corresponding plants and the possible effects on nuclear waste disposal. More specifically, on the technological side, it is all about development periods, technical security requirements and licensability, proliferation risks and implementation periods. The presentation of the results of some hypothetical P&T scenarios is intended to help to assess the impacts on radioactive waste present in Germany, necessary facilities and operating periods. Thus, pyro-chemical and hydrochemical separation processes, special transuranic fuels based on mixed oxides (MOX) or uranium-free fuel types and critical fast reactors, subcritical (accelerator-driven) reactors, as well as molten salt reactors, are considered. One difficulty is that the multiple recycling of the transuranics changes the fuel composition. Detailed statements about these changes are only possible with complex simulation calculations and their influence on safe reactor operation. So far, this has not happened on an international scale. In the modelling presented here, an attempt was made to represent the restrictions that the reactor design has on the fuel composition more precisely, at least insofar as the element composition of the fuel remains the same for the duration of the scenario. Conclusions presented from the analysis of the hypothetical scenarios affect, among other things, necessary operating periods and the number of plants and changes achieved in the stock of both transuranics and fission products.