Long-Lived-Radionuclide Transmutation Scenarios

For nuclear power stations, the main function of Essential Service Water System (ESWS) is to discharge the waste heat from reactor core and spent fuel pool to the environment controllably, which is directly related to the safety and economy of nuclear power stations. Usually ESWS use open water from sea, rivers, lakes, reservoirs, as heat transfer medium. Extremely harsh environmental conditions may disable system functions and even lead to ESWS failure, directly reduce the safety and economy of nuclear power stations, and cause serious nuclear accidents. Failure of ESWS is one of the main reasons that lead to the Fukushima nuclear accident because of the loss of electricity after the earthquake and tsunami. Based on the typical ESWS configuration and conditions of serving nuclear power stations in China, the influence of environmental conditions on the function of water system is studied, and the corresponding measures are analyzed. These conditions can be divided into three categories: temperatures, water levels, and physical and chemical characteristics. Temperatures affect cooling characteristic of ESWS mainly. Nuclear power stations in tropical areas need to focus on cooling capacity might be reduced by high temperature. Those in cold region need attention to excessive cooling and freezing problems caused by low temperature. The influence of water levels is mainly fluid transport capacity and selection of equipment to ESWS. When the range of natural water level is too wide, designers shall consider measures to narrow it, such as the construction of highly reliable reservoir. Inland nuclear power stations shall try to ensure the reliability of ESWS; prevent water level changes beyond the scope of design caused by drought and flood disasters. The effects of physical and chemical properties are derived from the open water characteristics, including high salinity, high chloride ion concentration, carrying solid particles, suspended solids, and aquatics, and so on. These characteristics will cause the equipment and pipeline eroded or even damaged, aqueducts of intake and output jammed, heat exchangers of the final heat sink weakened and other negative effects, resulting in ESWS performance decline. Some of these factors are the characteristics of station site natural environment, some others are changes caused by human activities. Some factors are sustained, long-term; some others may be sudden, temporary. Influence on these factors need to be taken measures from many aspects, including structure, biological disinfection, special materials and equipment, environmental protection measures around the nuclear power station, and so on. On the whole, the environmental factors that affect ESWS in the nuclear power stations are wide, and the influence mechanism is more complex. These factors ultimately act on ESWS, but most of them cannot be banished inside of ESWS or the final heat sink system. Against the negative effects from environmental conditions, it has to be considered from all steps in the engineering of nuclear power stations, including design, construction and operation. All the measures shall be suitable to local conditions, in order to ensure the safety and economy of nuclear power stations.

Objective: The purpose of this study was to investigate the accumulation of natural and artificial radioisotopes in medicinal plant raw materials, specifically examining greater plantain (Plantago major L.) leaves harvested from various territories of the Voronezh region, which differ in anthropogenic impact. Methods: The specific activity of long-lived artificial radionuclides (cesium-137, strontium-90) and natural radionuclides (thorium-232, potassium-40, radium-226) was measured in soil and greater plantain leaves using an MKGB-01 RADEK spectrometer-radiometer. Samples were collected from diverse areas with varying levels of industrial, agricultural, and urban influences. Correlation analysis and accumulation coefficients were used to assess the transfer of radionuclides from soil to plant. Results and Discussion: The results indicated that all collected greater plantain samples complied with radiation safety standards. Cesium-137 and strontium-90 were present in higher concentrations than natural radionuclides, but all levels were within permissible limits. A strong correlation between the radionuclide content in soil and plant leaves was observed, confirming that the contamination primarily occurs through soil. Accumulation coefficients for cesium-137 ranged from 1.81 to 3.29, with an average of 2.55. The study also identified trends of decreased accumulation with increased radionuclide concentrations in the soil, suggesting physiological mechanisms regulating radionuclide uptake. Conclusions: Greater plantain leaves from the Voronezh region accumulate radionuclides primarily through soil contamination. The study confirms that the leaves meet radiation safety standards for medicinal plant raw materials. Accumulation of cesium-137 was notably high, while other radionuclides showed lower levels. The results highlight the importance of monitoring radionuclide contamination in medicinal plants, particularly in urbanized and industrialized areas.

Based on the German decision to minimise transport of spent fuel casks between nuclear power plants, reprocessing plants and central storage facilities several on-site storage facilities were licensed until the end of 2003. Because of the large amount of Type B(U) transport casks which are going to be used for long-term interim storage the question of timelimited Type B(U) licence maintenance during the storage period of up to 40 years has been discussed under different aspects. This paper describes present technical aspects of the discussion. A main aspect of qualification of transport casks for interim storage is the long-term behaviour of the metallic seal–lid system. Here results are presented from current long-term experimental tests with metallic 'Helicoflex' seals in which pool water is enclosed. This series of tests has been performed by the Federal Institute for Materials Research and Testing (BAM) on behalf of the Federal Office for Radiation Protection (BfS) since 2001. Finally, the paper presents a German concept for an exchange of experience, know-how and state-of-the-art between authorities and technical experts with regard to cask dispatch in nuclear facilities. BAM has taken over a central role in this so-called 'coordinating institution for cask dispatching information' ('KOBAF') which entails management of an online database of cask-specific documents and a technical working group meeting twice a year. The goal is to keep comparable technical standards for all nuclear sites and storage facilities which are going to load and dispatch casks of the same or similar types under the responsibility of different German state governments for the coming decades.

Spontaneous stabilization of HTGRs without reactor scram and core cooling—Safety demonstration tests using the HTTR: Loss of reactivity control and core cooling

13 - Nuclear fuel fabrication

Indonesia is experiencing an increasing demand for electrical energy, which can be met through alternative sources such as nuclear energy generated in nuclear reactors at Nuclear Power Plants (NPPs). Small Modular Reactors (SMRs) can be implemented in Pressurized Water Reactors (PWRs). The fuel can be arranged in hexagonal or square pin configurations in such reactors, each potentially influencing the reactor’s neutronic performance. This research aims to determine the multiplication factor (keff) and excess reactivity, as well as the characteristics of neutron flux, fission rate, and minor actinide production resulting from using hexagonal and square fuel pin arrangements. This research was conducted by performing a neutronic analysis on the 300 MWth PWR using UN-PuN fuel with a comparison between the two fuel pin configurations. Neutronic calculations were carried out using OpenMC code based on the Monte Carlo method with the ENDF/B-VIII.0 library. The results indicate that different fuel pin arrangements yield distinct neutronic characteristics. The hexagonal fuel pin arrangement results in lower keff and excess reactivity values while exhibiting higher neutron flux, fission rate, and minor actinide concentrations compared to the square fuel pin arrangement.Keywords: Fuel Pin Arrangements, OpenMC, PWR, UN-PuN

Four types of surveillance programs were (are) realized in Slovak NPP’s: • “Standard Surveillance Specimen Program” (SSSP) was finished in Jaslovske´ Bohunice V-2 Nuclear Power Plant (NPP) Units 3 and 4; • “Extended Surveillance Specimen Program” (ESSP), was prepared for Jaslovske´ Bohunice NPP V-2 with aim to validate the SSSP results; • For the Mochovce NPP Unit 1 and 2 was prepared completely new surveillance program “Modern Surveillance Specimen Program” (MSSP), based on the philosophy that the results of MSSP must be available during all NPP service life; • For the Bohunice V-1 NPP was finished “New Surveillance Specimen Program” (NSSP) coordinated by IAEA, which gave arguments for prolongation of service life these units for minimum 20 years; • New Advanced Surveillance Specimen Program (ASSP) for Bohunice V-2 NPP (units 3 and 4) and Mochovce NPP (units 1, 2) is approved now. ASSP is dealing with the irradiation embrittlement of heat affected zone (HAZ) and RPV’s austenitic cladding, which were not evaluated till this time in surveillance programs. SSSP started in 1979 and was finished in 1990. ESSP program started in 1995 and will be finished in 2007, was prepared with aim of: • increasing of neutron fluence measurement accuracy; • substantial improvement the irradiation temperature measurement; • fixed orientation of samples to the centre of the reactor core; • minimum differences of neutron dose for all the Charpy-V notch and COD specimens; • the dose rate effect evaluation. In the year 1996 was started the new surveillance specimen program for the Mochovce RPV’s unit-1 and 2, based on the fundamental postulate — to provide the irradiation embrittlement monitoring till the end of units operation. The “New Surveillance Specimen Program” (NSSP) prepared in the year 1999 for the Bohunice V-1 NPP was finished in the year 2004. Main goal of this program was to evaluate the weld material properties degradation due to the irradiation and recovery efficiency by annealing too. The results showed us that the NPP V-1 RPV’s can be operated without any technical limitation for minimum next 20 years. Advanced Surveillance Specimen Program (ASSP) for Bohunice V-2 NPP (units 3 and 4) and Mochovce NPP (units 1 and 2) was prepared as the part of research project dealing with the WWER-440 units ageing management.

Reactivity Initiated Accident (RIA) leads to an unwanted increase in fission rate and power in a region of the reactor core confined around the position of occurrence. The power excursion due to such events may cause fuel rods failures and a subsequent release of radioactive material into the primary coolant of reactor, in severe cases, this release could damage nearby fuel assemblies. In nuclear power plants, RIAs are due to control system faults, e. g. control elements ejection/insertion, or rapid changes in temperature or pressure of moderator. In Boiling Water Reactors (BWRs), the control rod drop accidents (RDAs) at cold zero power have been deeply investigated, in fact, notwithstanding they are less frequent in comparison with the control rod ejection event in PWRs, in this kind of plant these conditions are the most severe in case of a RIA occurrence. RDA transient, comprised in the design basis events considered in safety analysis, may cause rod failures especially at high burnup. To simulate a RIA, a peaked power pulse is applied to a pre-irradiated and re-instrumented rodlet aiming at investigating the most important phenomena that could lead to the rupture of cladding tubes. This paper is focused on the investigation of the TRANSURANUS fuel performance code capability to predict the thermomechanical state of rodlets subjected to RIA tests. To this purpose the FK-1 test, carried out at the Nuclear Safety Research Reactor (NSRR) of Japan Atomic Energy Research Institute (JAERI), was simulated. This experiment is part of a set of 12 tests performed at the NSRR facility to study the performance under a reactivity initiated accident of BWR rodlets with burnup between 41 and 61 MWd/kgHM. In the FK-1 test, a STEP I BWR rodlet, previously irradiated in the Fukushima Daiichi Nuclear Power Station (Unit 3) operated by the Tokyo Electric Power COmpany (TEPCO) up to 45 MWd/kgHM, was subjected to a peak enthalpy insertion of 544 J/g. In this paper the code findings for the FK-1 test are discussed on the basis of the experimental data and the predictions of other stand-alone codes for transient analysis. The FK-1 predictions of FRAPTRAN (2001), FALCON (2003) and SCANAIR (ver. 3–2) are reported. The choice of fuel relocation model and important cladding properties (swelling, thermal expansion, thermal conductivity) was made relying on preliminary calculations whose results are also presented. Notwithstanding a satisfactory agreement between predictions and experimental data and a good agreement in the presented code-to-code comparison were envisaged, these results also emphasized the need to improve the models for FGR, heat transfer to plenum. Investigations are also required to ascertain possible contribution from fission gas to pelles thermal expansion. Ongoing modeling activity, performed at the ITU Joint Research Centre, is focused on a new model for FGR, ENEA (Agenzia nazionale per le nuove tecnologie, l’energia e lo sviluppo economico sostenibile) is expected, in the near term, to give a contribution to refine the model for plenum gas temperature. These activities should improve the description of RIA transient and further investigations on NSRR tests will be performed with newly developed models. The work presented in this paper will be part of ENEA contribution in FUMEX III project leaded by the International Atomic Energy Agency (IAEA) and aimed at the improvement of fuel codes predictions at high burnup.

The Chernobyl accident resulted in the contamination of the environment with long-lived radionuclides, including transuranium elements. The results, more accurate than the data, for the production of the basic long-lived radio-nuclides in the reactor core where the accident occurred. It is of interest to examine the Americium-241 contamination in greater detail and to compare the predicted estimates and the experimental data on the dynamics of the Am-241 accumulation in the environment. The maximum possible radiation consequences of Am-241 accumulation can be observed on territories subjected mainly to contamination as a result of precipitation of finely dispersed particles of the fuel composition. This radionuclide contributes to the internal irradiation dose resulting from inhalation and migration along the food chain.

Knowing exactly the nuclear core power of a nuclear reactor is one of the most important parameters for the operator; it is vital for safety as well as for economical matters. The secondary calorimetric is the only one where one can pilot on; it is a combination of measured parameters, of which the feedwater (FW) flow towards the steam generators is the most significant one. This feedwater flow can be measured by means of an ultrasonic flow meter, “LEFM CheckPlus™ system” instead of the commonly used venturis or diaphragms. In the Belgian Nuclear Power Plant (NPP) Doel 4, a new ultrasonic “LEFM CheckPlus™” feedwater flow measuring system has been installed in April 2008. The paper describes the consequences of the installation, as the total error on the secondary calorimetric decreases from the previous 1,3% to the current 0,8% with a possibility of further reduction to 0,4%. Additionally, the economical effects of the installation are calculated for a 1000 MWe power plant with venturi meters undergoing fouling. For the NPP Doel 4 it was an economically interesting investment since the payback period was only 45 days. Finally, the possibility of consuming the margin on the secondary calorimetric for a mini-power uprate is inspected, technically and economically. It is concluded that such a mini-power uprate is an interesting option for the NPP owner.

One of the lessons learned from Fukushima accident is that the existing procedures used in Nuclear Power Plants (NPPs) are not executed effectively and quickly enough after such an extended accident, for the accident is complex and people are too nervous in such a situation. Thus, emergency system that helps to raise diagnosis efficiency is necessary. In the paper, a quick diagnosis system on injection estimation of reactor core recovery and decay heat removal injection estimation is developed to meet the urgent needs and strengthen requirements for the training and application among utilities and nuclear regulators. The system will assist regulators to quickly know whether the currently flow will probably recover the reactor core, or whether the current injection capacity is sufficient to quench and recover the reactor core, directly after input present parameters into the system. In the system, Matlab method is used, and intuitive insights are considered, which is propitious to give immediate graphical interface and reduce possibility of human error.

The main part of radiation exposure to humans is formed by natural sources of ionizing radiation, of which radon is the predominant. A number of epidemiological studies have proven its detrimental effect on health — up to 15% of lung cancer cases are associated with daughter products of radon decay. Despite the very low radon content in the atmospheric air, the threat to health is reinforced by the fact that a person stays indoors for a long time, where radon activity is usually higher.The study aims to assess the radiation safety of students and staff of educational institutions in Tobolsk.A radonometric survey of 6 buildings of educational institutions in Tobolsk was conducted. As part of the work, the average annual value of the equivalent equilibrium volumetric radon activity, radiation doses and radiation risks caused by radon inhalation were determined for the studied premises. Radon activity was measured weekly by express method during the heating period.The average annual radon activity in the air of the studied rooms is estimated at 51–59 Bq/m3 (110–134 Bq/m3, taking into account the uncertainty of the estimate). The annual effective doses of internal radiation and radiation risks of students and staff associated with radon inhalation while in educational institutions were less than 0.6 mSv/year and 7.9×10–5–1×10–4, respectively. With exposure to residential and outdoor areas, radiation doses increase to 1.45 mSv/year, and the risk of death from lung cancer increases to 2.4×10–4.According to the results of measurements of radon activity, no non-compliance with radiation safety standards was found among the studied rooms. The effective radiation dose for Tobolsk school workers and students does not exceed an acceptable level (5 mSv/year). The risk of death from lung cancer can be classified as low, acceptable to staff, but not to the general population.Ethics. This study did not require the approval of the Ethics Committee.Acknowledgement. The author is grateful to the educational institutions of Tobolsk that participated in the research. The author is grateful to the reviewers for suggestions and comments that improved the quality of the article.Funding. The work was carried out within the Framework of the study "Ecological and geochemical transformations of soils of ecosystems of the Ob-Irtysh basin under the influence of natural and man-made factors" (R&D reg. number: 1024023000029-9-1.5.4).Conflict of interest. The author declares no conflict of interest.Received: 15.09.2025 / Accepted: 20.09.2025 / Published: 30.10.2025

Application of Data-Driven technology in nuclear Engineering: Prediction, classification and design optimization

The time-dependent radiation transport for a demonstration scale liquid-metal-cooled fast breeder reactor that has undergone a severe loss of sodium coolant is calculated with both a discrete ordinates and a diffusion theory solution for the real neutron flux shape. It is found that diffusion theory underpredicts reactivity levels by about $6 when compared to discrete ordinates. It is also found that the use of an initial adjoint neutron flux throughout the transient as a reactivity weighting function could seriously underpredict reactivity levels for a severely degraded reactor core. In both cases, there was an immediate termination of the excursion.The uncertainty of being limited to two fuel fields for an end of equilibrium cycle reactor core in SIMMER-II during the transient was greater than that due to microscopic cross-section shielding factor iteration and interpolation schemes. Fifty-energy-group reactivity coefficients were best duplicated in collapsing to a ten-energy-group set with an entire reactor integrated bilinear neutron energy flux spectrum.

The aging management system for the mechanical components of nuclear power plants (NPPs) must be established and used by the licensee in such a way that the quality of safety relevant components is guaranteed for the completely designed lifetime of the NPP. This demands an extensive plant life management with special emphases on the knowledge of the degradation in material properties. The basic safety concept in Germany observes this circumstance. Lifetime extension of the German NPPs is an aim of the current valid coalition agreement of the German government. Operational extension of interim storage facilities requires, in comparison to the aging management system for NPP, an aging management system adapted to the special circumstances of spent fuel storage casks. Extension of interim storage periods for spent fuel casks beyond the designed lifetime requires, in comparison to the components of an NPP, an increasing knowledge of material degradation with potential impact on cask integrity, e.g. le...