Radiation risk assessment for nuclear power plant staff in different emergency situations

The paper presents the results of radiation risk assessment for the staff of a nuclear power plant design during design basis accident (spent nuclear fuel assembly falling on fuel in reactor core or storage pool during refueling operations) and a beyond design basis accident (large leakage of the primary coolant with failure of the active part of the emergency cooling system and complete blackout for 24 h). The assessment is based on state-of-the-art radiation risk models from the United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR) and the International Commission on Radiological Protection (ICRP). The calculation of risk indicators for occupational exposure of NPP staff in emergency situations was carried out on the basis of data obtained using a computational module created in the COMSOL 5.6 multiphysics software, doses from a radioactive cloud and internal exposure due to inhalation for such radionuclides as 134Cs, 137Cs, 131I, 133I, 90Sr. A feature of this approach is the detailed consideration of the NPP industrial site infrastructure, which allows obtaining a more accurate assessment of the radionuclide air distribution and fallout.

The H2020 McSAFER Project: Main Goals, Technical Work Program, and Status

This paper describes the main objectives, technical content, and status of the H2020 project entitled “High-performance advanced methods and experimental investigations for the safety evaluation of generic Small Modular Reactors (McSAFER)”. The main pillars of this project are the combination of safety-relevant thermal hydraulic experiments and numerical simulations of different approaches for safety evaluations of light water-cooled Small Modular Reactors (SMR). It describes the goals, the consortium, and the involved thermal hydraulic test facilities, e.g., the COSMOS-H (KIT), HWAT (KTH), and MOTEL (LUT), including the experimental programs. It also outlines the different safety assessment methodologies applied to four different SMR-designs, namely the CAREM (CNEA), SMART (KAERI), F-SMR (CEA), and NuScale. These methodologies are multiscale thermal hydraulics, conventional, low order, and high fidelity neutron physical methods used to demonstrate the inherent safety features of SMR-core designs under postulated design-basis-accident conditions. Finally, the status of the investigations is shortly discussed followed by the dissemination activities and an outlook.

Radiation Dose Evaluation of Typical Design Basis Accident for Advanced PWR in China

The loss of coolant accident (LOCA) is one of the typical design basis accidents for nuclear power plant. Radionuclides leak to the environment and cause harm to the public in LOCA. Accurate evaluation of radioactivity and radiation dose in accident is crucial. The radioactivity and radiation dose model in LOCA were established, and used to analyze the radiological consequence at exclusion area boundary (EAB) and the outer boundary of low population zone (LPZ) for Hualong 1. The results indicated that the long half-life nuclides, such as 131I, 133I, 135I, 85Kr, 131mXe, 133mXe and 133Xe, released to environment continuously, while the short half-life nuclides, such as 132I, 134I, 83mKr, 85mKr, 87Kr, 88Kr, 135mXe and 138Xe, no longer released to environment after a few hours in LOCA. 133Xe may release the largest radioactivity to environment, more than 1015Bq. Inhalation dose was the major contribution to the total effective dose. The total effective dose and thyroid dose of Hualong 1 at EAB and the outer boundary of LPZ fully met the requirements of Chinese GB6249.

Cold Spray Technology for Oxidation-Resistant Nuclear Fuel Cladding

Light water reactors (LWR) use zirconium-alloy fuel claddings; the tubes that hold the uranium-dioxide fuel pellets. Zr-alloys have very good neutron transparency; but during a loss of coolant accident or beyond design basis accident (BDBA) they can undergo excessive oxidation in reaction with the surrounding steam environment. Relatively thin oxidationresistant coatings on Zr-alloy fuel cladding tubes can potentially buy coping time in these off-normal scenarios. In this study; cold spraying; solid-state powder-based materials deposition technology has been developed for deposition of oxidation-resistant Cr coatings on Zr-alloy cladding tubes; and the ensuing microstructure and properties of the coatings have been investigated. The coatings when deposited under optimum conditions have very good hydrothermal corrosion resistance as well as oxidation resistance in air and steam environments at temperatures in excess of 1100 °C; while maintaining excellent adhesion to the substrate. These and other results of this study; including mechanical property evaluations; will be presented.

Experimental research of the air permeability of concrete in the compressed zone of the containment for power units no. 1, 2 of the South-Ukrainian NPP

The determination of the air permeability of concrete in the compressed zone of NPP’s containment under conditions of excessive air pressure in the subshell space of the reactor compartment under a maximum design basis accident is necessary to confirm the localizing functions of the containment when extending the lifetime of power units. Approaches to determining the localizing ability (air permeability) of concrete in the compressed zone of the containment of NPPs with WWER-1000 type reactors under the impact of excessive air pressure under conditions of maximum design basis accident are discussed in the article. The designed testing installation for experimental studies of the air permeability of concrete by the method of stationary radial filtration is described and the results of experimental researches of the air permeability on the installation are presented for samples simulating the composition of concrete used under construction of the containment for Power Units No. 1, 2 (series B-302 and B-338, respectively) SD “South-Ukrainian NPP”. A comparative analysis of the results of abovementioned experimental researches with the results of experimental and theoretical work on the study of air permeability of the similar samples-imitators of concrete by the method of stationary linear filtration, as well as the results of measurements for samples of concrete sampled directly from the compressed zone of containment for Power Unit No. 2 of SD “South-Ukrainian NPP” was fulfilled. The results are explained by processes of compaction of the concrete structure under a complex stress state that occurs under specific hardening conditions and during longtime compression by pre-stressed reinforcing ropes, which takes place under conditions of real containment operation. It was concluded that the simulators made according to the currently accepted technology cannot provide a correct determination of the localizing ability of the NPP’s containment, since the conditions and time of containment concrete hardening, as well as the longtime constant compressive action of reinforcing ropes, cannot be correctly modeled in full under the formation of samples-imitators. The issue of the containment concrete samplesimitators creating in laboratory conditions requires additional study, improvement of technology and the development of new approaches with maximum physical modeling of the conditions characteristic of the operation of the NPP containment.

Assessment of the Dynamic Loads Applied to the RBMK-1000 Subreactor Room Walls and Slab during Vapor Explosion

One of the conditions for the safe operation of a nuclear power plant (NPP) unit is a comprehensive design and experimental justification of its failure-free operation in all operating modes and limitation of accident radiation consequences, including those in the case of severe beyond design basis accidents. According to the nuclear power industry development plans in Russia, new NPPs equipped with RBMK-1000 reactors are not supposed to be constructed in the future. Although the assigned service life of RBMK-1000 based power units that remain in operation is close to expiration, these power units account for most of the electricity generation in the total amount of nuclear power capacities in Russia (about 40%); therefore, the relevant industry organizations have decided to extend their operation. This article analyzes the severe accident evolvement scenario at an RBMK-based NPP during the stage of severe core damage, in the course of which fuel-containing masses collapse into the subreactor space filled with water. Once fuel-containing masses emerge in the sub-reactor room, they come in interaction with the reactor base concrete. There is a potential danger of the concrete floor slab melting and the corium collapsing into the bubbler pool water. The main strategy foreseen for keeping the molten core within the reactor space boundaries involves decay heat removal from the reactor and cooling of the support metal structures by supplying water. However, the filling of the subreactor space with liquid may give rise to conditions under which vapor explosion can occur. The maximum dynamic impact applied to the RBMK-1000 subreactor room walls in the event of possible interaction between the molten corium and water during a severe beyond design basis accident is estimated. It is shown that when the corium melt interacts with a large amount of water in the subreactor room, the kinetic energy of the resulting water vapor is sufficient to cause significant destruction of the power unit building. When the water level in the subreactor room falls below one meter, the destruction hazard becomes less probable. The mass of hydrogen released as a result of the interaction is also estimated.

Results of Structural Analysis of WWER-1000/320 Containment Behavior under Severe Accidents

The paper presents the review and evaluation of the containment prestressing system within reinforced concrete structures under seismic loads and severe accidents. Given the complex design of the containment, the detailed finite element model has been developed and used to describe real containment behavior. Containment stress and strain state was calculated by modern LIRA software. The first stage analyzed the results of WWER-1000/320 containment stress and strain state calculation under a combination of loads caused by maximum design basis accident (MDBA) and safe shutdown earthquake (SSE) and defined minimum acceptable tension of tendons. The research determines the minimum acceptable tension of tendons in the containment prestressing system, and evaluates the strength and reliability of containment structures under a combination of loads in normal operation + design-basis accident + maximum design earthquake (NO + DBA + MDE). The verification calculations have been performed using tendon tension of 780 ton-force in the cylindrical part of the containment and 760 ton-force in the containment dome. The second stage covered the analysis of severe accident parameters (pressure and temperature) and the results of calculation. Stress and strain state in ZNPP-1 containment has been calculated, parameters (pressure and temperature) under which the containment can loss its protective and isolation functions have been identified, calculation results have been analysed and conclusions of containment structural integrity and ensuring the implementation of the design confining functions have been made. Based on the calculation results, it can be concluded that strength of the containment cylindrical part during a beyond design-basis accident cannot be ensured under parameters t (temperature) = 120°С, p (pressure) = 0.6 MPa.