NPP Design Research Articles

CREATION AND USE OF THE LOCAL COORDINATE SYSTEM AT THE INDUSTRIAL SITE Khmelnitsky NPP

Today, nuclear power plants make up the main part of electricity production in Ukraine, providing the needs of domestic and industrial consumers, and after the full-scale invasion of the Russian Federation, systematic shelling of critical infrastructure, distribution stations and generation facilities, they became, in fact, the basis of Ukraine's energy security. Today, four nuclear power plants are in operation on the territory of Ukraine: Zaporizhzhya NPP (occupied by the Russian Federation), South Ukrainian NPP, Rivne NPP and Khmelnytsky NPP. The total capacity of operating power units on the controlled territory of Ukraine is 7,835 MW. The Chernobyl NPP is in the decommissioning stage. An important element in the design, construction and operation of nuclear power plants is to ensure the safe and reliable operation of construction structures of buildings and structures. To ensure these conditions during NPP design, design organizations develop work programs for conducting geodetic monitoring of buildings and structures both during construction and operation. In this article, we will consider the method of use GNSS measurements with its own base station and the creation of a local coordinate system to ensure the performance of geodetic monitoring of NPP buildings and structures.

Application of information technology to create the digital acoustic models of VVER

The IT used to create DAMPS is based on an interdisciplinary approach. DAMNR is presented as a self-oscillatory system, belonging to a special class of nonlinear dissipative systems capable of generating undamped oscillations. The use of DAMNR in the design and operation of NPP makes it possible to identify the sources of occurrence of ASW in the coolant, the conditions for their occurrence and frequency, and determine the influence of geometric parameters and the layout of coolant circuit equipment on the interaction of neutronic-physical, thermal–hydraulic and vibroacoustic processes. The use of DAMNR makes it possible to optimize engineering and design solutions when creating NPP by eliminating the conditions that cause the occurrence of unwanted self-oscillations of the coolant and vibroacoustic resonances in normal and emergency operating modes, as well as under conditions of shock impacts and seismic loads. Universal models and analysis methods were verified at Novovoronezh NPP.

Consideration of the tail dependence of copula functions in the selection of combinations of natural hazards

In accordance with the Federal rules and regulations of the Russian Federation (paragraphs 2.8, 2.9 of RR-064-17) and international standards (IAEA SSG-3), the OUAE (the object of the use of atomic energy) project should provide information on interrelated processes, phenomena and factors of natural origin identified during engineering surveys and studies, and the parameters of relevant combinations of external hazards are determined, for which an analysis of their impact on the safety of the NPP as part of the NPP design and beyond-design accidents should be performed. Combinations of hazards can have a stronger impact on the safety of the station than each impact considered separately, and the frequency of occurrence of the combination can be comparable to the frequency of occurrence of individual impacts. Currently, probabilistic approaches are not presented in the regulatory documents of the Russian Federation and international norms. The paper [1] presents a methodology based on the theory of copula functions, the practice of which has shown that the determination of the frequency and intensity of combinations is influenced by taking into account the empirical parameters of the distribution function: correlation coefficients, tail dependencies. The analysis of the effect of taking into account the value of tail dependencies on the result of choosing a model copula function within the selection of combinations of external natural factors that should be taken into account in the project of the OUAE is carried out. It is shown that the approach presented in [1] needs to be supplemented with an estimate of the value of the upper tail dependence based on statistical data, since due to a more complete application of the properties of copula functions, the depth of justification of the results increases.

Study on Function Extension of Pressurizer Fast Depressurization System for Gen III PWR NPPs

A nuclear power plant of Gen Ⅲ pressurized water reactor (PWR) in China has installed the pressurizer fast depressurization system for fast depressurization of the primary loop under severe accidents. With respect to this NPP, according to the general method and process of applying the probabilistic safety assessment (PSA) to the NPP design improvement, this study, focusing on the improvement plan of extending the pressurizer fast depressurization system function to the primary feed & bleed for depressurization to enable the use of this system as a standby pressure relief measure for the pressurizer safety valve, performs the PSA modeling and the feasibility evaluation and demonstration. Finally, this study shows that the improvement plan can reduce the core damage frequency of a NPP greatly, without any new negative effect, so the plan is feasible and can be implemented. Consequently, the NPPs should fully tap the potential of existing system equipment to further improve the safety and economical efficiency of NPPs.

SYSTEM DYNAMICS MODEL DEVELOPMENT BY ANALYZING CAUSES OF FUKUSHIMA DAIICHI NPP UNIT-1 ACCIDENT

— Fukushima Daiichi NPP accident occurred on March 11, 2011. A system dynamics model is developed by analyzing the chronological progression and causes of the Fukushima Daiichi NPP unit-1 accident applying causeeffect relationship and qualitative causal loop and feedback loop. The accident progression demonstrates that the safety system of the NPP could not function properly, resulting in the loss of core cooling and confinement of radioactive materials functions, and a severe accident with hydrogen explosion occurred. The system dynamics model shows that earthquake, tsunami induced flooding, design failure of Tsunami height, loss of all AC power, loss of DC power, loss of ultimate heat sink, failure of HPCI system, failure of IC system, design complexity of isolation valves of IC and lack of training on IC operation, delayed water injection to reactor, delayed operation of containment venting, lacking of compressed air, high reactor pressure, and harsh radiological condition are the causes of this severe accident. High reactor pressure and harsh radiological condition were produced during the accident, which positively boost accident progression by demonstrating the effect of system dynamics. The conservative design to protect power sources from natural events and the loss of all power for an extended period must be considered during future design of NPP.

SYSTEM DYNAMICS MODEL DEVELOPMENT BY ANALYZING CAUSES OF FUKUSHIMA DAIICHI NPP UNIT-1 ACCIDENT

Fukushima Daiichi NPP accident occurred on March 11, 2011. A system dynamics model is developed by analyzing the chronological progression and causes of the Fukushima Daiichi NPP unit-1 accident applying causeeffect relationship and qualitative causal loop and feedback loop. The accident progression demonstrates that the safety system of the NPP could not function properly, resulting in the loss of core cooling and confinement of radioactive materials functions, and a severe accident with hydrogen explosion occurred. The system dynamics model shows that earthquake, tsunami induced flooding, design failure of Tsunami height, loss of all AC power, loss of DC power, loss of ultimate heat sink, failure of HPCI system, failure of IC system, design complexity of isolation valves of IC and lack of training on IC operation, delayed water injection to reactor, delayed operation of containment venting, lacking of compressed air, high reactor pressure, and harsh radiological condition are the causes of this severe accident. High reactor pressure and harsh radiological condition were produced during the accident, which positively boost accident progression by demonstrating the effect of system dynamics. The conservative design to protect power sources from natural events and the loss of all power for an extended period must be considered during future design of NPP.

Upgrading of Regulatory Documents: Probabilistic Safety Analysis in Support of Decision-Making in the Design and Operation of NPP

Upgrading of Regulatory Documents: Probabilistic Safety Analysis in Support of Decision-Making in the Design and Operation of NPP

Radiological consequences and lessons of the Chernobyl NPP and «Fukushima-1» NPP radiation accidents

35 years have passed since the Chernobyl NPP accident, 10 years – since the «Fukushima-1» NPP accident. At the present time extensive activities on the remediation of the consequences of two major large-scale radiation disasters are performed in the Belorussia, Russian Federation, Ukraine, and Japan. Releases of radiologically significant radionuclides after the Chernobyl NPP accident correspond to 14 exaBecquerel – higher up to an order of magnitude compared to 3 emergence power units of the «Fukushima-1» NPP. The significantly lower release rate and deposition of 80% of the radionuclides released into the atmosphere on the surface of the Pacific Ocean lead to lower up to several orders of magnitude radioactive contamination of the Japanese territory compared to the territories of the former USSR and neighboring countries after the Chernobyl NPP accident. Collective dose to the public due to the Chernobyl NPP accident is higher up to several orders of magnitude compared to the dose to the Japanese population after the «Fukushima-1» accident. No statistically reliable long-term medical consequences are expected for all groups of the Japanese public, additionally exposed due to «Fukushima-1» accident. 134 emergency workers have developed acute radiation sickness due to the Chernobyl NPP accident. Emergency workers with doses higher than 150 mSv had increased radiation-induced morbidity with leukemia and solid cancers. Among the individuals, that were kids or adolescents in the exposure period after the Chernobyl NPP accident and residing on the territories of Belorussia, Ukraine and four most radioactively contaminated regions of the Russian Federation, morbidity with thyroid cancer is increase by a factor of 10 compared to the pre-accidental levels. The following lessons of the Chernobyl NPP and «Fukushima-1» NPP can be derived: faults in the NPP design and lack of response after the recognition of the faults; lack of timely full-scale prophylactic with iodine; unjustified resettlement of the residents of the radioactively contaminated territories several years after the accident.

MULTIPLE CHOICES OF REACTOR CORE NUCLEAR DESIGN FOR ACP100’S APPLICATION IN DIFFERENT SCENARIOS

ACP100 NPP designed by CNNC (China National Nuclear Corporation) is a 125MWe, multi-purpose small modular reactor based on pressurized water reactor technology; it adopts the integrated reactor technology. Different application scenarios bring up different design requirements: some require high compactness, but others care more about a longer cycle length, and some may require a fully mature and conservative design; thus, multiple design choices need to be proposed. Also, the same and most important thing cared by all users is that, the design needs to be validated to satisfy the current nuclear safety standards, and lower cost would be always preferred. Core nuclear design is a key part of the whole NPP design. Basically, nuclear design target of ACP100 is to achieve a reasonable good balance during longer cycle length, larger discharge exposure for fuel assemblies, and maximally using the mature technologies, and of course, with sufficient reactivity control ability for safety assurance. Aiming at satisfying all these different needs maximally, a strategy of supplying multiple nuclear design choices is proposed for ACP100: choice 1. Boron-free plan, this is a compact design with no need for chemistry and volume system, no need for daily boron adjustment and relative waste storage; choice 2. Boron and rod co-controlled plan, this is similar with large commercial PWRs, with a lower power peak factor and suitable for broad location sites. Both choices load 57 units of the same type fuel assemblies CF3S (with height reduced from CF3 fuel assemblies) per cycle, and both adopt partial reload and shuffle fuel management strategy to achieve larger discharge exposure. Gd is loaded in the fuel rods in both choices to help control reactivity. Choice 1 loads much more control rod clusters than choice 2, and of course, reactivity adjustment and compensation during operation is totally different between them. Using suitable and reliable software to simulate the core, through large amount of optimization, both choices achieve a 24-month fuel cycle length; the average discharge exposure of fuel assemblies reach about 40000MWd/tU, which is competitive among SMRs, especially for boron-free ones; and sufficient reactivity control ability and safety margin is validated to fully meet the reactor safety requirements.

Infrastructure risk-oriented advantages of low-power nuclear power plants

Variants of classification of risks of NPP design over the entire life cycle are proposed. The concept of integral risk is introduced and, on its basis, approaches to the creation of a low-risk nuclear power system based on small and medium power units are formulated. The key role of the human factor in the formation of risks is reflected. Comparison of the risks of NPPs of large and small capacity is carried out. .

Small Nuclear Power Plants – Primary Avenue for Reducing Risks and their Consequences in Nuclear Power

The potential for improving safety and systems efficiency via unit power reduction of NPP in spite of an increase of the specific costs is investigated. An approach to significant reduction of the risk due to the human factor by switching to small- and medium-size power-generating units is considered. This approach can becharacterized as a method of risk reduction by means of ‘power insurance.’ The qualitative technical and economic effects from the reduction of different forms of risk in the implementation of small- and mediumsize NPP designs as compared with large NPP are analyzed: public acceptance; search for investors and their financial risk; availability of NPP sites; cost increase due to longer construction times; radiation and nuclear safety risks; electric power interruptions; minimization of the reserve power in the power system; and, adaptability to the rate of regional energy growth. Variants of the classification of integral risks over the entire life cycle of an NPP design are presented.

Conceptual Design of Small NPP with RITM-200 Reactor

A small ground-based nuclear power plant with two RITM-200 reactors is examined. The plant is intended for electric and thermal energy production and desalination of water. The general layout of the site, arrangement of the basic engineering and technical structures, characteristics of the reactor and steam turbine plants, and systems and equipment of the power plant are examined. RITM-200 is used in the universal nuclear icebreakers now under construction. Its compactness and capacity for operating in a maneuvering regime, flexibly following the required power, makes it possible to attain high technical and economic performance and plant efficiency. The design of a power plant with RITM-200 complies with the current global requirements and trends in the field of safety of nuclear power plants. The requisite number of active and passive safety systems are provided for this purpose. Optimization R&D to improve the technical and economic characteristics of the plant and increase its commercial attractiveness is to be performed at later stages of the design process.

Validation &Verification context for NPP design and construction

Nuclear power, as one of the most sophisticated and technologically advanced industry sectors, sets very demanding safety requirements. Fulfilment thereof is mostly to be implemented and justified at the design and construction stages. The process of Validation and Verification (V&V) for such facilities should cover all possible disciplines and cases, therefore a systematic approach to its organization is needed, which is not fully. The article objective is to determine the reasons for such strict safety requirements and explore the methods for achieving it, including the V&V process, as well as determining the factors (V&V dimensions) forming the V&V context, which is going to be used for choosing a specific type of V&V activities (to be conducted during the design and construction phases). Then, the article shall determine the interrelations between disciplines that use the V&V results depending on the context. V&V Context is setting up the parameters defining specific V&V activities. Generally, the V&V Context includes the following dimensions: Discipline (Construction, Electrical, Instrumentation and Control), Hierarchy (Design Levels – Plant, System, Components and Structures levels), NPP Construction Project phases (Design, Construction, Installation, Commissioning). The V&V process is a complex one and it will be different for each combination of the above-mentioned V&V dimensions together forming the V&V Context.

Modernization of Fire Extinguishing Installations at Ukrainian NPPs

Fire safety of nuclear power plants is a relevant task, which shall be solved using the latest approaches taking into account NPP design and construction practices, including those in foreign countries.
 The paper describes results of analysis of NPP fire safety in Ukraine based on national and international assessments, identifies problems in this area and presents priority measures aimed at NPP safety improvement, in particular modernization of fire protection of safety system rooms and implementation of gas and powder fire extinguishers at Ukrainian NPPs.

Разномасштабные задачи тепломассообмена в атомной энергетике

This paper presents a review of the results obtained in OKB «GIDROPRESS» taking into account the experience gained in the field of numerical modeling of heat and mass transfer in various nuclear power facilities. The benefits of using CFD technology in the NPP design is the ability, based on a limited number of experimental data, to study in detail the heat and mass transfer processes in the facilities in order to confirm or improve design solutions at the advanced scientific and technical level. A number of problems are presented, for solution of which the modern numerical methods of computational hydrodynamics with high-performance computer technology were used. New possibilities of computational modeling with the use of modern supercomputer computing technologies, as well as accompanying computational complexity and difficulties of analysis of results are shown. Examples of use of the considered technology for modeling experimental rigs and full-scale objects under different operating conditions are given. The automatic post-processing of the results is shown. The procedure allows to analyze large problems with mesh size up to 1 billion control volumes by integral parameters, such as mass flow distribution at the core inlet and outlet, distribution of coolant heat-up over the core fuel assemblies, etc.

Verification of Benchmark Experiments by Determination of the Leakage Neutron Spectrum from the Surface of a Depleted Uranium Sphere with Embedded Radionuclide Sources

Computational validation is presented for previous benchmark experiments designed to determine the energy spectra of the neutron current emanating from the surface of a uranium sphere into which small sources were arranged successively at the center. Such experiments must meet stringent reliability and error requirements, since they are intended for the fi nal check of new or adjusted systems of neutron constants recommended for the calculation of neutron fi elds. The comprehensiveness of the validation was secured by the fact that experiments with the three sources 252 Cf, Pu–Be, and Pu–B were validated and the calculations were performed using three systems of neutron constants BNAB-78, BBF, and BUGLE-96. The codes used to calculate the ionizing radiation during the design and operation of NPP and other facilities using atomic sources are now being improved in connection with the development of nuclear power and advances in computing technology. Complex three-dimensional problems can now be solved taking account of more stringent error requirements of the results. As a rule the, libraries of the interaction constants for neutrons interacting with the nuclei of different isotopes used in the calculations make an appreciable contribution to the error. For this reason, there is now a need to check them with the aid of benchmark experiments. Considering the importance of the results heightened requirements are now imposed on the reliability, ordinarily evaluated by performing verifi cation by means of comprehensive calculations. The simplest geometry is chosen for the experiments. The materials studied are fabricated in the form of spheres; small radionuclide sources are used in the experiments so that they can be modeled by point isotropic sources located at the center of a sphere. The geometry chosen for the experiments makes it possible to measure reliably the energy distribution of the neutron current from unit surface area of a sphere, which is the main result, and moreover to use simple one-dimensional codes and neutron-constants libraries which have proven themselves well in practice. The experimental data presented in the form of the energy distribution of the neutron current are informative, and when validating new or improved libraries of neutron constants they make it possible to fi nd on the energy scale appreciable discrepancies with experiment and introduce the corresponding corrections. The aim of the present work was to validate the results of previous benchmark experiments on determining the energy spectra of the neutron current emanating from the surface a uranium sphere. The choice of uranium is associated with the fact that it is one of the main materials used in nuclear technology.

Analysis of damaged welds no. 111 in the PGV-1000 steam generator and damage repair proposals

Reliability of complex facilities, such as nuclear power plants, under construction or in operation is fundamental to the safety of humans and the natural environment. The key reliability-related factors are absence of errors in structural design and calculations, proper selection of materials and the manufacturing technology, quality of the materials used and the onsite welding operations, conditions of operation and in-service inspection.Despite the fact that much attention is given in the course of the NPP design and operation to ensuring reliability, the experience of operation has demonstrated that there is a potential for crack formation in welded joint No. 111 of the PGV-1000 M steam generator. The crack nucleation and growth mechanism has not been yet unambiguously identified.This paper presents the results of a study into the causes for the metal damage in the region of welded joint No. 111 between the hot header and the steam generator vessel's nozzle of Dn1200.

核电厂中主要弱β释放体的放射性量及控制方法研究

随着核电厂中对正常满功率运行时裂变产物和腐蚀产物控制水平的不断提高，主要的弱β释放体(如氚和C-14)对公众的辐射影响变得日益重要。GB 6249-2011中对流出物中氚和C-14的单机组和厂址排放总量规定了控制值，也对内陆厂址排放口下游1 km处受纳水体中氚浓度进行了限制。本文根据核电厂中氚和C-14的主要产生途径，采用理论方法对其产生量进行分析，并从其产生途径和设计的角度，提出了切实可行的对释放到环境中的氚和C-14进行控制的方法和措施，以用于优化电厂设计，为环境影响评价和安全分析提供审评的依据。 As the control technology of fission and corrosion products develops during the full power operation in the nuclear power plant, the impact from the main β emitters (like tritium and Carbon-14) to people becomes increasingly important. Certain limits of the tritium and carbon-14 concentration in the effluences for both single unit and site are established in GB6249-2011, also, it includes the tritium concentration restriction in the 1 km downstream water from the discharge canal for inland sites. Based on the main generation pathway, a theoretical method is applied to analyze the tritium and carbon-14 generations in this paper, and some tritium and carbon-14 release control recommended measures are provided from both the generating pathway and design aspects. Also, these measures could be used for the NPP design optimization, safety analysis and environment assessment.

Tornado Regime at the Belarus NPP and the Threshold Probability of Tornado Danger

Data on tornado passage are collected and analyzed as part of engineering and hydrometeorological research for the construction of the Belarus NPP. Statistical analysis of the data showed that the probability of a tornado passing through the NPP site is 2.94∙10–6 reactor-yr and the computed tornado magnitude is 2.5 on the Fujita scale. The main working characteristics of a likely tornado incorporated in the NPP design are determined. It is noted that it would be expedient to refine the threshold probability of tornado danger by taking account of the tornado loads on the NPP.

Next Generation of Soil-Structure Interaction Models for Design of Nuclear Power Plants

The author here shares his vision of next-generation models for seismic soil-structure interaction (SSI) analysis. These models should combine reasonable considerations of wave effects in half-infinite soil with a correct representation of nonlinearity in the structure, and in both the so-called near field, i.e., in that part of soil near a base mat, and in the soil-structure contact surface. The far field, i.e., all of the soil except for the near field, is treated as a linear horizontally layered medium, as is currently done in the well-known program SASSI. The importance of considering nonlinear effects even in very stiff structures like NPPs was shown by the March 2011 Great East Japan Earthquake that hit northeastern Japan’s Pacific coast. Although the idea of calculating SSI wave effects in the time domain has been around for several decades ago, current NPP design practices are linear. Next-generation SSI models should enable practical time-domain analysis. The author suggests a road map – the sequence of problems to be solved to achieve a proposed level. Some of these problems have already been solved, at least in principle, but other solutions are yet to be found. The author describes the current status of his research and ideas about implementing modern computational techniques such as parallel computation.