Mechanical Analysis and Evaluation

Traditionally, nuclear power plants in the U.S. provide baseload power to the power grid because they have less flexibility for ramping their output power than natural gas peaking plants. However, achieving climate goals to reduce the consumption of fossil‐based natural gas places pressure on nuclear power plants and other power generators to ramp up their power output to balance grid generation with demand. This paper presents the modelling and performance analysis of a nuclear reactor system (NRS) coupled to a liquid‐metal battery (LMB) to improve its dynamic response and enable its black start capability. The NRS and LMB thermal behaviour are modelled in Dymola, while the electrical dynamics of the LMB and power grid are modelled in RTDS‐RSCAD. Both simulation platforms are coupled and share their thermal and electrical data using a Transmission Control Protocol/Internet Protocol (TCP/IP) communication protocol. The dynamic performance of the NRS‐LMB integration is tested on the IEEE 9 bus, which demonstrates its ability to respond and provide frequency and voltage regulation. The black start capability of the NRS‐LMB is also evaluated by simulating a grid outage and using the LMB to supply the auxiliary loads required to bring the NRS back online as soon as possible. The results show that coupling an NRS to an LMB improves the system dynamic performance and enables it to black start after being disconnected from the grid for several days.

The present work is concerned with heat exchanger development for molten salt service, including the proposed molten salt reactor (MSR), a homogeneous reactor in which the fuel is dissolved in a circulating fluid of molten salt. It is an outgrowth of recent work done under the Next Generation Nuclear Plant (NGNP) program; what the two reactor systems have in common is an inherently safe nuclear plant with a high outlet temperature that is useful for process heat as well as more conventional generation The NGNP program was tasked with investigating the application of a new generation of nuclear power plants to a variety of energy needs. One baseline reactor design for this program is a high temperature, gas-cooled reactor (HTGR), which provides many options for energy use. These might include the conventional Rankine cycle (steam turbine) generation of electricity, but also other methods: for example, Brayton cycle (gas turbine) electrical generation, and the direct use of the high temperatures characteristic of HTGR output for process heat in the chemical industry. Such process heat is currently generated by burning fossil fuels, and is a major contributor to the carbon footprint of the chemical and petrochemical industries. The HTGR, based on graphite fuel elements, can produce very high output temperatures; ideally, temperatures of 900 °C or even greater, which has significant energy advantages. Such temperatures are, of course, at the frontiers of materials limitations, at the upper end of the performance envelope of the metallic materials for which robust construction codes exist, and within the realm of ceramic materials, the fabrication and joining of which, on the scale of large energy systems, are at an earlier stage of development. A considerable amount of work was done in the diffusion welding of materials of interest for HTGR service with alloys such as 617 and 800H. The MSR output temperature is also materials limited, and is projected at about 700 °C. (RR E) A different set of alloys, such as Alloy N and 242, are needed to handle molten salts at this temperature. The diffusion welding development work described here builds on techniques developed during the NGNP work, as applied to these alloys. There is also the matter of dissimilar metal welding, since alloys suitable for salt service are generally not suited for service in gaseous oxidizing environments, and vice versa, and welding is required for the Class I boundaries in these systems, as identified in the relevant ASME codes.

The determination of Design Ground Motion time history (or response spectra) is the primary and critical step to derive correct Seismic Design Inputs for a Nuclear Power Plant (NPP) design. Historically Design Ground Motion (design SSE input) for a NPP was determined by early version procedure provided in RG 1.60. It was based on a theory of deterministic approach; the resulting ground motion is given in acceleration response spectra located at free surface of a site. As a transition point, 1997 was the year where new procedure was developed and recommended in RG 1.165 based on the new theory of SSE ground motion probabilistic approach. RG 1.165 was authorized for application on all new NPPs’ design after 1997. With the advancing of PSHA approach, RG 1.165 was withdrawn and replaced with new RG 1.208 in 2008. RG 1.208 established an effective way through the similar probabilistic approach used in RG 1.165 by improving PSHA method. Both RG1.165 results and RG 1.208 results are focused on addressing site-specific design, its Ground Motion Response Spectra (GMRS) and Ground Motion Time History (converted from GMRS) are used as design inputs to specific Nuclear Island (NI) seismic design. To accomplish a Standard Design Certification, the RG 1.60 DRS is used to develop the Certified Seismic Design Response Spectra (CSDRS) by modifying control points on original RG 1.60 curves to broaden the spectra in higher frequency range. In reality, CSDRS serves as a good approach to define DRS and Design Ground Motion Time History for standard design of new NPPs in current timeframe, hence envelop the site-specific GMRS given in RG 1.208. In this paper, through the comparison of above US NRC regulatory requirements and Chinese regulatory requirements, gives recommendations on the determination of Design Ground Motion Response Spectra (or Time History), which serves as the basis for deriving seismic design inputs at required specific location (e.g. the bottom of NI foundation level) for potential “GEN III & Plus” plants in China.

The basic design solutions and characteristics of the VBER-300 reactor system for the power-generating units of 150–300 MW(e) nuclear power plants and regional nuclear heat-and-electricity plants are described. The reactor system implemented as a unit is based on the technologies and solutions used for marine nuclear power systems, which have been corroborated by experience in operating nuclear-powered icebreakers. The technical-economic advantages of floating power-generating units are substantiated.

The seismic analysis for the OPR1000 (previously entitled Korea Standard Nuclear Power Plant, now Optimized Power Reactor 1000 MW-Class) was carried out based on the SSE to assure the integrity of the reactor coolant pressure boundary and to secure a safe shutdown of reactor as well as the OBE to ensure the functionality of all the necessary systems for the continued operation of the reactor without undue risk to the health and safety of the public. In fact, the OBE with some supplementary functions, however, tends to control the design of nuclear power plants instead of the SSE due to its conservative analysis methodology. As a result of analysis work by means of more sophisticated techniques and continuously raised questions by the nuclear industry about the adequacy of using the OBE with at least one-half the maximum vibratory ground acceleration of the SSE as a design input, the OBE elimination design is exceptionally allowed through the alternative design rule by KINS (Korea Institute of Nuclear Safety: Independent nuclear regulatory expert organization), as long as the OBE is established as less than or equal to one-third of the SSE. Evaluated in this paper was the effect of the application of the OBE elimination-related alternative design rule on the KEPIC (Korea Electric Power Industry Code)/ASME Code Class 1 piping systems of an OPR1000-based nuclear power plant. The SSE (e.g. 0.3g) at least three-folds larger than the OBE (e.g. 0.1g) ground acceleration could decrease the flexibility of piping systems by adding more restraints to mitigate the increased SSE-induced seismic loads. Therefore, the flexibility of four piping subsystems, whose acquisition is one of the main purposes of the OBE elimination, was evaluated for an APR1400-based nuclear power plant (Korea’s Advanced Power Reactor 1400 MW-Class), which has practically adopted the application of the OBE elimination for the first time in Korea.

Traditional machine learning algorithms have problems such as overfitting, low accuracy, and difficulty in hyperparameter optimization when performing fault diagnosis.In order to improve the accident diagnosis ability of nuclear power plant reactor system, this paper combines Bayesian optimization (BO) algorithm with eXtreme Gradient Boosting (XGBoost) algorithm to develop a reactor accident diagnosis model.First, data preprocessing and feature quantity analysis are performed on accident data samples.Then, the BO algorithm is used to optimize the hyperparameters of the XGBoost model. Finally, the BO-XGBoost model is used to diagnose the operating conditions of seven nuclear power plants, and the diagnostic effects of various traditional machine learning classification algorithms are compared and analyzed.The results show that the BO-XGBoost model can achieve more efficient and accurate identification of reactor accident types, and the model has better generalization ability.This research can help nuclear power plant operators to accurately identify the types of reactor accidents, assist decision-making, and ensure the safe operation of nuclear power plants.

Offshore platform has long been in a complex and time-varying marine environment, would be inevitably influenced by environmental load and accidental loading, and result in the occurrence of various kinds of accidents or even disasters. The risk assessment for ocean platform has become an important issue in marine engineering field. However, the quantitative calculation of risks in the marine engineering industry is extremely difficult. The reason may attribute to the fact that there are few relevant statistical databases; and the mechanism of collision damage of marine structures is complex. The numerical simulation of impact caused by falling object is carried out by nonlinear finite element method (FEM). The PLS-RBF network is used for nonlinear interpolation purpose by combining the partial least square regression with the radial basis function (RBF) neural network. Therefore, the trained network can be implemented to obtain accurate collision results under various conditions by nonlinear interpolation. The feasibility and effectiveness of the proposed mapping approach for nonlinear finite analysis are testified by the simulation validation.

Nuclear power plant of VVER-440/213 designed by the former USSR have a number of special features. As a consequence of these features the transient behaviour of such a reactor system should be different from the PWR system behaviour. To study the transient behaviour of the Hungarian Paks Nuclear Power Plant of VVER-440/213-type both analytical and experimental activities have been performed. The experimental basis of the research in the PMK-2 integral-type test facility , which is a scaled down model of the plant. Experiments performed on this facility have been used to assess thermal-hydraulic system codes. Four tests were selected for {open_quotes}Standard Problem Exercises{close_quotes} of the International Atomic Energy Agency. Results of the 4th Exercise, of high international interest, are presented in the paper, focusing on the essential findings of the assessment of computer codes.

The impact of turbine missiles on barriers in nuclear power plants could cause severe damages and endanger public safety. This paper evaluates the available models and empirical methods that predict the extent of damages of turbine missile impact on composite and multiple barriers. A method of analysis is used in which the missile' s residual velocity after perforating each layer becomes the impact velocity for the next layer. Recommendations are made as to the most accurate models for barriers made of concrete and steel layers. INTRODUCTION Turbine missiles are irregularly shaped fragments that result from the failure or fracture of steam generator turbines [1] . The impact of missiles on surrounding facilities in nuclear power plants could cause severe damages and endanger public safety [2] . Various methods of damage analysis for composite and multiple barriers are reviewed. To evaluate the performance of available empirical methods , test results for 45 full-scale and small scale experiments were assembled from the literature for a variety of testing situations, missile types and orientations, targets and materials [3] . Missile impacts are often categorized as either hard or soft. Hard impacts such as that of a steel missile on concrete barriers essentially generate no significant deformation in the projectile [4] . On the other hand, in soft impacts [5] such as that of a steel missile on metal targets relative hardness plays an Transactions on the Built Environment vol 8, © 1994 WIT Press, www.witpress.com, ISSN 1743-3509

Offshore platforms work in complex and harsh environments, in addition to normal work load, environment loads, also threatened by accidental loads, such as the impact of ships. Major disaster caused by accidental loading will not only result in casualties and major economic losses, but also the surrounding coastal environment and marine ecosystem caused by pollution and destruction So it is essential to analyze and assess the behaviors of platform subjected to the accidental loading. The work presented in this paper focus on the risk of ship collision. On the basis of research on the risk analysis method and structural collision damage, offshore platform collision risk quantitative assessment system is built to achieve the risk assessment of offshore platforms from the traditional experience to scientific prediction type.

This paper presents the design methodology of a surface mounted brushless generator. A simple magnet configuration is presented for the optimum design. The design analysis and performance evaluation are carried out using finite element simulation techniques. The analytical design method is compared with finite element analysis. The simulation program is run on the prototype model after defining all input values and the values of electromagnetic parameters produced at various points under different speed and load conditions are tabulated

Possible genetic damage in the Czech nuclear power plant workers

One of the requirements for licensing a nuclear power plant in the U.S is the capability to survive and recover from a station blackout according to the U.S Nuclear Regulatory Commission (USNRC). Station blackout is the loss of all off-site and onsite power simultaneously. Therefore, experimental test facilities are being constructed and operated to test the performance of the related safety systems in a nuclear power plant. Design and construction of a test facility creates the need to perform scaling analysis to ensure proper representation of key components and phenomena of interest. One of the main outcomes of the scaling analysis is the quantitative estimation of the Similarity Level (SL), which requires derivation of dimensionless scaling parameters and prediction of appropriate input values for the scaling parameters. To study the performance of the Reactor Core Isolation Cooling (RCIC) system, the Nuclear Heat Transfer Systems (NHTS) Laboratory at Texas A&M University has constructed and is operating a RCIC test facility. This paper presents the scaling analysis with reference to a full-size RCIC system and the RCIC system turbine was used as the main component for scaling. The input parameters for dimensionless scaling parameters were obtained through experimental measurements and CFD analysis. The CFD analysis is for the ZS-1 RCIC system turbine model. The STAR-CCM+ CFD code was used in this study to create and run simulations for steady state normal and abnormal operating conditions for the NHTS-developed CAD models. The input for the dimensionless scaling parameters was estimated. Input parameters were collected both experimentally and from CFD simulations and inserted into these equations. As a result, a high degree of similarity was confirmed, with a minimum of 82% between the NHTS and full-size RCIC systems. The 82% represents the amount of transfer properties conserved between the two systems. Consequently, this high similarity level allows the NHTS RCIC system to be used to study the behavior of the full-size RCIC system under Beyond Design Bases Accident (BDBA) conditions. Future work is to study and model other components of the RCIC system such as the suppression chamber to estimate similarity levels and study their effects on behavior of the system under BDBA.

Stress and strain considerations in the design of mechanical components in nuclear power plants

At present, the water levels of high-temperature and high-pressure vessels such as voltage stabilizers and steam generators on ships' nuclear power plants are measured by the pressure principle. Affected by the load conditions in the vessels, the differential pressure water level measurement has problems of low accuracy and low reliability. The radar type liquid level gauge has the advantages of high precision and high reliability, and is an effective means of measuring the liquid level of high temperature and high pressure containers. In this paper, a FMCW system split guided wave radar based on the principle of frequency domain reflection (FDR) is proposed for liquid level measurement. The antenna probe part and the electronics part adopt a split structure, and the electronics part is installed in the secondary instrument cabin with better environmental conditions., the radar is easy to meet the index requirements of shock, vibration and irradiation. The high-frequency signal emitted by the radar antenna is conducted along the coaxial rod probe, and the signal will be reflected only when it encounters the surface of the material, which effectively eliminates the influence of obstacles such as the heating plate in the container, the internal reference pipe water replenishment device and the reference pipe connection. Non-blind zone measurement of liquid level to meet the liquid level measurement requirements of high temperature and high pressure vessels in ship nuclear power plants.