Commercial Nuclear Power Plants Research Articles

Implications of HALEU fuel on the design of SMRs and micro-reactors

Since the construction of the first commercial nuclear power plant in 1957, the nuclear power industry has operated under the philosophy of economy of scale - the idea that increased power plant size accounts for higher economic efficiency. However, there has been a recent shift in direction; small modular reactors (SMRs) and micro-reactors are being considered as potentially wise investments for commercial power producers in that they can provide advantages that large-scale reactors may not possess in terms of reactor safety and investment risk. However, this may come at the risk of a higher levelized cost of electricity (LCOE). LCOE may be reduced by enriching the fuel passed its regulated limit of 5 wt% (w/o) 235U. The high assay low-enriched uranium (HALEU) fuel (5–20 w/o 235U) is introduced to increase the plant capacity factor, which thereby decrease fuel supply costs and reduce the LCOE. While decreasing plant LCOE seems like a clear advantage, several issues may result from increasing enrichments to the HALEU level in an SMR or micro-reactor design. This paper aims to shed light on these issues and address how they may affect the overall reactor design by using HALEU fuels in these reactors.This paper first discusses the notable effects on a reactor design with higher enrichment, then analyzes a SMR case study based on the NuScale’s 160 MWth SMR design. The case study reveals that the SMR with higher enriched fuel was able to double both fuel burnup and cycle time with an average core enrichment of 8.34 w/o and a maximum average assembly enrichment of 9.10 w/o. Moreover, this higher enriched core was found to operate with a maximum global peaking factor of 1.86, well below the published limit of 2.0. Likewise, the maximum axial flux offset of −2.4% and the maximum boron concentration of 1757 ppm both remain within their respective safety constraints. Notable fission poisons, such as 149Sm and 135Xe, were also found to sharply increase in the HALEU core. Additionally, the average fuel temperature and peak cladding temperature fell within their respective safety constraints. Core-averaged flux, fluence, cladding creep, and post-shutdown decay heat were also investigated. Lastly, the higher enriched core was found to reduce LCOE by approximately 1.23 $/MWh.

Foreword: Special section on the Seismic Analysis and Risk Assessment of Nuclear Facilities

The seismic design and risk assessment of nuclear facilities is critically important, not just for safety, but also for economics—the latter especially for commercial nuclear power plants (NPPs). T...

CHF enhancement of downward-facing saturated pool boiling on the SCGS-modified surfaces with multi-scale conical pin fin structures

Enhancing the critical heat flux (CHF) of external reactor vessel cooling (ERVC) to achieve the strategy of in-vessel retention (IVR) has been deemed as an essential requirement for the security of modern commercial PWR nuclear power plants. A new type of multi-scale conical pin fin structures had been designed to meet this need. These conical structures of four different axial heights (1.0, 1.5, 2.0 & 2.5mm) and similar radial geometric parameters were fabricated on four heating surfaces of copper substrates by supersonic cold gas spray technique (SCGS), and the conical pin fin structures were combined with a thin layer of micro porous structures. Transient-state (quenching) and steady-state (heating) pool boiling experiments had been respectively conducted on 6 different downward-facing surfaces under different inclination angles including 5°, 30°, 45°, 60°, 75° & 90° (in vertical), which included above-mentioned four pin fin surfaces, a plain surface and a coated surface covered with a 1mm-thick layer of uniform porous coating fabricated by SCGS with copper particles. The experimental results showed that the CHF increased with the inclination angles. The CHF of all the SCGS-modified surfaces were significantly enhanced compared with those of the plain surface. The maximum CHF value and the maximum CHF enhancement belonged to the 2.0mm-height pin fin surface whose CHF values were more than 1.7 times of those of the plain surface, and up to 2.1 times in 90°. The 1.0mm pin fin surface had the greatest performance in boiling heat transfer coefficient (HTC) and its CHF-enhancing performance also greatly exceeded that of 1.0mm uniform porous coating surface. Remarkable performances of multi-scale conical pin fin structures had been experimentally confirmed, hence the safety margin of IVR-ERVC will be greatly enhanced after adopting them. In addition, the CHF enhancement mechanism of the multi-scale conical pin fin structures was analyzed.

Techno-economic analysis of long-duration energy storage and flexible power generation technologies to support high-variable renewable energy grids

Techno-economic analysis of long-duration energy storage and flexible power generation technologies to support high-variable renewable energy grids

Corrosion in Advanced Nuclear Reactors

Commercial nuclear power plants in the United States are light water reactors (LWRs) that use water as a coolant, with temperatures ranging between 280°C and 320°C. Water purity is tightly controlled; nevertheless, the high temperatures expose components in the water circuit to degradation by corrosion by many different types of environmental attack. Uniform corrosion occurs across the entire surface of a material and is prevalent in engineering systems to some extent. Site-specific corrosion processes, such as crevice corrosion, intergranular attack, or galvanic corrosion, are common in complex engineering systems that consist of multiple materials joined by welds or other solid state joining processes. Even within a single component, if second phase strengthening is used or if surface defects are present, localized corrosion can occur.

Integrated Security System (ISS) Design and Evaluation for Commercial Nuclear Power Plant

Physical security system, which is also called physical protection system, is very crucial in the nuclear industry for protecting staff, visitors, buildings, assets, and nuclear materials against theft, sabotage, and harmful activities. Theft of nuclear materials has a major impact on the essence of nuclear safeguards. Sabotage of a nuclear facility could endanger the public at large. Reviewing the published literature, it is found that there are no complete physical security system designs based on an integrated network of electronic devices that are devoted to commercial NPPs. And there is no definite evaluation factor that was set to approve such a system. This paper is an evolving solution to this deficiency by proposing an unpreceded integrated security system design applicable to a commonly structured physical layout of any commercial NPP. This proposal provides comprehensive security coverage for the NPP boundaries employing a high level of integration for all subsystems communicated via an IP data network controlled by central management software. This paper is proposing also testing procedures to be followed to evaluate the proposed design. The security system effectiveness will be calculated using mathematical codes by assuming external intrusion attack scenarios. Attributes of each attack scenario will be numerically introduced to the evaluation software EASI and ASSESS codes developed by Sandia Labs, USA. This paper also proposes a threshold value of such security system effectiveness which should be achieved by the commercial NPP security system to achieve the so-called security license.

Experimental study on the advective heat flux of a heat exchanger for passive cooling of spent fuel pools by temperature anemometry grid sensor

In commercial nuclear power plants spent fuel assemblies are usually stored in actively cooled water pools. The continuous decay heat release represents a potential risk in case of a station black out scenario. Thus two-phase passive heat removal systems are a key technology to enhance the safety of nuclear power plants. Such systems work only by the energy provided from the heat source, e.g. by the maintenance of a natural convection cooling. A heat transfer loop using air as an unlimited heat sink consists of a primary heat exchanger in the spent fuel pool water and a secondary heat exchanger located in ambient air. Thus the measurement of the heat flux, which gets transferred from the pool to the ambient air, is an important task. If one would measure heat flux, flow rates and temperatures in many positions by help of local probes, the natural flow would get strongly disturbed. For that reason we introduce a heat flux measurement around the secondary heat exchanger located in ambient air, which applies temperature and velocity measurement by an anemometric principle.A 6.5m long flow channel with an electrical heated finned tube heat exchanger was set up at the TOPFLOW facility at HZDR. Since the tubes of a heat exchanger would be tilted in a passive heat removal system, i.e. to allow drainage of the condensed heat transfer medium, different tiled angles were adjusted to 0° (horizontal), 20°,30° and 40°. The frontal velocity was varied between 0.5ms and 4ms and three thermocouples were placed up- and downstream of the heat exchanger respectively. A Temperature Anemometry Grind Sensor (TAGS) was located downstream the heat exchanger. It consists of a wire grid with platinum resistance elements, which are placed in the small sub-channels of a flow straightener to generate laminar flow profiles. Two methods were used to calculate the heat flux: arithmetical average and weighting of the flow area. The results of velocity was compared with the average velocity measured by the volume flow control and out of the velocity and temperature the heat flux was calculated and compared with electrical supplied heat flux. The calculated average velocity measured by the TAGS corresponds well with the velocity measured by the volume flow controller up to approximately 3ms with a maximum deviation of ±5%, but underestimates the velocity measured by the volume flow controller at higher velocities. The heat flux was calculated by five methods, 1.) from the three thermocouples up- and downstream of the heat exchanger, 2.) from the average temperatures measured by the TAGS, 3.) from the weighted temperature measured by the TAGS, 4.) from the average temperature and velocity measured by the TAGS and 5.) from the weighted temperature and velocity measured by the TAGS. In this order the accuracy of methods increases compared to the electrical supplied heat flux. For the last method the maximum deviation was 6.5% for all tilt angles. This measurement concept determines the heat flux without disturbing the flow in the loop.

Validation of helical steam generator design for the experimental power reactor

The steam generator is an important component in any commercial nuclear power plants to deliver required steam from the reactor heat. Different with the steam generator design of PWR, other design in a small modular reactor (SMR) with smaller thermal capacity incorporates the helical-coiled tubes having advantage of additional heat transfer surface to deliver a superheated steam. The purpose of this research is to model the helical-coiled steam generator in the experimental power reactor developed by BATAN using RELAP5/SCDAP/Mod3.4 code as thermal hydraulic code to obtain a validation of technical data in the conceptual design document. After a number of steady-state simulations, operational thermal hydraulic parameters such as outlet helium and steam temperature, tube pressure drop, and heat capacity have been calculated to be close with the design data. It has been found out, that the different heat transfer characteristics in the helical-coiled tubes need to be addressed specifically in the RELAP5 input data by considering the new estimated heat transfer coefficient or by modifying the heat transfer area. In general, the developed steam generator model is ready to be integrated with the experimental power reactor model that can be used for safety anaysis purposes.

Neutronic performance of fully ceramic microencapsulated of uranium oxycarbide and uranium nitride composite fuel in SMR

The existing commercial nuclear power plants (PWR and BWR) utilize the oxide fuels, i.e., UO2. This fuel selection is not questionable, where the safety and economy are the top priority in the nuclear industry. In this work, the potential advantages of microencapsulated fuels to System Integrated Modular Advanced Reactor “SMART” has been explored. The UN and UCO have been considered as the candidate fuel material for the SMART reactor. The detailed design and fuel assembly configurations with different fuel types of the SMART reactor have been investigated for criticality with depletion (burn up), fuel and moderator temperature coefficients, and power peaking factor using OpenMC. The fissile and fertile fuel elements with coated particles are embedded in graphite matrices. The results are presented with a review of attributes and potential benefits. The nitride fuel has an advantage of mechanical stability, enhanced thermal conductivity, and high fuel density compared to dioxide fuel (UO2). The melting point of standard fuel, UN, and UCO are similar. Thus, they have higher safety margins under NPP's operating conditions. Therefore, nitride (UN) and carbide (UCO) fuel types are more attractive than standard uranium oxides because they are safer and beneficial. The dioxide fuel (UO2) is considered as the reference and compared with candidate material. The neutronic assessment within SMART core specification examined the effective multiplication factor, thermal flux distribution, axial and radial power distribution, and power peaking factor at the beginning, and end of the fuel cycle.

Component detection in piping and instrumentation diagrams of nuclear power plants based on neural networks

Piping and Instrumentation Diagrams (P&IDs) are the most commonly used engineering drawings to describe components and their relationships, and they are one of the most important inputs for data analysis in Nuclear Power Plants (NPP). In traditional analysis, the information related to the components is extracted manually from the P&IDs. This usually takes large amounts of effort and is error-prone. With the rapid development in the area of computer vision and deep learning, automatically detecting components and their relationships becomes possible. In this paper, we aim to use the latest neural network models to automatically extract information on components and their identifications, from the P&IDs in NPPs. We use a Faster Regional Convolutional Neural Networks (Faster RCNN) architecture called ResNet-50, to detect the components in the P&IDs. Compared to common object detection, object detection for P&IDs poses unique challenges to these methods. For example, the P&IDs symbols are much smaller than the background, and detecting such small objects remains a challenging task for modern neural networks. To address these challenges, we 1) propose several techniques for data augmentation that effectively solve the problem of training data shortage, and 2) propose a feature grouping strategy for detecting components with distinct features. Besides, we introduce a SegLink model for text detection, which can automatically extract components’ identifications from P&IDs. We also develop a method for building a data structure to reflect the relationships between components (e.g., to which pipe a component is connected, or what are the downstream or upstream components of one specific component) based on the extracted information. This data structure can be further used for plant safety analysis, and operation and maintenance cost optimization. Sensitivity analysis and comparison with other Convolutional Neural Networks (CNNs) are performed. The results of these analyses are also discussed in this paper.This analysis framework has been tested on the P&IDs from a commercial NPP. The Average Precision for components, which is used to measure the performance of the proposed method, is about 98%. The success rates of component-text mapping and component-pipe mapping are 270/275 and 319/319, respectively. It is worth noting that this framework is generic and can also be applied to P&IDs of non-nuclear industries.

Pyrogenic carbon for decontamination of low-level radioactive effluents: Simultaneous separation of 137Cs and 60Co

Radiocesium (137Cs) and radiocobalt (60Co) represent main dose-significant radionuclides in low-level radioactive liquid effluents released from commercial nuclear power plants (NPPs). Although pyrogenic carbon-based adsorbents (biochars, BC) have been studied for toxic metal removal, the reports of their use in the adsorption of radionuclides are rare. This article focuses on the competitive removal of Cs+ and Co2+ ions from radioactive solutions to waste biomass derived biochar and the multi-component isotherms. BC exhibited a good affinity to both Cs and Co, and in single-component systems the maximal adsorption capacity Qmaxexp reached 97.4 ± 5.0 (Cs+) and 74.5 ± 0.4 μmol g−1 (Co2+). 3D adsorption surfaces obtained by Sheindorf-Rebuhn-Steintuch (SRS) and extended Freundlich models indicate that the sorption of Cs+ ions from the Cs–Co binary system is significantly affected by Co2+ ions and during the adsorption competitive and interaction effects between Cs+ and Co2+ ions occur. Moreover, mono- and divalent cations present in the solution acted as competing ions, decreasing the Kd values. SEM-EDX maps, FTIR and XPS spectra confirmed that both biochar mineral and carbon fraction participated in Cs and Co immobilization. Beside interactions of Cs+ and Co2+ ions with O-containing moieties on the biochar surface (chemisorption), mineral components and porosity of BC (physisorption) were involved in both Cs and Co removal. Finally, the adsorption loading and fast kinetics for both Cs and Co indicate suitability of biochar to decontaminate low-level radioactive effluents contaminated by 137Cs and 60Co.

Syntactic and semantic information extraction from NPP procedures utilizing natural language processing integrated with rules

Procedures play a key role in ensuring safe operation at nuclear power plants (NPPs). Development and maintenance of a large number of procedures reflecting the best knowledge available in all relevant areas is a complex job. This paper introduces a newly developed methodology and the implemented software, called iExtractor, for the extraction of syntactic and semantic information from NPP procedures utilizing natural language processing (NLP)-based technologies. The steps of the iExtractor integrated with sets of rules and an ontology for NPPs are described in detail with examples. Case study results of the iExtractor applied to selected procedures of a U.S. commercial NPP are also introduced. It is shown that the iExtractor can provide overall comprehension of the analyzed procedures and indicate parts of procedures that need improvement. The rich information extracted from procedures could be further utilized as a basis for their enhanced management.

Performance analysis of the second stage accumulators in BNPP-1 in response to MB-LOCA with SBO: A support for practical modification in the injection line

During the last decades, the substitution of passive safety systems instead of active safety systems has increased the reliability of new commercial nuclear power plants. This effort is much more significant for generation (Gen) III + reactors in comparison with Gen II and III reactors. Hydro-accumulators, as a part of the emergency core cooling system (ECCS), is a common passive system among different Gens to encounter the loss of cooling accidents (LOCAs). Recently, the application of advanced and multi-stage accumulators has become successful in GEN III + reactors. Bushehr nuclear power plant (BNPP-1) as a Russian VVER-1000 is somehow categorized between II and III Gens, but it is the first VVER-1000 equipped with two stages of accumulators. This article investigates the influence of Second Stage Accumulators (SSAs) of BNPP-1 on postponing the core damage time in an MB-LOCA with Station Blackout (SBO). Also, SSA's injection line is upgraded after the engineering judgments of start-up tests, and accordingly, each line was equipped with a Flow Restriction Component (FRC). Besides, the effect of FRCs on postponing of the core damage time is assessed in this work. Extracted results achieved from both start-up tests and RELAP5 simulations showed that, although the SSAs are useful in supplying water in MB-LOCA with SBO as a Design Extension Condition (DEC), the application of FRCs are much more effective in postponing the core damage time and without them, SSAs discharge instantly. The results of this study verify the utilization of SSAs with their associated FRCs. Also, the results confirm that some design modifications in BNPP-1 as a commercial reactor can provide resistance against a DECs. Moreover, this experience gained in BNPP-1 could be turned into a basis for safety systems designs in newer VVER-1000s and also VVER-1200 reactors.

Study on Typical Design Basis Conditions of HPR1000 With Nuclear Safety Analysis Code ATHLET

The third-generation nuclear power plant Hua-long Pressurized Reactor (HPR1000) is developed based on the experience of Chinese commercial Nuclear Power Plant (NPP) designs, construction, operation and maintenance. It improves the concept of defense in depth and strengthens severe accident prevention and mitigation strategies. The HPR1000 has implemented a number of active and passive innovative safety systems and accident management procedures for design basis conditions, e.g., the employment of Medium Pressure Rapid Cooldown (MCD) and Atmospheric Steam Dump System (ASDS) for the activation of Middle Head Safety Injection (MHSI), the application of Secondary Passive Residual Heat Removal System (SPRHR) for the residual heat removal. In the article, calculations are carried out for HPR1000 nuclear power plant with nuclear system safety analysis code ATHLET (Analysis of Thermal-Hydraulics of Leaks and Transient) 3.1 [1]. By means of conservative deterministic safety analysis approach, transient analyses concerning selected typical design basis conditions, i.e., Large Break Loss-Of-Coolant Accident (LB-LOCA), Small Break Loss-Of-Coolant Accident (SB-LOCA), Steam Generator Tube Rupture accident (SGTR) and Feed water Line Break (FLB) are performed. The ATHLET results are also compared with the results performed by CGN (China General Nuclear) developed code LOCUST with similar assumptions. The comparisons indicate that, although some discrepancies are detected, the trends of system responses predicted by the two codes are generally in agreement with each other for different accident scenarios. The results also demonstrate that the acceptance criteria for each accident can be met with significant safety margin. Thus, the effectiveness of safety system configuration and accident management procedures is guaranteed.

Clean Fuel from Nuclear Power

Abstract The current commercial hydrogen production has a significant carbon footprint. Now, projects co-funded by the U.S. Department of Energy and commercial nuclear utilities with operating nuclear power facilities aim to change that by exploiting the capabilities of nuclear power plants. This article delves into four projects aimed at demonstrating technology to make hydrogen from water on an industrial scale using energy from an operating commercial nuclear power plant.

A novel experiment for coherent elastic neutrino nucleus scattering: CONUS

The CONUS experiment (COherent elastic NeUtrino nucleus Scattering) aims at detecting coherent elastic neutrino nucleus scattering of reactor antineutrinos on Germanium. The experiment will be set up at the commercial nuclear power plant of Brokdorf, Germany, at a distance of ∼17 m to the reactor core. The recoil of the nuclei hit by the antineutrinos is detected with four high-purity point contact Germanium detectors with a very low threshold and an overall mass of about 4 kg. To suppress the background, the setup is equipped with a shell-like passive shield and an active muon veto system. The shield and the muon veto have successfully been tested at the shallow depth laboratory at Max-Planck-Institut für Kernphysik. Monte Carlo simulations have been performed to reproduce the prompt muon-induced background and to examine the induced neutron spectrum. Currently, the low threshold Germanium detectors are characterized and the experiment is prepared for commissioning.

Improvement in reliability quantification to support BS EN 61511 failure probability analysis

Estimation of failure rates provides a key input to quantitative risk assessment (QRA) quantification. International functional safety standard such as BS EN 61511 specifies use of realistic and credible failure data in failure probability analysis. In traditional reliability assessment, mean time to failure is one of the most common approaches to field failure data analysis. Unfortunately, new technology, such as hydrogen failure data is extremely limited. One possible way is to use surrogate failure data from other settings such as commercial nuclear power plants, chemical plants, and offshore oil and natural gas platforms. The proposed Bayesian framework addresses the requirements by allowing industry knowledge about failure rates to be incorporated in a prior gamma distribution and periodic updating process with new survival data as it becomes available. Monte Carlo simulation is adopted which make it practical to solve uncertainty in the failure rate estimation and update these models with many trials in seconds. The result shows that the process of updating failure rate with more samples of new observations and modelling failure data uncertainty using Monte Carlo simulation can be effective in improving reliability quantifications in the existing BS EN 61511 standard.

Preliminary surface soil area factor for elevated residual radioactivity of Kori Unit 1 considering adjacent Unit 2

Based on the U.S. standard decommissioning procedure guidance manual, Multi-Agency Radiation Site Survey and Investigation Manual (MARSSIM), DCGL (Derived Concentration Guideline Level) can be used to classify contaminated sites and suggests a methodology for assessing site deregulation safety as an essential criterion for determining decontamination activities. In this study, area factors for DCGLEMC, which is DCGL applicable in hot spot, were derived by adjusting the contaminated area at the Kori Unit 1, the first commercial Nuclear Power Plant (NPP) in South Korea. DCGLW and DCGLEMC were investigated for Kori Unit 1 according to changes in preliminary contaminated baseline area including the area of Unit 2. Area factors for surface soil were divided into scenarios, gamma and beta emitters. Sensitive parameters and exposure pathways were identified for each specific contaminated area. Additionally, the trend of DCGLEMC was analyzed in the industrial worker scenario due to baseline area and contaminated area changes.

Neutron-induced background in the CONUS experiment

CONUS is a novel experiment aiming at detecting elastic neutrino–nucleus scattering in the almost fully coherent regime using high-purity germanium (Ge) detectors and a reactor as antineutrino source. The detector setup is installed at the commercial nuclear power plant in Brokdorf, Germany, at a short distance to the reactor core to guarantee a high antineutrino flux. A good understanding of neutron-induced backgrounds is required, as the neutron recoil signals can mimic the predicted neutrino interactions. Especially events correlated with the reactor thermal power are troublesome. On-site measurements revealed such a correlated, highly thermalized neutron field with a maximum fluence rate of (745pm 30),hbox {cm}^{-2},hbox {day}^{-1}. These neutrons, produced inside the reactor core, are reduced by a factor of sim 10^{20} on their way to the CONUS shield. With a high-purity Ge detector without shield the gamma -ray background was examined including thermal power correlated ^{16}hbox {N} decay products and neutron capture gamma -lines. Using the measured neutron spectrum as input, Monte Carlo simulations demonstrated that the thermal power correlated field is successfully mitigated by the CONUS shield. The reactor-induced background contribution in the region of interest is exceeded by the expected signal by at least one order of magnitude assuming a realistic ionization quenching factor.

Study on the Experiences of Subsurface Soil Remediation at Commercial Nuclear Power Plants in the United States

Regulatory agency and licensee are preparing for the site restoration of Kori unit 1, the first commercial NPP in Korea, scheduled for 2031. Developing regulatory guidelines and strategies is essential for effective restoration work. Unfortunately, Korea does not have experience of site restoration of commercial NPPs. Therefore, it is important to review cases from experienced countries to establish a strategy and regulatory standards. The U.S. has had numerous soil remediation experiences using RESRAD and MARSSIM. However, formalized evaluation methodologies for subsurface soil have not yet been established in MARSSIM. This survey focused on subsurface soil remediation by reviewing the five decommissioned NPPs under regulation of the US NRC. Overall process of remediating a contaminated subsurface soil and groundwater was reviewed to identify considerations and lessons that could be applicable in Korea. In addition, an applied methodology for evaluation of contaminated subsurface soil and related major issues between regulatory agency and licensees were reviewed in detail to support establishment of remediation strategy for Kori unit 1.