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Guest Editorial

ASME J of Nuclear Rad Sci. 2018;4(4):040301-040301-1. doi:10.1115/1.4040963.
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The 25th edition of the International Conference on Nuclear Engineering (ICONE-25)1 was held in Shanghai, China in July, 2017, where over 1000 attendees joined the conference. Every year ICONE brings together nuclear engineers, scientists, physicists, and practitioners from Industry, Academia, and Government from around the globe to share information and viewpoints on the latest topics of interest for nuclear power. Technical papers presented during the five days of the conference are second to none technically and cover a wide range of topics that are germane to nuclear-power research, construction, and operation. The technical papers and presentations encompass 16 different technical tracks that include Nuclear Power Plant Operation, Maintenance, Safety, Accident Analysis, Thermal-Hydraulics, Fuel Cycles, Computational Fluid Dynamics, Education, Codes and Standards, and Student Track. The information contained in the over 700 technical papers presented at ICONE-25 is typical of every ICONE and represents some of the finest technical work produced in the Nuclear Engineering field. Included in these tracks is a very robust student-paper track, where more than 60 students from North America, Europe, China, Japan, and Asia are hosted by the three organizing societies (ASME, JSME, and CNS). All of the student works are judged, and best papers and best posters are recognized and awards presented. The top student-paper award is the Akiyama Medal, which was awarded this time to Mr. Nailiang Zhuang from the Harbin Engineering University (China). This part of the conference has been a huge success and continues to grow.

Commentary by Dr. Valentin Fuster

Research Papers

ASME J of Nuclear Rad Sci. 2018;4(4):041001-041001-11. doi:10.1115/1.4040422.

The effect of structural state (solution annealed (SA) and after 40% cold work (CW)) and surface finishing (turning, grinding, and polishing) on the corrosion behavior of austenitic 1.4970 (15-15 Ti) steel in flowing (2 m/s) Pb-Bi eutectic containing 10−7 mass% dissolved oxygen at 400 °C and 10−6 mass% O at 500 °C is investigated. At 400 °C for ∼13,000 h, the corrosion losses are minor for steel in both structural states and for surfaces finished by turning and grinding—a thin Cr-based oxide film is formed. In contrast, the polished surface showed initiation of solution-based corrosion attack with the formation of iron crystallites and preferential propagation along the grain boundaries. The depth of corrosion attack does not exceed 10 μm after ∼13,000 h. At 500 °C for 2000 h, the samples in both structural states showed general slight oxidation. Cold-worked steel underwent a severe groove-type and pit-type solution-based attack of 170 μm in maximum depth, while the SA sample showed only sporadic pit-type corrosion attack to the depth of 45 μm in maximum.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(4):041002-041002-9. doi:10.1115/1.4040887.

Accident tolerant fuels (ATF) and steam generator (SG) auxiliary feedwater (AFW) extended operation are two important methods to increase the coping time for nuclear power plant safety response. In light of recent efforts to investigate such methods, we investigate both FeCrAl cladding oxidation kinetics and SG AFW sensitivity analyses, for the Surry nuclear power plant Short-Term Station Blackout simulation using the MELCOR YR 1.8.6 systems code. The first part describes the effects of FeCrAl cladding oxidation kinetics. Zircaloy cladding and two different oxidation models of FeCrAl cladding are compared. The initial hydrogen generation time (>0.5 kg) is used as the evaluation criterion for fuel degradation in a severe accident. Results showed that the more recent oxidation correlation by ORNL predicts much less hydrogen generation than Zircaloy cladding. The second part investigates the effects of three different methods of AFW injection into the SG secondary side. We considered three different methods of water injection; i.e., constant water injection into the secondary side (case 1); water injection based on secondary side water level in boiler region (case 2); water injection based on secondary side water level in the downcomer region (case 3). The case of constant water injection is the most straightforward, but it would have the tendency to overfill the SG with excess water. Water injection with downcomer level control is more reasonable but requires DC power to monitor level and to control AFW injection rate.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(4):041003-041003-8. doi:10.1115/1.4039971.

Use of FCM fuel in light water reactors is an attractive option for existing and future generations of these reactors to make them accident tolerant in nature. This work focuses on the neutronic study of the use of burnable material in various configurations to control the excess reactivity and to keep the moderator temperature coefficient of reactivity (MTC) feedback negative for entire cycle length. Erbia and gadolinia, two conventional materials are used in three different configurations including quadruple isotropic (QUADRISO), bi-isotropic (BISO), and Matrix Mix forms. The results obtained from the implicit random treatment of the double heterogeneity of tri-structural isotropic (TRISO), QUADRISO, and BISO particles show that the erbia is the best material to be used in QUADRISO and Matrix Mix configurations with lowest reactivity swing for the life cycle and residual poison well below 0.5%. Gadolinia is usable in FCM environment only in the BISO form where enhanced self-shielding controls the depletion performance of the material. The gadolinia has almost zero residual poison at end of cycle (EOC); however, it has relatively large reactivity swing, which will need more micromanagement of the control rods during the plant operations. At the beginning of cycle (BOC), erbia-loaded assemblies have shown an increase in negative value of MTC compared with reference due to presence of resonance peak in erbium near 1 eV. The finally recommended material-configuration combinations have shown the excess reactivity containment in desired manner with good depletion performance and negative feedback of the MTC for life cycle.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(4):041004-041004-7. doi:10.1115/1.4040432.

A new Japanese nuclear regulation involves estimating the possible damage to plant structures due to intentional aircraft impact. The effect of aircraft impact needs to be considered in the existing nuclear power plants. The structural damage and fuel dispersion behavior after aircraft impact into plant structures can be evaluated using finite element analysis (FEA). FEA needs validated experimental data to determine the reliability of the results. In this study, an analysis method was validated using a simple model such as a cylindrical tank. Numerical simulations were conducted to evaluate the impact and dispersion behavior of a water-filled cylindrical tank. The simulated results were compared with the test results of the VTT Technical Research Centre of Finland (VTT). The simulations were carried out using a multipurpose FEA code LS-DYNA®. The cylindrical tank was modeled using a shell element, and the tank water was modeled using smoothed particle hydrodynamics (SPH) elements. First, two analysis models were used to evaluate the effect of the number of SPH elements. One had about 300,000 SPH elements and the other had 37,000 SPH elements. The cylindrical tank ruptured in the longitudinal direction after crashing into a rigid wall, and the filled water dispersed. There were few differences in the simulated results when using different numbers of SPH elements. The VTT impact test was simulated with an arbitrary Lagrangian-Eulerian (ALE) element to consider the air drag. The analytical dispersion pattern and history of dispersion velocity ratio agreed well with the impact test results.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(4):041005-041005-6. doi:10.1115/1.4039967.

Shanghai Nuclear Engineering Research and Design Institute (SNERDI) has been studying seismic risk analysis for nuclear power plant for a long time, and completed seismic margin analysis for several plants. After Fukushima accident, seismic risk has drawn an increasing attention worldwide, and the regulatory body in China has also required the utilities to conduct a detailed analysis for seismic risk. So, we turned our focus on a more intensive study of seismic probabilistic safety assessment (PSA/PRA) for nuclear power plant in recent years. Since quantification of seismic risk is a key part in Seismic PSA, lots of efforts have been devoted to its research by SNERDI. The quantification tool is the main product of this research, and will be discussed in detail in this paper. First, a brief introduction to Seismic PSA quantification methodology is presented in this paper, including fragility analysis on system or plant level, convolution of seismic hazard curves and fragility curves, and uncertainty analysis as well. To derive more accurate quantification results, the binary decision diagram (BDD) algorithm was introduced into the quantification process, which effectively reduces the deficiency of the conventional method on coping with large probability events and negated logic. Finally, this paper introduced the development of the seismic PSA quantification tool based on the algorithms discussed in this paper. Tests and application have been made for this software based on a specific nuclear power plant seismic PSA model.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(4):041006-041006-7. doi:10.1115/1.4040372.

Companies involved in the nuclear energy domain, like component and platform manufacturers, system integrators, and utilities, have well-established yearly trainings on Nuclear Safety Culture. These trainings are typically covered as part of the annual quality assurance-related refresher trainings, introductory courses for new employees, or indoctrinations of temporary staff. Gradually, security awareness trainings are also addressed on a regular basis, typically with a focus on information technology, the daily office work, test bay, or construction site work environment, and some data protection and privacy-related topics. Due to emerging national nuclear regulation, steadily but surely, specialized cybersecurity trainings are foreseen for integrators and utilities. Beyond these safety, physical security and cybersecurity specific trainings, there is a need to address the joint part of these disciplines, starting from the planning phase of a new nuclear power plant (NPP). The engineers working on safety, physical protection, and cybersecurity must be aware of these interrelations to jointly elaborate a robust instrumentation and control architecture (defense-in-depth, design basis events, functional categorization and systems classification) and a resilient security architecture (security by design, security grading, zone model or infrastructure domain, security conduits, forensic readiness, security information, and event management). This paper provides more in-depth justification of when and where additional training is needed, due to the ubiquitous deployment of digital technology in new NPPs. Additionally, for existing NPPs, the benefits of conveying knowledge by training on specific interfaces between the involved disciplines will be discussed. Furthermore, the paper will address the need of focused training of management stakeholders, as eventually, they must agree on the residual risk. The decision-makers are in charge of facilitating the interdisciplinary cooperation in parallel to the allocation of resources, e.g., on security certifications of products, extended modeling-based safety and security analyses and security testing coverage.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(4):041007-041007-6. doi:10.1115/1.4039969.

In order to study the contribution of manganese (Mn) atoms in copper (Cu) precipitates to hardening in body centered cubic (BCC) structure iron (Fe) matrix, the interactions of a 1/2 〈111〉 {110} edge dislocations with nanosized Cu and Cu–Mn precipitates in BCC Fe have been investigated by using molecular dynamics method (MD). The results indicate that the critical resolved shear stresses (τc) of the Cu–Mn precipitates are larger than that of Cu precipitates. Meanwhile, τc of the Cu–Mn precipitates show a much more significant dependence on temperature and size compared to Cu precipitates. Mn atoms exhibit strong attraction to dislocation segment in Cu precipitate and improve the fraction of transformed atoms from BCC phase to nine rhombohedron (R) phase for big size precipitates. Those all lead to the higher resistance to the dislocation glide. Eventually, these features confirmed that the appearance of Mn atoms in Cu precipitates greatly facilitates the hardening in BCC Fe matrix.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(4):041008-041008-9. doi:10.1115/1.4040650.

Cybersecurity incidents are stressful, complex in nature, and are frequently not systematically considered in daily tasks. When correctly managed, operational readiness procedures ensure the availability of data required to successfully and quickly recover from a security incident, while lessening the adverse effect. Therefore, protective measures, such as implementation of data diodes, are playing an essential role in defending instrumentation and control (I&C) systems. In addition, applicability of the newest forensic and digital evidence-related standards to the nuclear domain is being evaluated. Results of such evaluation are being considered in the three-dimensional and two-dimensional modeling of cybersecurity relevant assets. The development of the new IEC 63096, downstream standard of IEC 62645, will also support the proposed evaluation and modeling. However, IEC 63096 covers not only forensic and incident management-related security controls but also a broad range of cybersecurity controls. This paper will further explore the security degree-specific selection and overall assignment of forensic-related security controls for the nuclear domain. Results from ongoing prototype developments will be used to demonstrate possible alternative selections and assignments, along with their contribution to different security metrics.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(4):041009-041009-6. doi:10.1115/1.4040368.

Accumulative test data indicate that the effects of the light water reactor (LWR) environment could cause the fatigue resistance of primary pressure boundary components materials to be significantly reduced. Environmentally assisted fatigue (EAF) is the abbreviation of the environmentally assisted fatigue. In 2007, Nuclear Regulatory Commission (NRC) issued RG. 1.207. It was updated in 2014. And, it requires that the effects of LWR environment on the fatigue life reduction of metal components should be considered for new design plants. And it suggests to use environmental correction factor, Fen, to account for EAF. NRC regulation (NUREG), NUREG/CR-6909 (NRC, 2013, “Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials,” U.S. Nuclear Regulatory Commission, Argonne, IL, Standard no. NUREG/CR-6909), presents the detail Fen calculation formula. Fen is a function of temperature, strain rate, dissolved oxygen level in water, and sulfur content of the steel. Accordingly, Fen calculation will present a comparatively conservative result. Depending on the experience of the primary pressure boundary piping transient operation, Fen varies during each transient. More uncertainty and confusion are raised during the application of the Fen method. The research work in this paper includes: first, the typical character of piping thermal transient is derived based on the existing experience. Second, small specimen EAF tests are conducted depending on the above derived combined loading characters. Then effort is taken to improve the application of the Fen method for the combined multitransient loading conditions. And the results are compared with those of the lowest instantaneous Fen method and equalization of the weighted Fen method. Finally, a designed test plan is presented.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(4):041010-041010-6. doi:10.1115/1.4039970.

The generic concept of security controls, as initially deployed in the information security domain, is gradually used in other business domains, including industrial security for critical infrastructure and cybersecurity of nuclear safety instrumentation & control (I&C). A security control, or less formally, a security countermeasure can be any organizational, technical, or administrative measure that helps in reducing the risk imposed by a cybersecurity threat. The new IAEA NST036 lists more than 200 such countermeasures. NIST SP800-53 Revision 4 contains about 450 pages of security countermeasure descriptions, which are graded according to three levels of stringency. In order to facilitate and formalize the process of developing, precisely describing, distributing, and maintaining more complex security controls, the application security controls (ASC) concept is introduced by the new ISO/IEC 27034 multipart standard. An ASC is an extensible semiformal representation of a security control (extensible markup language or javascript object notation-based), which contains a set of mandatory and optional parts as well as possible links to other ASCs. A set of ASCs may be developed by one company and shipped together with a product of another company. ISO/IEC 27034-6 assumes that ASCs are developed by an organization or team specialized in security and that the ASCs are forwarded to customers for direct use or for integration into their own products or services. The distribution of ASCs is supported and formalized by the organization normative frameworks (ONFs) and application normative frameworks (ANFs) deployed in the respective organizational units. The maintenance and continuous improvement of ASCs is facilitated by the ONF process and ANF process. This paper will explore the applicability of these industry standards based ASC lifecycle concepts for the nuclear domain in line with IEC 62645, IEC 62859, and the upcoming IEC 63096. It will include results from an ongoing bachelor thesis and master thesis, mentored by two of the authors, as well as nuclear-specific deployment scenarios currently being evaluated by a team of cybersecurity Ph.D. candidates.

Topics: Security , Maintenance
Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(4):041011-041011-5. doi:10.1115/1.4040888.

The dynamic reliability of the main and startup feedwater control system in nuclear power plant is evaluated by conducting the Markov cell-to-cell mapping technology (Markov/CCMT) methodology. All the equipment failure modes and potential failure states within the system are analyzed. This process illustrates the uncertainty in the physical process of the system. Furthermore, the failure probability and cut-set of the system can be computed to provide a more comprehensive and accurate response to the system characteristics and reflect the two types of interaction within the system. In contrast to the traditional static probability safety assessment, the Markov/CCMT methodology remedies the defect in terms of event sequence setting, control loop, multiple top event competition, uncertainty of the analysis result, as well as the insufficient analysis of human-caused failure. The reliability analysis of the main and startup feedwater control system (FWCS) based on the self-developed Markov/CCMT reliability analysis software verifies the feasibility and engineering application value of the methodology and software.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(4):041012-041012-5. doi:10.1115/1.4040367.

Spent fuel pool (SFP) stores fuel assemblies removed from the reactor over the years. SFP and its accident mitigation measures may fail simultaneously at the time of the earthquake, which may cause serious accident consequences. This paper uses probabilistic safety assessment (PSA) method to quantitatively evaluate the risk of SFP for a CPR1000 unit caused by seismic events. Quantitative analysis results show that seismic events' risk is the highest in all internal events and external events for SFP. In order to reduce the risk of SFP, more attention should be paid to improve seismic capacity or reduce the common failure for systems and components associated with SFP under the earthquake situation.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(4):041013-041013-6. doi:10.1115/1.4040369.

The control function for process, heating, ventilation, and air conditioning, and electrical systems in nuclear power plant (NPP) are represented by control logic diagram. To develop distributed control system (DCS), the designer and supplier should complete the activities of control logic configuration, testing, and verification, which are based on control logic diagram. Design verification is an effective method to ensure the correctness of control logic design. This paper represents a system, which is capable of implementing control logic design verification automatically for NPP instrumentation and control (I&C) system, as well as an overview of the procedure and some examples by using this system. With the design data (including control requirements and control logic diagrams in computer-readable format) and simulation technology, this system automatically performs design verification based on different rules and confirms the design outputs meet the inputs—the control requirements of plant's systems. Finally, a conclusion about the design verification system and future scenarios is given.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(4):041014-041014-8. doi:10.1115/1.4040371.

The intercooled cycle (IC) is a simplified novel proposal for generation IV nuclear power plants (NPP) based on studies demonstrating efficiencies of over 45%. As an alternative to the simple cycle recuperated (SCR) and the intercooled cycle recuperated (ICR), the main difference in configuration is no recuperator, which reduces its size. It is expected that the components of the IC will not operate at optimum part power due to seasonal changes in ambient temperature and grid prioritization for renewable sources. Thus, the ability to demonstrate viable part load performance becomes an important requirement. The main objective of this study is to derive off-design points (ODPs) for a temperature range of −35 °C to 50 °C and core outlet temperatures (COTs) between 750 °C and 1000 °C. The ODPs have been calculated using a tool designed for this study. Based on the results, the intercooler changes the mass flow rate and compressor pressure ratio (PR). However, a drop of ∼9% in plant efficiency, in comparison to the ICR (6%) was observed for pressure losses of up to 5%. The reactor pressure losses for IC have the lowest effect on plant cycle efficiency in comparison to the SCR and ICR. Characteristic maps are created to support first-order calculations. It is also proposed to consider the intercooler pressure loss as a handle for ODP performance. The analyses brings attention to the IC an alternative cycle and aids development of cycles for generation IV NPPs specifically gas-cooled fast reactors (GFRs) and very-high-temperature reactors (VHTRs), using helium.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(4):041015-041015-9. doi:10.1115/1.4040494.

The so-called in-vessel retention (IVR) was considered as a severe accident management strategy and had been certified by Nuclear Regulatory Commission (NRC) in U.S. as a standard measure for severe accident management since 1996. In the core meltdown accident, the reactor pressure vessel (RPV) integrity should be ensured during the prescribed time of 72 h. However, in traditional concept of IVR, several factors that affect the RPV failure were not considered in the structural safety assessment, including the effect of corium crust on the RPV failure. Actually, the crust strength is of significant importance in the context of a severe reactor accident in which molten core material melts through the reactor vessel and collects on the lower head (LH) of the RPV. Consequently, the RPV integrity is significantly influenced by the crust. A strong, coherent crust anchored to the RPV walls could allow the yet-molten corium to fall away from the crust as it erodes the RPV, therefore thermally decoupling the melt pool from the coolant and sharply reducing the cooling rate. Due to the thermal resistance of the crust layer, it somewhat prevents further attack of melt pool from the RPV. In the present study, the effect of crust on RPV structural behaviors was examined under multilayered crust formation conditions with consideration of detailed thermal characteristics, such as high-temperature gradient across the wall thickness. Thereafter, systematic finite element analyses and subsequent damage evaluation with varying parameters were performed on a representative RPV to figure out the possibility of high temperature induced failures with the effect of crust layer.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(4):041016-041016-7. doi:10.1115/1.4040364.

Plants exhibit complex responses to change in environmental conditions such as radiant heat flux, water quality, airborne pollutants, and soil contents. We seek to utilize natural chemical and electrophysiological response of plants to develop novel plant-based sensor networks. Our present work focuses on plant responses to nuclear radiation—with the goal of monitoring plant responses as benchmarks for detection and dosimetry. In our study, we used plants including Cactus, Arabidopsis, Dwarf mango (pine), Euymus, and Azela. We demonstrated that these plants Chlorophyll-a (F680) to Chlorophyll-b (F735) ratio can be changed according to the radiation dose amount. The recovery processes and speed are different for different plants.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(4):041017-041017-9. doi:10.1115/1.4040423.

The management of long-lived radionuclides in spent fuel is a key issue to achieve the closed nuclear fuel cycle and the sustainable development of nuclear energy. The partitioning-transmutation method is supposed to efficiently treat the long-lived radionuclides. Accordingly, the transmutation of long-lived minor actinides (MAs) is significant for the postprocessing of spent fuel. In the present work, the transmutations in pressurized water reactor (PWR) mixed oxide (MOX) fuel are investigated through the Monte Carlo neutron transport method. Two types of MAs are homogeneously incorporated into MOX fuel assembly with different mixing ratios. In addition, two types of design of semihomogeneous loading of 237Np in MOX fuels are studied. The results indicate an overall nice efficiency of transmutation in PWR with MOX fuel, especially for 237Np and 241Am, which are primarily generated in the current uranium oxide fuel. In addition, the transmutation efficiency of 237Np is excellent, while its inclusion has no much influence on other MAs. The flattening of power and burnup are achieved by semihomogeneous loading of MAs. The uncertainties of Monte Carlo method are negligible, while those due to nuclear data change little the conclusions of the transmutation of MAs. The transmutation of MAs in MOX fuel is expected to be an efficient method for spent fuel management.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(4):041018-041018-6. doi:10.1115/1.4040029.

As the first fast reactor of China, the safety of China Experimental Fast Reactor (CEFR) is extremely important, and will decide the future of Chinese fast reactor project. The fuel failure detection system of CEFR provides surveillance and protection for the first barrier-fuel cladding of CEFR, so it is one of the most important systems for the safety of CEFR. As tag gas method is an important method for fuel-failure location in fast reactor, CEFR has a medium-term and long-term plan of using this method to locating failed fuel assemblies. This paper introduces the main principle of tag gas method, summarizes the application of this method, and compares the advantages and disadvantages of each fuel failure location method. Combining the design characteristics of CEFR, this work analyzes the selection principle of tag gas isotopes and the effects on heat transfer capability of fuel element while tag gas filled in. Meanwhile, according to the detection ability of mass spectrometer and the foreign advanced utilization experiences of tag gas method, some suggestions are provided.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(4):041019-041019-9. doi:10.1115/1.4040370.

This paper presents a power management and distribution system (PMAD) for a growing Martian colony. The colony is designed for a 15-year operation lifetime, and will accommodate a population that grows from 6 to 126 crewmembers. To provide sufficient power, a nuclear fission surface power (FSP) system is proposed with a total capacity of 1 MWel. The system consists of three 333 kWel fission reactors. Direct current (DC) transmission with 2000 voltage direct current (VDC) is found to provide the best power density and transmission efficiency for the given configuration. The grounding system consists of grounding rods, grounding grids, and a soil-enhancement plan. A regenerative fuel cell using a propellant tank recycled from the lander was found to have the best energy density and scalability among all the options investigated. The thermal energy reservoir, while having the worst storage efficiency, can be constructed through in situ resource utilization (ISRU), and is a promising long-term option. A daily load following a 12-h cycle can be achieved, and the power variation will be less than 10% during normal operation. Several main load-following scenarios are studied and accommodated, including an extended dust storm, nighttime, daytime, and transient peak power operation. A contingency power operation budget is also considered in the event that all of the reactors fail. The system has a power distribution efficiency of 85%, a storage efficiency of 50%, and a total mass of 13 Mt.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(4):041020-041020-7. doi:10.1115/1.4040936.

The continuous generation of graphite dust particles in the core of a high-temperature reactor (HTR) is one of the key challenges of safety during its operation. The graphite dust particles emerge from relative movements between the fuel elements or from contact to the graphitic reflector structure and could be contaminated by diffused fission products from the fuel elements. They are distributed from the reactor core to the entire reactor coolant system. In case of a depressurization accident, a release of the contaminated dust into the confinement is possible. In addition, the contaminated graphite dust can decrease the life cycle of the coolant system due to chemical interactions. On one hand, the knowledge of the behavior of graphite dust particles under HTR conditions using helium as the flow medium is a key factor to develop an effective filter system for the discussed issue. On the other hand, it also provides a possibility to access the activity distribution in the reactor. The behavior can be subdivided into short-term effects like transport, deposition, remobilization and long-term effects like reactions with material surfaces. The Technische Universität Dresden has installed a new high-temperature test facility to study the short-term effects of deposition of graphite dust particles. The flow channel has a length of 5 m and a tube diameter of 0.05 m. With helium as the flow medium, the temperature can be up to 950 °C in the channel center and 120 °C on the sample surface, the Reynolds number can be varied from 150 up to 1000. The particles get dispersed into the accelerated and heated flow medium in the flow channel. Next, the aerosol is passing a 3 m long adiabatic section to ensure homogenous flow conditions. After passing the flow straightener, it enters the optically accessible measurement path made from quartz glass. In particular, this test facility offers the possibility to analyze the influence of the thermophoretic effect separately. For this, an optionally cooled sample can be placed in the measuring area. The thickness of the particle layer on the sample is estimated with a three-dimensional laser scanning microscope. The particle concentration above the sample is measured with an aerosol particle sizer (APS). Particle image velocimetry (PIV) detects the flow-velocity field and provides data to estimate the shear velocity. In combination with the measured temperature-field, all necessary information for the calculation of the particle deposition and particle relaxation times are available. The measurements are compared to results of theoretical works from the literature. The experimental database is relevant especially for computational fluid dynamics (CFD)-developers, for model development, and model verification. A wide range of phenomena like particle separation, local agglomeration of particles with a specific particle mass, and selective remobilization can be explained in this way. Thus, this work contributes to a realistic analysis of nuclear safety.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(4):041021-041021-7. doi:10.1115/1.4039968.

In this study, the applicability of Monte Carlo code particle and heavy ion transport code system (PHITS) [Sato et al. (2013, “Particle and Heavy Ion Transport Code System PHITS, Version 2.52,” J. Nucl. Sci. Technol., 50(9), pp. 913–923)] to the equipment design of sampler and detector in the radiation monitoring system was evaluated by comparing calculation results with experimental results obtained by actual measurements of radioactive materials. In modeling a simulation configuration, reproducing the energy distribution of beta-ray emitted from specific nuclide by means of Fermi Function was performed as well as geometric arrangement of the detector in the sampler volume. The reproducing and geometric arrangement proved that the calculation results are in excellent matching with actual experimental results. Moreover, reproducing the Gaussian energy distribution to the radiation energy deposition was performed according to experimental results obtained by the multi-channel analyzer. Through the modeling and the Monte Carlo simulation, key parameters for equipment design were identified and evaluated. Based on the results, it was confirmed that the Monte Carlo simulation is capable of supporting the evaluation of the equipment design.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(4):041022-041022-4. doi:10.1115/1.4040495.

The sputtering of graphite due to the bombardment of hydrogen isotopes is crucial to successfully using graphite in the fusion environment. In this work, we use molecular dynamics to simulate the sputtering using the large-scale atomic/molecular massively parallel simulator (lammps). The calculation results show that the peak values of the sputtering yield are between 25 eV and 50 eV. When the incident energy is greater than the energy corresponding to the peak value, a lower carbon sputtering yield is obtained. The temperature that is most likely to sputter is approximately 800 K for hydrogen, deuterium, and tritium. Below the 800 K, the sputtering yields increase with temperature. By contrast, above the 800 K, the yields decrease with increasing temperature. Under the same temperature and incident energy, the sputtering rate of tritium is greater than that of deuterium, which in turn is greater than that of hydrogen. When the incident energy is 25 eV, the sputtering yield at 300 K increases below an incident angle at 30 deg and remains steady after that.

Commentary by Dr. Valentin Fuster

Technical Brief

ASME J of Nuclear Rad Sci. 2018;4(4):044501-044501-4. doi:10.1115/1.4040366.

In an emergency condition of nuclear power plant, operators have to mitigate the accident in order to remove the decay heat and to prevent the release of radioactive material to the environment following the emergency operating procedures (EOPs). The action of operators on a component, for example, changing the parameter level of a component, which is described in a procedure step, will impact other components of the plant and the plant behavior. Nowadays, the advanced main control rooms have been equipped with the computer-based procedures (CBPs) which provide some features and benefits which are not available in paper-based procedures. However, most of CBPs do not provide information of the impact of the counteractions on each procedure step (components influenced and future plant behavior) although it is useful for operators to understand the purpose of the procedure steps before making decisions and taking the actions. This paper discusses the functional information and the method to generate the information using multilevel flow modeling (MFM) model of operator actions on some procedure steps of a simplified EOP of pressurized water reactor (PWR) plant, as an example.

Commentary by Dr. Valentin Fuster

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