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### Editorial

ASME J of Nuclear Rad Sci. 2019;5(1):010201-010201-3. doi:10.1115/1.4041694.
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The ASME Journal of Nuclear Engineering and Radiation Science (NERS) celebrates its fourth year. On this occasion, I would like to congratulate all members of our Journal Board, reviewers listed in the following editorial, authors, and readers on this occasion! Below are the latest Journal statistics (also, see Table 1), changes, and plans for the future. First of all, I would like to express my great appreciation to those Associate and Guest Editors (AEs and GEs), whose terms have expired in 2018, for their valuable contributions to the Journal and nuclear engineering in general.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(1):010202-010202-2. doi:10.1115/1.4041823.
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The Reviewers of the Year Award is given to reviewers who have made an outstanding contribution to the journal in terms of the quantity, quality, and turnaround time of reviews completed during the past 12 months. The prize includes a Wall Plaque, 50 free downloads from the ASME Digital Collection, and a one year free subscription to the journal.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(1):010203-010203-1. doi:10.1115/1.4041695.
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Clean energy technologies appear to be in the spotlight of recent discussions of policy and opinion makers all around the world. Although these discussions do not always appear very technical, they do provide opportunities to show the way toward scientific and technically sound solutions benefiting all people on the planet. These opportunities are definitely extended toward ASME (which promotes the art, science, and practice of multidisciplinary engineering and allied sciences around the globe) in general and the ASME Nuclear Engineering Division (NED, which focuses on the design, analysis, development, testing, operation and maintenance of reactor systems and components, nuclear fusion, heat transport, nuclear fuels technology, and radioactive waste) in particular. Yes, nuclear energy is a clean energy. Probably, the most abundant of all that is available to humanity today.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(1):010204-010204-1. doi:10.1115/1.4041888.
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The Chinese Nuclear Society (CNS) congratulates readers, reviewers, and editors of the ASME Journal of Nuclear Engineering and Radiation Science (ASME J. NERS) and nuclear experts and specialists all around the world with significant achievements and new milestones in nuclear-power industry!

Commentary by Dr. Valentin Fuster

### Guest Editorial

ASME J of Nuclear Rad Sci. 2019;5(1):010301-010301-2. doi:10.1115/1.4041822.
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It is my pleasure to celebrate the fifth anniversary of the ASME Journal of Nuclear Engineering and Radiation Science (NERS), as a former chair of Division of Power and Energy Systems (PES), Japan Society of Mechanical Engineers (JSME). The NERS is a very important journal for nuclear researchers and engineers.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(1):010302-010302-3. doi:10.1115/1.4041973.
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The AP1000® (Advanced Plant) is a 1100-MWel-class pressurized water reactor with innovative passive-safety features and extensive plant simplifications that enhance construction, operation, maintenance, and safety. The passive-safety features of the plant use natural driving forces such as pressurized gas, gravity flow, natural circulation flow, and natural convection. These features do not rely on active components (such as pumps, fans, or diesel generators) or safety-grade support systems (such as alternative current (AC) power, component cooling water, service water, and heating, ventilating, and air conditioning heating, ventilating, and air conditioning (HVAC)).

Commentary by Dr. Valentin Fuster

### Research Papers

ASME J of Nuclear Rad Sci. 2019;5(1):011001-011001-13. doi:10.1115/1.4040649.

Local subassembly faults (LFs) have been considered to be of greater importance in safety evaluation in sodium-cooled fast reactors (SFRs) because fuel elements were generally densely arranged in the subassemblies (SAs) in this type of reactors, and because power densities were higher compared with those in light water reactors. A hypothetical total instantaneous flow blockage (HTIB) at the coolant inlet of an SA gives most severe consequences among a variety of LFs. Although an evaluation on the consequences of HTIB using SAS4A code was performed in the past study, SAS4A code was further developed by implementing analytical model of power control system in this study. An evaluation on the consequences of HTIB in an SFR by this developed SAS4A code was also performed in this study. It was clarified by the analyses considering power control system that the reactor would be safely shut down by the reactor protection system triggered by either of 116% over power or delayed neutron detector (DND) trip signals. Therefore, the conclusion in the past study that the consequences of HTIB would be much less severe than that of unprotected-loss-of-flow (ULOF) was strongly supported by this study. Furthermore, SAS4A code was newly validated using four in-pile experiments which simulated HTIB events. The validity of SAS4A application to safety evaluation on the consequence of HTIB was further enhanced in this study. Thus, the methodology of HTIB evaluation was established in this study together with the past study and is applicable to HTIB evaluations in other SFRs.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(1):011002-011002-15. doi:10.1115/1.4041269.

Safety analyses at the high flux isotope reactor (HFIR) are required to qualify experiment targets for the production of plutonium-238 (238Pu) from neptunium dioxide/aluminum cermet (NpO2/Al) pellets. High heat generation rates (HGRs) due to fissile material and low melting temperatures require a sophisticated set of steady-state thermal simulations in order to ensure sufficient safety margins. These simulations are achieved in a fully coupled thermo-mechanical analysis using comsolmultiphysics for four different preliminary target designs using an evolving set of pre- and postirradiation data inputs, and subsequently evolving solution scopes, from the unique pellet and target designs. A new comprehensive presentation of these preliminary analyses is given and revisited analyses of the first prototypical target designs are presented to reveal the effectiveness of evolving methods and input data.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(1):011003-011003-8. doi:10.1115/1.4041272.

Molten salt reactor (MSR), the only one using liquid fuel in the six types “Generation IV” reactors, is very different from reactors in operation now and has initiated very extensive interests all over the world. This paper is primarily aimed at investigating the breeding characteristics of high-power thorium molten salt reactor (TMSR) based on the two-fluid molten salt breeder reactor (MSBR) with a superior breeding performance. We explored the optimized structure to be a thorium-based molten salt breeder reactor with different core conditions and different postprocessing programs, and finally got the breeding ratio of 1.065 in our TMSR model. At last we analyzed the transient security of our optimized model with results show that the temperature coefficient of core is −3 pcm/K and a 2000 pcm reactivity insertion can be successfully absorbed by the core if the insertion time is more than or equal to 5 s and the core behaves safely.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(1):011004-011004-12. doi:10.1115/1.4041791.

Advanced heavy water reactor (AHWR) is a natural circulation-based, light water-cooled, heavy water-moderated pressure tube type of nuclear reactor. In AHWR, the steam separation takes place in horizontal steam drums purely based on gravity separation principle. It is a known fact that efficiency of gravity separation is affected by the carryover phenomenon, i.e., conveyance of water droplets by the steam. To minimize the carryover, it is advised to reduce the superficial velocities of phases. Lowering the flow velocities also results in to lower pressure drop, which is very much desired. However, careful attention must be given to carryover phenomenon during design. An experimental test facility known as air–water loop (AWL) simulating the scaled down steam drum of AHWR with air–water mixture has been designed and experimental work performed on carryover phenomenon is presented here. Comparison of measured entrainment fraction with existing correlations and other visual observations are described. Numerical simulations with Euler–Lagrangian method have been carried out for which droplet size distribution measured experimentally is used as an input.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(1):011005-011005-12. doi:10.1115/1.4039778.

Thorium-based fuel cycles can improve fuel sustainability within the nuclear power industry. The Canadian supercritical water-cooled reactor (SCWR) concept uses this path to achieve the sustainability requirement of the Gen-IV Forum. The study of thorium dioxide/thoria ThO2-based fuel irradiation behavior is significantly less advanced than that of uranium dioxide (UO2) fuel, although ThO2 possesses superior thermal conductivity, thermal expansion, higher melting temperature, and oxidation resistance that may improve both fuel performance and safety. The fuel and sheath modeling tool (FAST), a fuel performance model for UO2 fuel, was developed at the Royal Military College of Canada (RMCC). FAST capability has been extended to include thoria (ThO2), thorium uranium dioxide (Th,U)O2, and thorium plutonium dioxide (Th,Pu)O2 as fuel pellet materials, to aid in designing and performance assessment of Th-based fuels, including SCWR (Th,Pu)O2 fuel. The development and integration of ThO2 and (Th,U)O2 models into the existing FAST model led to the multipellet material FAST (MPM-FAST). Model development was performed in collaboration between RMCC and Canadian Nuclear Laboratories (CNL). This paper presents an outline of the ThO2 and (Th,U)O2 MPM-FAST model, a comparison between modeling results with postirradiation examination (PIE) data from a test conducted at CNL, and an account of the knowledge gap between our ability to model ThO2 and (Th,U)O2 fuel compared to UO2. Results are encouraging when compared to PIE data.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(1):011006-011006-12. doi:10.1115/1.4040937.

Specimens produced from technically pure iron and two different heats of ferritic/martensitic steel T91 are investigated after exposure to oxygen-containing flowing lead–bismuth eutectic (LBE) at 400 °C, 10−7 mass% dissolved oxygen, and flow velocity of 2 m/s, for exposure times between around 1000 and 13,000 h. The occurring phenomena are analyzed and quantified using metallographic cross sections prepared after exposure. While pure iron mostly shows solution underneath or in the absence of a detached and buckled oxide scale, solution in T91 occurs only in a few spots on the sample surface. However, in the case of one of the investigated heats, a singular event of exceptionally severe solution-based corrosion is observed. The results are compared especially with findings at 450 and 550 °C and otherwise similar conditions as well as austenitic steels tested in the identical experimental run.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(1):011007-011007-12. doi:10.1115/1.4040651.

This paper analyzes and reviews various seismic damage at the Fukushima Daiichi Nuclear Power Station (F1) caused by the Great East Japan Earthquake (the Earthquake) on March 11 in 2011. Moreover, descriptions of various F1 accident reports on on-site seismic damage are comparatively analyzed. At first, impacts of the Earthquake and tsunami as well as damage at four influenced Nuclear Power Stations (NPSs), including F1, are comparatively analyzed. Although no safety-related equipment were seismically damaged at F1, there occurred various on-site seismic damage which should be learned at NPSs worldwide in preparation for next possible beyond-design-basis disasters. Particularly damaged were the main administration building as well as various on-site high-voltage equipment to receive off-site electric power. Owing largely to this on-site damage F1 lost the off-site power, eventually leading the entire NPS to station black-out. Other on-site seismic damage includes loss of an emergency data transmission system, roads, coolant water tanks, leakage of radio-contaminated water from spent fuel pools (SFP), presumably the Unit-1/2 exhaust stack among others.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(1):011008-011008-5. doi:10.1115/1.4041192.

Molten salt reactor (MSR) has been recognized as one of the next-generation nuclear power systems. Most MSR concepts are the variants evolved from the Oak Ridge National Laboratory (ORNL's) molten-salt breeder reactor (MSBR), which employs molten-salt as both fuel and coolant, and normally graphite is used as moderator. Many evaluations have revealed that such concepts have low breeding ratio and might present positive power coefficient. Facing these impediments, thorium molten salt reactor (TMSR) with redesigned lattice is proposed in this paper. Based on comprehensive investigation and screening, important lattice parameters including molten salt fuel composition, solid moderator material, lattice size, structure and lattice pitch to channel diameter (P/D) ratio are redesigned. In this paper, a fuel composition without BeF2 is adopted to increase the solubility for actinides (ThF4, UF4), and BeO is introduced as moderator to improve neutron economy. Moreover, lattice size and structure with cladding to separate fuel and moderator were also optimized. With these lattice parameters, TMSR has a high breeding ratio close to 1.14 and a short doubling time about 15 years. Meanwhile, a negative power coefficient is maintained. Based on this lattice design, TMSR can have excellent performance of safety and sustainability. SONG/TANG-MSR codes system is applied in the simulation, which is independently developed by Shanghai Nuclear Engineering Research & Design Institute (SNERDI).

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(1):011009-011009-7. doi:10.1115/1.4040493.
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Leaking valves have forced shutdown in many nuclear power plants. The myth of zero leakage or adequate sealing must give way to more realistic maximum leak rate criterion in design of nuclear bolted flange joints and valve packed stuffing boxes. It is well established that the predicting leakage in these pressure vessel components is a major engineering challenge to designers. This is particularly true in nuclear valves due to different working conditions and material variations. The prediction of the leak rate through packing rings is not a straightforward task to achieve. This work presents a study on the ability of microchannel flow models to predict leak rates through packing rings made of flexible graphite. A methodology based on experimental characterization of packing material porosity parameters is developed to predict leak rates at different compression stress levels. Three different models are compared to predict leakage; the diffusive and second-order flow models are derived from Naiver–Stokes equations and incorporate the boundary conditions of an intermediate flow regime to cover the wide range of leak rate levels and the lattice model is based on porous media of packing rings as packing bed ($Dp$). The flow porosity parameters ($N, R$) of the microchannels assumed to simulate the leak paths present in the packing are obtained experimentally. The predicted leak rates from different gases ($He, N2, and Ar$) are compared to those measured experimentally in which the set of packing rings is mainly subjected to different gland stresses and pressures.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(1):011010-011010-6. doi:10.1115/1.4041693.

Transient performance of China supercritical water-cooled reactor (SCWR) with the rated electric power of 1000 MWel (CSR1000) core during some typical transients, such as control rod (CR) ejection and uncontrolled CR withdrawal, is analyzed and evaluated with the coupled three-dimensional neutronics and thermal-hydraulics SCWR transient analysis code. The 3D transient analysis shows that the maximum cladding surface temperature (MCST) retains lower than safety criteria 1260 °C during the process of CR ejection accident, and the MCST retains lower than safety criteria 850 °C during the process of uncontrolled CR withdrawal transient. The safety of CSR1000 core can be ensured during the typical transients under the salient fuel temperature and water density reactivity feedback and the essential reactor protection system.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(1):011011-011011-6. doi:10.1115/1.4039884.

In the case of an accident in a nuclear power plant with combined initiating events (loss of ultimate heat sink and station blackout), an additional heat removal system could transfer the decay heat from the core to an ultimate heat sink (UHS). One specific additional heat removal system, based upon a Brayton cycle with supercritical carbon dioxide (CO2) as working fluid, is currently investigated within the European Union-funded project “sCO2-HeRo” (supercritical carbon dioxide heat removal system). It serves as a self-launching, self-propelling, and self-sustaining decay heat removal system used in severe accident scenarios. Since this Brayton cycle produces more electric power than it consumes, the excess electric power can be used inside the power plant, e.g., for recharging batteries. A small-scale demonstrator is attached to the pressurized water reactor (PWR) glass model at Gesellschaft für Simulatorschulung (GfS), Essen, Germany. In order to design and build this small-scale model, cycle calculations are performed to determine the design parameters from which a layout can be derived.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(1):011012-011012-5. doi:10.1115/1.4041563.

The inertial confinement fusion (ICF) program has been mainly concentrating on the indirect drive approach for the last three decades, due to relaxed requirements on driver-beam uniformity and reduced sensitivity to hydrodynamic instabilities. The optimal designs are important for maximum conversion of driving energy to X-rays, and finally, symmetrical irradiation of the capsule. The view factor (VF) evaluation is an important parameter providing significant radiation heat transport information in specific geometries. The present study is aimed at the VF calculations for closed cavities. The VF calculations include the case of energy transfer from one infinitesimal surface element of the enclosure to other similar infinitesimal surface elements of the cavity. Similarly, the obstructed VF is also calculated when multiple obstructions are present in the cavity. Two distinct computer programs are developed by programming in FORTRAN-90 to evaluate unobstructed VF and obstructed VF for a square geometry. The calculations are based on the crossed strings method, which is more reliable for simple geometries. The shadow effect method is used for the obstructed VF calculations. The results of the developed programs are benchmarked using the summation rule. In the case of no obstacles in the cavity, VF calculations solely obey the summation rule. However, in the presence of obstacles in the cavity, obstructed VF calculations showed the acceptable difference in comparison with the summation rule.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(1):011013-011013-9. doi:10.1115/1.4041691.

The quantities of leak rate through sealing systems are subjected to strict regulations because of the global concern on radiative materials. The maximum tolerated leak is becoming a design criterion in pressure vessel design codes, and the leak rate for an application under specific conditions is required to be estimated with reasonable accuracy. In this respect, experimental and theoretical studies are conducted to characterize gasket and packing materials to predict leakage. The amount of the total leak is the summation of the permeation leak through the sealing material and the interfacial leak generated between the sealing element and its mating surfaces. Unfortunately, existing models used to predict leakage do not separate these two types of leaks. This paper deals with a study based on experimental testing that quantifies the amount of these two types of leaks in bolted gasketed joints and packed stuffing boxes. It shows the contribution of interfacial leak for low and high contact surface stresses and the influence of the surface finish of 0.8 and 6.3 μm (32 and 250 μin) resulting from phonographic grooves in the case of a bolted flange joint. The results indicate that most leakage is interfacial, reaching 99% at the low stress while interfacial leak is of the same order of magnitude of permeation leak at high stresses reaching 10−6 and 10−8 mg/s in both packing and gaskets, respectively. Finally, particular focus is put on the technique of precompression to improve material sealing tightness.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(1):011014-011014-8. doi:10.1115/1.4040889.

The choice of materials is of great concern in the construction of Gen IV supercritical water reactors (SCWR), particularly the fuel cladding, due to the harsh environment of elevated temperatures and pressures. A material's performance under simulated conditions must be evaluated to support proper material selection by designers. In this study, aluminide and Cr-modified aluminide coated 304, as well as bare stainless steel 304 as a reference material, were tested in supercritical water (SCW) at 700 °C and 25 MPa for 1000 h. The results showed that all three samples experienced weight loss. However, the aluminide coated 304 had 20 to 40 times less weight loss compared to Cr-modified aluminide coated and bare stainless steel 304 specimens, respectively. Based on scanning electron microscope/energy dispersive X-ray spectroscopy (SEM/EDS) and X-ray diffraction (XRD) analysis results, spinel and hematite Fe2O3 formed on bare 304 after 1000 h in SCW while alumina was observed on both coated specimens, i.e., aluminide and Cr-modified aluminide surfaces. Oxide spallation was observed on the bare 304 and Cr-modified aluminide surface, contributing to a larger weight loss. Based on the results from this study, pure aluminide coating with Al content of 10–11 wt % demonstrated superior performance than bare 304 and Cr-modified aluminide coated 304.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(1):011015-011015-8. doi:10.1115/1.4041194.

In this study, based on the code Fuel ROd Behavior Analysis (FROBA), a thermal–mechanical analysis code initially developed for traditional UO2-Zr fuel elements by our research group, a modified version was developed to perform the fuel performance simulation of accident tolerant fuels (ATFs), named FROBA-ATF. Compared with initial version, the cladding could be divided into arbitrary number control volumes with different materials in the new code, so it can be used to perform the calculation for multilayer coatings. In addition, a new nonrigid pellet–cladding mechanical interaction (PCMI) calculation model was established in the new code. The FROBA-ATF code was used to predict PCMI performance of two kind fuels with coated claddings, including the internal surface coating and external surface coating. The calculation result indicates that because the coating surface was close to the inner surface of the cladding where also was the PCMI surface, the absolute value of the combine pressure of internal surface-coated cladding was substantial larger than that of the external surface-coated cladding, which might be harmful the coating behavior. However, the internal surface-coated mode can provide a protection for alloy due to the isolation from direct contact with fuel pellets, which can result in a much lower equivalent stress of zirconium body during the PCMI.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(1):011016-011016-7. doi:10.1115/1.4041193.

Guided wave (GW) testing is regularly used for finding defect locations through long-range screening using low-frequency waves (from 5 to 250 kHz). By using magnetostrictive sensors, some issues, which usually limit the application to nuclear power plants (NPPs), can be fixed. The authors have already shown the basic theoretical background and simulation results concerning a real steel pipe, used for steam discharge, with a complex structure. On the basis of such theoretical framework, a new campaign has been designed and developed on the same pipe, and the obtained experimental results are now here presented as a useful benchmark for the application of GWs as nondestructive techniques. Experimental measures using a symmetrical probe and a local probe in different configurations (pulse-echo and pitch-catch) indicate that GW testing with magnetostrictive sensors can be reliably applied to long-term monitoring of NPPs components.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(1):011017-011017-9. doi:10.1115/1.4041278.

Due to the recent high interest on in-vessel melt retention (IVR), development of detailed thermal and structural analysis tool, which can be used in a core-melt severe accident, is inevitable. Although RELAP/SCDAPSIM is a reactor analysis code, originally developed for U.S. NRC, which is still widely used for severe accident analysis, the modeling of the lower head is rather simple, considering only a homogeneous pool. PECM/S, a thermal structural analysis solver for the reactor pressure vessel (RPV) lower head, has a capability of predicting molten pool heat transfer as well as detailed mechanical behavior including creep, plasticity, and material damage. The boundary condition, however, needs to be given manually and thus the application of the stand-alone PECM/S to reactor analyses is limited. By coupling these codes, the strength of both codes can be fully utilized. Coupled analysis is realized through a message passing interface, OpenMPI. The validation simulations have been performed using LIVE test series and the calculation results are compared not only with the measured values but also with the results of stand-alone RELAP/SCDAPSIM simulations.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(1):011018-011018-8. doi:10.1115/1.4041338.

Several types of radioactive gases are released from the nuclear reactor. In order to measure the activity of such gases, it is necessary to calculate the accurate efficiency. Practically, efficiency calibration with gaseous sources is not very easy because of the low half-lives of the noble gases. For this purpose, Monte Carlo (MC) simulation was performed to study the full energy peak efficiency of two n-type high-purity Germanium (HPGe) detectors. Two spheres of xenon and krypton composition sources with two nuclides ($Xe133 and Kr85$) and two-point sources were simulated, covering the energy range from 81 keV to 604 keV. Self-absorption correction factors were calculated with GEANT4 for two gas sphere samples and obtained good efficiency agreement with the experimental results. The simulation was performed for various gas samples with different densities and observed their effects on the full energy peak efficiency value of two detectors. The corresponding self-absorption correction factors were calculated for each gaseous sample and investigated that the self-absorption correction factors not only depend on the sample characteristics but also on the detector geometry and source to detector distance. The dependence of the full energy peak efficiency on the side cap wall material and their thicknesses were also carried out for some particular photon energies.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(1):011019-011019-8. doi:10.1115/1.4041276.

Previous analyses of generation IV (GEN IV) helium gas turbine cycles indicated the possibility for high turbine entry temperatures (TETs) up to 1200 °C in order to improve cycle efficiency, using improved turbine blade material and optimum turbine cooling fractions. The purpose of this paper is to understand the effect on the levelized unit electricity cost (LUEC) of the nuclear power plant (NPP), when the TET is increased to 1200 °C from an original TET of 950 °C and when an improved turbine blade material is used to reduce the turbine cooling fraction. The analyses focus on the simple cycle recuperated (SCR) and the intercooled cycle recuperated (ICR). The baseline LUECs of the NPPs were calculated as \$61.84/MWh (SCR) and \$62.16/MWh for a TET of 950 °C. The effect of changing the turbine blades improved the allowable blade metal temperature by 15% with a reduction in the LUEC by 0.6% (SCR) and 0.7% (ICR). Furthermore, increasing the TET to 1200 °C has a significant effect on the power output but more importantly it reduces the LUECs by 22.7% (SCR) and 19.8% (ICR). The analyses intend to aid development of the SCR and ICR including improving the decision making process on choice of cycles applicable to the gas-cooled fast reactors (GFRs) and very high-temperature reactors (VHTRs), where helium is the coolant.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(1):011020-011020-8. doi:10.1115/1.4040377.

In this study, we numerically evaluated the performance of a steam methane reforming (SMR) reactor heated using high-temperature helium for hydrogen production. The result showed that with an increase in the reactant gas inlet velocity, the temperature at the same reactor length position decreased. The maximum gas temperature difference at the gas collection chamber reached approximately 55 °C. The outlet temperature difference increased to 35 °C when the inlet temperature increased from 370 °C to 570 °C. A higher inlet temperature did not have a positive effect on the system's thermal efficiency. The methane conversion rate increased by 68%, and the hydrogen production rate increased by 55%, when the helium inlet velocity increased from 2 m/s to 22 m/s. When the helium inlet temperature increased by 200 °C, the highest temperature of the reactant gas increased by 132 °C. In the SMR for hydrogen production using a high-temperature gas-cooled reactor (HTGR), low reactant-gas inlet velocity, suitable inlet temperature, high inlet velocity, and a high HTGR outlet temperature of helium were preferable.

Commentary by Dr. Valentin Fuster

### Technical Brief

ASME J of Nuclear Rad Sci. 2019;5(1):014501-014501-3. doi:10.1115/1.4041274.

The hydrogen emission of zirconium hydride at high temperature is a challenging issue for many researchers. The hydrogen emission content of zirconium hydride pins should be evaluated to confirm the application feasibility. The comparison of theory analysis and experiment data indicated that Richardson's law could offer a conservative result for calculating the hydrogen emission content of zirconium hydride pins at high temperature. Furthermore, the methods of preventing hydrogen loss should be developed for the purpose of extending the work temperature or time. The results showed that a ZrO2 layer prepared for zirconium hydride could not prevent hydrogen loss after exposure at 650 °C in an inert environment and ZrO2 transformed into Zr3O gradually due to the opposite movement of hydrogen and oxygen. Finally, a further improvement to prevent hydrogen loss was developed. The zirconium hydride with a ZrO2 layer in the cladding of He+CO2 exhibited no significant reduction of hydrogen content. It is helpful to prevent the hydrogen loss by increasing the oxygen potential on the outside of ZrO2 layer.

Commentary by Dr. Valentin Fuster