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IN THIS ISSUE

### Guest Editorial

ASME J of Nuclear Rad Sci. 2015;1(4):040301-040301-1. doi:10.1115/1.4031266.

Nuclear energy is faced with major challenges ranging from public perception and acceptance to nonproliferation and continued operation and enhanced safety of existing fleet with lessons learned from recent Fukushima Daiichi accidents. The industry itself recognizes the ultimate importance of ensuring overall safety, where research and development play an important role in supporting safe operation of nuclear power plants. Understanding the mechanisms and underlying physics and developing measures to prevent and/or mitigate decay of plant performances are still challenging tasks for utilities and technical support organizations. Having in mind the global nature of these nuclear technology issues, it is imperative that professional forums, such as international journals and technical professional meetings, provide a global communication channel to foster the exchange of ideas and critical information and to enhance cross fertilization of research and development activities through our professional community.

Commentary by Dr. Valentin Fuster

### Research Papers

ASME J of Nuclear Rad Sci. 2015;1(4):041001-041001-14. doi:10.1115/1.4031098.

In recent years, several small modular reactor (SMR) designs have been developed. These nuclear power plants (NPPs) not only offer a small power size (less than 300 MWe), a reduced spatial footprint, and modularized compact designs fabricated in factories and transported to the intended sites, but also passive safety features. Some light water (LW)-SMRs have already been granted by Department of Energy: NuScale and mPower. New LW-SMRs are mainly inspired by the early LW-SMRs (such as process-inherent ultimate safety (PIUS), international reactor innovative and secure (IRIS), and safe integral reactor (SIR)). LW-SMRs employ significantly fewer components to decrease costs and increase simplicity of design. However, new physical challenges have appeared with these changes. At the same time, advanced SMR (ADV-SMR) designs (such as PBMR, MHR Antares, Prism, 4S, and Hyperion) are being developed that have improved passive safety and other features. This paper quantitatively and qualitatively compares most of the LW- and ADV-SMRs with respect to reactors, nuclear fuel, containment, reactor coolant systems, refueling, and emergency coolant systems. Economic and financing evaluations are also included in the paper. The detailed comparisons in this paper elucidate that one reactor is not superior to the others analyzed in this study, as each reactor is designed to meet different needs.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2015;1(4):041002-041002-11. doi:10.1115/1.4029927.

This paper describes a sequence of residual stress measurements made to determine a two-dimensional map of biaxial residual stress in a stainless steel weld. A long stainless steel (316L) plate with an eight-pass groove weld (308L filler) was used. The biaxial stress measurements follow a recently developed approach, comprising a combination of contour method and slitting measurements, with a computation to determine the effects of out-of-plane stress on a thin slice. The measured longitudinal stress is highly tensile in the weld- and heat-affected zone, with a maximum around 450 MPa, and compressive stress toward the transverse edges around −250 MPa. The total transverse stress has a banded profile in the weld with highly tensile stress at the bottom of the plate (y = 0) of 400 MPa, rapidly changing to compressive stress (at y = 5 mm) of −200 MPa, then tensile stress at the weld root (y = 17 mm) and in the weld around 200 MPa, followed by compressive stress at the top of the weld at around −150 MPa. The results of the biaxial map compare well with the results of neutron diffraction measurements and output from a computational weld simulation.

Topics: Stress
Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2015;1(4):041003-041003-7. doi:10.1115/1.4030982.

Monolithic plate-type fuels comprise a high-density, low-enrichment, U10Mo fuel foil encapsulated in a cladding material. This concept generates several fabrication challenges, including flatness, centering, or thickness variation. There are concerns whether these parameters have implications on overall performance. To investigate these inquiries, the effects of the foil flatness were studied. For this, a representative plate was simulated for an ideal case. The simulations were repeated for additional cases with various foil curvatures to evaluate the effects on the irradiation performance. The results revealed that the stresses and strains induced by fabrication process are not affected by the flatness of the foil. Furthermore, fabrication stresses in the foil are relieved relatively fast in the reactor. The effects of the foil flatness on peak irradiation stress-strains are minimal. There is a slight increase in temperature for the case with maximum curvature. The major impact is on the displacement characteristics. While the case with a flat foil produces a symmetrical swelling, if the foil is curved, more swelling occurs on the thin-cladding side and the plate bows during irradiation.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2015;1(4):041004-041004-10. doi:10.1115/1.4029918.

State-of-the-art neutron detectors lack capabilities required by the fields of homeland security, health physics, and even for direct in-core nuclear power monitoring. A new system being developed at Purdue’s Metastable Fluid and Advanced Research Laboratory in conjunction with S/A Labs, LLC provides capabilities that the state-of-the-art lacks, and simultaneously with beta ($β$) and gamma ($γ$) blindness, high ($>90%$ intrinsic) efficiency for neutron/alpha spectroscopy and directionality, simple detection mechanism, and lowered electronic component dependence. This system, the tensioned metastable fluid detector (TMFD), provides these capabilities despite its vastly reduced cost and complexity compared with equivalent present day systems. Fluids may be placed at pressures lower than perfect vacuum (i.e., negative), resulting in tensioned metastable states. These states may be induced by tensioning fluids just as one would tension solids. The TMFD works by cavitation nucleation of bubbles resulting from energy deposited by charged ions or laser photon pile-up heating of fluid molecules, which are placed under sufficiently tensioned (negative) pressure states of metastability. The charged ions may be created from neutron scattering or from energetic charged particles such as alphas, alpha recoils, and fission fragments. A methodology has been created to profile the pressures in these chambers by laser-induced cavitation (LIC) for verification of a multiphysics simulation of the chambers. The methodology and simulation together have led to large efficiency gains in the current acoustically tensioned metastable fluid detector (ATMFD) system. This paper describes in detail the LIC methodology and provides background on the simulation it validates.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2015;1(4):041005-041005-11. doi:10.1115/1.4031031.

A joint experimental and numerical campaign is conducted to provide validation dataset of high-fidelity fluid–structure interaction (FSI) models of nuclear fuel assemblies during seismic loading. A refractive index-matched (RIM) flow loop is operated on a six-degree-of-freedom shake table and instrumented with nonintrusive optical diagnostics. The test section can house up to three full-height fuel assemblies. To guarantee reproducible and controlled initial conditions, special care is given to the test section inlet plenum; in particular, it is designed to minimize secondary pulsatile flow that may arise due to ground acceleration. A single transparent surrogate $6×6$ fuel subassembly is used near prototypical Reynolds number, $Re=105$ based on hydraulic diameter. To preserve dynamic similarity of the model with prototype, the main dimensionless parameters are matched and custom spacer grids are designed. Special instruments are developed to characterize fluid and structure response and to operate in this challenging shaking environment. In parallel to the earlier experiments, we also conducted fully coupled direct numerical simulations, where the equations for the fluid and the structure are simultaneously advanced in time using a partitioned scheme. To deal with the highly complex geometrical configuration, which also involves large displacements and deformations, we utilize a second-order accurate, immersed boundary (IB) formulation, where the geometry is immersed in a block-structured grid with adaptive mesh refinement (AMR). To explore a wide parametric range, we will consider several subsets of the experimental configuration. A typical computation involves 60,000 cores, on leadership high-performance computing facilities (i.e., IBM Blue-Gene Q–MIRA).

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2015;1(4):041006-041006-13. doi:10.1115/1.4030833.

Two mathematical models (a one-dimensional (1D) and a two-dimensional (2D)) were adopted to study, numerically, the thermal-hydrodynamic characteristics of flow inside the cooling channels of a nuclear thermal rocket (NTR) engine. In the present study, only one of the cooling channels of the reactor core is simulated. The 1D model adopted here assumes the flow in this cooling channel to be unsteady, compressible, turbulent, and subsonic. The governing equations of the compressible flow in the cooling channel are discretized using a second-order accurate (MacCormack) finite-difference scheme. The steady-state results of the proposed model were compared to the predictions by a commercial CFD code. The 2D CFD solution was obtained in two domains: the coolant (gaseous hydrogen) and the ZrC fuel cladding. The wall heat flux which varied along the channel length (as described by the nuclear variation in the nuclear power generation) was given as an input. Numerical experiments were carried out using both codes to simulate the thermal and hydrodynamic characteristics of the flow inside a single-cooling channel of the reactor for a typical Nuclear Engine for Rocket Vehicle Application (NERVA)-type NTR engine. It is concluded that both models predict successfully the steady-state axial distributions of temperature, pressure, density, and velocity of gaseous hydrogen flow in the NTR cooling channel.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2015;1(4):041007-041007-8. doi:10.1115/1.4031033.

Oxygen redistribution with a high-temperature gradient is an important fuel performance concern in fast-breeder reactor (FBR) and light-water reactor (LWR) $(U,Pu)O2$ fuel under irradiation, and affects fuels properties, power distribution, and fuel overall performance. This paper studies the burnup dependent oxygen and heat diffusion behavior in a fully coupled way within $(U,Pu)O2$ FBR and LWR fuels. The temperature change shows relatively larger impact on oxygen to metal (O/M) ratio redistribution rather than O/M ratio change on temperature, whereas O/M ratio redistributions show different trends for FBR and LWR fuels due to their different deviations from the stoichiometry of oxygen under high-temperature environments.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2015;1(4):041008-041008-7. doi:10.1115/1.4030961.

During the course of a severe accident in a nuclear power plant, water can be collected in the sump containment through steam condensation on walls, cooling circuit leak, and by spray systems activation. Therefore, the sump can become a place of heat and mass exchanges through water evaporation and steam condensation, which influences the distribution of hydrogen released in containment during nuclear core degradation. The objective of this paper is to present the analysis of semi-analytical experiments on sump interaction between containment atmosphere for typical accidental thermal hydraulic conditions in a pressurized water reactor (PWR). Tests are conducted in the TOSQAN facility developed by the Institut de Radioprotection et de Sûreté Nucléaire in Saclay. The TOSQAN facility is particularly well adapted to characterize the distribution of gases in a containment vessel. A tests’ grid was defined to investigate the coupled effect of the sump evaporation with wall condensation, for air steam conditions, with noncondensable gases (He, $SF6$), and for steady and transient states (two depressurization tests).

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2015;1(4):041009-041009-9. doi:10.1115/1.4031032.

One-dimensional (1D) sensitivity computations were carried out for air–water countercurrent flows in a $1/15$-scale model of the hot leg and a $1/10$-scale model of the pressurizer surge line in a pressurized water reactor (PWR) to generalize the prediction method for countercurrent flow limitation (CCFL) characteristics in slightly inclined pipes with elbows. In the 1D model, the wall friction coefficient $fwG$ of single-phase gas flows was used. The interfacial drag coefficient of $fi=0.03$, an appropriate adjustment factor of $NwL=6$ for the wall friction coefficient $fwL$ of single-phase liquid flows ($NwG=1$ for $fwG$ of single-phase gas flows), and an appropriate adjustment factor of $Nde=6$ for the pressure loss coefficient $ζe$ of elbows in single-phase flows were determined to give good agreement between the computed and measured CCFL characteristics. The adjusted factors were used to compute and then discuss effects of the inclination angle and diameter on CCFL characteristics.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2015;1(4):041010-041010-12. doi:10.1115/1.4030963.

$Li2BeF4$, or flibe, is the primary candidate coolant for the fluoride-salt-cooled high-temperature nuclear reactor (FHR). Kilogram quantities of pure flibe are required for repeatable corrosion tests of modern reactor materials. This paper details fluoride salt purification by the hydrofluorination–hydrogen process, which was used to regenerate 57.4 kg of flibe originating from the secondary loop of the molten salt reactor experiment (MSRE) at Oak Ridge National Laboratory (ORNL). Additionally, it expounds upon necessary handling precautions required to produce high-quality flibe and includes technological advancements which ease the purification and analysis process. Flibe batches produced at the University of Wisconsin are the largest since the MSRE program, enabling new corrosion, radiation, and thermal hydraulic testing around the United States.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2015;1(4):041011-041011-8. doi:10.1115/1.4030199.

The present work analyzes the thermal-hydraulic behavior of the CANDU supercritical water reactor (SCWR) using a 1-D numerical model. The possibility of a static instability, the Ledinegg excursion, is investigated, which reveals it can occur only in a hypothetical condition, far from the proposed operating regime of the CANDU SCWR. The investigation demonstrates the possibility of density wave oscillations (DWOs), a dynamic instability, in the operating regime of the CANDU SCWR and its marginal stability boundary (MSB) is obtained. The phenomenon of the deterioration in heat transfer is observed, and the related investigation shows that the strong buoyancy effect is responsible for its appearance inside the heating section of the channel of the CANDU SCWR core. The MSB is found to be inadequate in determining the safe operating zone of the reactor because the wall temperature can exceed the allowable limit from metallurgical consideration. The investigations also determine the safe as well as stable zone where the CANDU SCWR should operate in order to avoid the maximum temperature limit and DWOs.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2015;1(4):041012-041012-9. doi:10.1115/1.4030046.

A thermal analysis was performed for the Advanced Gas Reactor test experiment (AGR-3/4) with time-varying gas gaps. The experiment was irradiated at the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). Several fuel irradiation experiments are planned for the AGR Fuel Development and Qualification Program that supports the development of the Very-High-Temperature Gas-Cooled Reactor (VHTR) under the Next-Generation Nuclear Plant (NGNP) project. AGR-3/4 combines two tests in a series of planned AGR experiments to test tristructural-isotropic (TRISO)-coated, low-enriched uranium oxy-carbide fuel. Forty-eight TRISO-fueled compacts were inserted into 12 separate capsules for the experiment (four compacts per capsule). The purpose of this analysis was to calculate the temperatures of each compact and graphite layer to obtain daily average temperatures using time (fast neutron fluence)-varying gas gaps and to compare with experimentally measured thermocouple data. A finite-element heat transfer model was created for each capsule using the commercial code ABAQUS. Model results are compared to thermocouple data taken during the experiment.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2015;1(4):041013-041013-6. doi:10.1115/1.4030799.

A lead–bismuth eutectic (LBE)-cooled accelerator-driven system (ADS) of 30 MeV and 0.5 mA proton beam has been simulated. The performance of this 15-kW ADS has been analyzed for three coolants (LBE, air, and water), all under variable and constant heat loads using the thermal hydraulic code RELAP5/Mod4.0. Steady-state simulation results for temperature of coolants match the reported design values within 3.2% of relative error. The effect of variation of mass flow rate on power extraction has also been evaluated for the three coolants, namely, LBE, air, and water.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2015;1(4):041014-041014-12. doi:10.1115/1.4030983.

The ability of coated particles of enriched uranium dioxide fuel encased in graphite to discontinue nuclear fission reaction without human action in the case of complete loss of cooling is a vital safety measure over traditional nuclear fuel. As a possible solution toward enhancing the safety of light water reactors (LWRs), it is envisaged that the fuel, in the form of loose, coated particles in a helium atmosphere, can be used inside the cladding tubes of the fuel elements. This study is therefore a first step toward understanding the heat-transfer characteristics under natural convective conditions within the fuel cladding tubes of such a revolutionary new fuel design. The coated particle fuels are treated as a bed, from which the heat is transferred to the cladding tube and the gas movement occurs due to natural convection. A basic unit cell model was used where a single unit of the packed bed was analyzed and taken as representative of the entire bed. The model is a combination of both analytical and numerical methods accounting for the thermophysical properties of sphere particles, the interstitial gas effect, gas temperature, contact interface between particles, particle size, and particle temperature distribution used in this study to investigate the heat-transfer effect. The experimental setup was a packed bed heated from below with gas circulation due to natural convection. This allows for the development of an appropriate, conservative thermal energy balance that can be used in determining the heat-transfer characteristics in homogeneous porous media. Success in this method, when validated with suitable correlation, such as Gunn, suggests that the heat-transfer phenomenon/characteristics in the fuel cladding tube of the new design can be evaluated using this approach for design purpose.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2015;1(4):041015-041015-12. doi:10.1115/1.4030365.

The process of heat transfer in a heavy liquid-metal coolant (HLMC) cross-flow around heat-transfer tubes has not been thoroughly studied yet. Therefore, it is of great interest to carry out experimental studies for determining the heat-transfer characteristics in lead coolant cross-flow around tubes. It is also interesting to explore the velocity and temperature fields in an HLMC flow. To achieve this goal, experts of the R.E. Alekseev Nizhny Novgorod State Technical University performed work aimed at experimental determination of the temperature and velocity fields in high-temperature lead coolant cross-flows around a tube bundle. The experimental studies were carried out in a specially designed high-temperature liquid-metal facility. The experimental facility is a combination of two high-temperature liquid-metal setups, i.e., FT-2 with a lead coolant and FT-1 with a lead-bismuth coolant, combined by an experimental site. The experimental site is a model of the steam generator of the BREST reactor facility. The heat-transfer surface is an in-line tube bank of diameter 17 mm and wall thickness of 3.5 mm, which is made of 10H9NSMFB ferritic–martensitic steel. The temperature of the heat-transfer surface is measured with thermocouples of diameter 1 mm installed in the walls of heat-transfer tubes. The velocity and temperature fields in a high-temperature HLMC flow are measured with special sensors installed in the flow cross-section between rows of heat-transfer tubes. The characteristics of heat transfer and velocity fields in a lead coolant flow were studied in different directions of the coolant flow: the vertical (“top-down” and “bottom-up” (Beznosov et al., 2013, “Experimental Studies of Thermal Hydraulics of a HLMC Flow Around Heat transfer Surfaces,” Proceedings of the 21st International Conference on Nuclear Engineering, ICONE21, Paper No. ICONE21-15248)) and the horizontal directions. The studies were conducted under the following operating conditions: the temperature of lead was $t=450–500°C$, the thermodynamic activity of oxygen was $a=10−5–100$, and the lead flow through the experimental site was $Q=3–6 m3/h$, which corresponds to coolant velocities of $V=0.4–0.8 m/s$. Comprehensive experimental studies of the characteristics of heat transfer in a lead coolant cross-flow around tubes have been carried out for the first time, and the dependences $Nu=f(Pe)$ for a controlled and regulated content of the thermodynamically active oxygen impurity and sediments of impurities have been obtained. The effect of the oxygen impurity content in the coolant and characteristics of protective oxide coatings on the temperature and velocity fields in a lead coolant flow have been revealed. This is because the presence of oxygen in the coolant and oxide coatings on the surface, which restricts the liquid-metal flow, leads to a change in the characteristics of the wall-adjacent region. The obtained experimental data on the distribution of the velocity and temperature fields in an HLMC flow permit studying the heat-transfer processes, and on this basis, create program codes for engineering calculations of HLMC flows around heat-transfer surfaces.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2015;1(4):041016-041016-10. doi:10.1115/1.4030834.

The purpose of the study was to investigate the application of deterministic safety analysis for a prevention strategy of the extended station blackout (SBO). A method for the assessment of portable pump flow rates for a steam generator (SG) makeup is proposed. The RELAP5/MOD3.3 computer code and input model of a two-loop pressurized water reactor (PWR) is used for analyses, assuming different injection start times, flow rates, and reactor coolant system (RCS) losses. The calculated results show the effectiveness of the proposed extended SBO prevention strategy. The results demonstrate the need for mitigation of the common cause failure (CCF) potential of the on-site and off-site power sources.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2015;1(4):041017-041017-10. doi:10.1115/1.4030504.

We have proposed laser-assisted depletion of $Gd152$ isotope from a natural isotopic mixture of Gd to enhance its functional efficiency as a burnable poison. Theoretical investigations on laser-assisted depletion of $Gd152$ isotope from natural gadolinium have been carried out for two-color resonant three-color photoionization pathways using density matrix formalism. Calculations have been carried out using a density matrix formalism to optimize conditions for high ionization efficiency without much sacrifice in the isotopic selectivity. Optimum conditions for good isotopic selectivity of $Gd152$ without significant sacrifice in the ion yield have been identified. Under appropriate conditions, all the 17 photoionization schemes are found to be useful for the laser-assisted separation of $Gd152$ isotopes which can be used for reactor applications. The effect of source, laser, and atom parameters on isotopic selectivity and ionization efficiency has been investigated. Among the photoionization schemes investigated, one of the photoionization scheme has been investigated in detail. Under optimized conditions, this photoionization scheme has resulted in high ionization efficiency ($>50%$) and high isotopic selectivity $(1.2×104)$.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2015;1(4):041018-041018-20. doi:10.1115/1.4030364.

A severe accident (SA) is defined as an incident involving melting of the nuclear reactor core and the release of fission products (FP) from the fuel and their associated risks. In the SA, the containment may fail, causing the public hazard of fission products released to the environment. This review elaborates the resolved issues of SAs under the condition of a hypothetical SA. SA research that has been performed over the years is briefly described, including various SA scenarios. The SA scenarios involve core melt scenarios from the beginning of core degradation to melt formation and relocation into the lower head and to the containment, the interactions of the molten corium with water and concrete, the behavior of fission products in- and ex-vessel, hydrogen-related phenomena, and all associated risks. The mitigation strategies that have been adopted in existing reactors and advanced light water reactors (ALWR) are also discussed. These mitigation measures can keep the reactor vessel or containment intact and terminate the SA progression. SA analysis codes are then summarized and divided into three categories, namely, systematic analysis codes, mechanism analysis codes, and single-function analysis codes. Next, the unresolved issues of SAs are proposed, including narrow gap cooling, melt chemical interactions, steam explosion loads, molten debris coolability, and iodine chemistry. Further experimental and theoretical research activities should be conducted to resolve these issues; consequently, some recommendations for further research work are also given in the last part of this review. This review aims to add to the knowledge and understanding of SA research in the past few decades and to benefit further research of SAs.

Commentary by Dr. Valentin Fuster

### Erratum

ASME J of Nuclear Rad Sci. 2015;1(4):047001-047001-6. doi:10.1115/1.4031143.

Some figures were incorrectly referenced within the text. The following pages show the correct text references.

Commentary by Dr. Valentin Fuster