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### Editorial

ASME J of Nuclear Rad Sci. 2015;1(1):010201-010201-2. doi:10.1115/1.4029141.
Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2015;1(1):010202-010202-1. doi:10.1115/1.4029140.
Commentary by Dr. Valentin Fuster

### Guest Editorial

ASME J of Nuclear Rad Sci. 2015;1(1):010301-010301-2. doi:10.1115/1.4029139.
Commentary by Dr. Valentin Fuster

### Research Papers

ASME J of Nuclear Rad Sci. 2015;1(1):011001-011001-19. doi:10.1115/1.4029420.
OPEN ACCESS

It is well known that electrical power generation is the key factor for advances in industry, agriculture, technology, and standards of living. Also, a strong power industry with diverse energy sources is very important for a nation’s independence. In general, electrical energy can be generated from (1) burning mined and refined energy sources such as coal, natural gas, oil, and nuclear; and (2) harnessing energy sources such as hydro, biomass, wind, geothermal, solar, and wave power. Today, the main sources for electrical energy generation are (1) thermal power, primarily using coal and secondarily natural gas; (2) “large” hydraulic power from dams and rivers; and (3) nuclear power from various reactor designs. The balance of the energy sources is from using oil, biomass, wind, geothermal, and solar, which have a visible impact just in some countries. This paper presents the current status and role of the nuclear-power industry in the world with a comparison of nuclear-energy systems to other energy systems.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2015;1(1):011002-011002-12. doi:10.1115/1.4026389.

The structural integrity of the containment vessel (CV) for a pressurized water reactor (PWR) plant under a loss-of-coolant accident is evaluated by a safety analysis code that uses the average temperature of gas phase in the CV during reactor operation as an initial condition. Since the estimation of the average temperature by measurement is difficult, this paper addressed the numerical simulation for the temperature distribution in the CV of an operating PWR plant. The simulation considered heat generation of the equipment, the ventilation and air conditioning systems (VAC), heat transfer to the structure, and heat release to the CV exterior based on the design values of the PWR plant. The temperature increased with a rise in height within the CV and the flow field transformed from forced convection to natural convection. Compared with the measured temperature data in the actual PWR plant, predicted temperatures in the lower regions agreed well with the measured values. The temperature differences became larger above the fourth floor, and the temperature inside the steam generator (SG) loop chamber on the fourth floor was most strongly underestimated, $−16.2 K$ due to the large temperature gradient around the heat release equipment. Nevertheless, the predicted temperature distribution represented a qualitative tendency, low at the bottom of the CV and increases with a rise in height within the CV. The total volume-averaged temperature was nearly equal to the average gas phase temperature. To improve the predictive performance, parameter studies regarding heat from the equipment and the reconsideration of the numerical model that can be applicable to large temperature gradient around the equipment are needed.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2015;1(1):011003-011003-10. doi:10.1115/1.4026391.

The fluoride-salt-cooled high-temperature reactor (FHR) is an advanced reactor concept that uses tristructural isotropic (TRISO) high-temperature fuel and low-pressure liquid salt coolant. A $20-MWth$ test reactor, as the key step in demonstrating the technical feasibility, is currently under design at Massachusetts Institute of Technology. This study focuses on the coupled conduction and convection heat transfer adopting a three-dimensional unit-cell model with one coolant channel and six one-third fuel compacts. The laminar, transitional, and turbulent flows are investigated with the use of computational fluid dynamic (CFD) software, CD-adapco STARCCM+. The model is validated against theory for developing laminar flow in the benchmark study with excellent agreement. The model is also benchmarked for transitional and turbulent flows by Hausen, Gnielinski, Dittus-Boelter, and Sieder-Tate correlations. Azimuthal distributions of temperature, heat flux, and heat transfer coefficient along the coolant-graphite interface were obtained for the asymmetric heat source, graphite materials, and two different types of salt coolant. The results show that the asymmetric power generation has little impact on peak fuel temperature, interface temperature, and heat transfer coefficient for a unit-cell module in laminar flow regime due to effective thermal conduction of the graphite matrix. In the turbulent flow regime, the effect on the azimuthal heat flux and heat transfer coefficient is more pronounced.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2015;1(1):011004-011004-9. doi:10.1115/1.4026392.

A new process for determination of in-service inspection (ISI) requirements was proposed on the basis of the system based code concept to realize effective and rational ISI by properly taking into account plant-specific features. The proposed process consists of two complementary evaluations, one focusing on structural integrity and the other one on detectability of defects before they would grow to an unacceptable size in light of plant safety. If defect detection was not feasible, structural integrity evaluation would be required under sufficiently conservative hypothesis. The applicability of the proposed process was illustrated through an application to the existing prototype fast breeder reactor, Monju.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2015;1(1):011005-011005-10. doi:10.1115/1.4026390.

Small-scale experiments were carried out to characterize the resuspension factor of radioactive lanthanum oxide powder in an environmentally controlled wind tunnel, with the majority using particle sizes less than $10 μm$ in order to assess the impact of wind resuspension stresses and surface roughness conditions on resuspension. Operational principles of the measuring devices used in the radionuclide resuspension experiments and corresponding uncertainties are discussed. The average bin-by-bin particle resuspension factors ($ki$) for particle sizes, in the range of $0.25–7.00 μm$ and $7.00–12.5 μm$ for downwind fallout locations, were calculated and are reported here as $1.14×10−3 1/m$ and $4.39×10−2 1/m$, respectively.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2015;1(1):011006-011006-10. doi:10.1115/1.4026387.

Nuclear power becomes more and more important in many countries worldwide as a basis for current and future electrical energy generation. The largest group of operating nuclear power plants (NPPs) equipped with water-cooled reactors (96% of all NPPs) has gross thermal efficiencies ranging from 30–36%. Such relatively low values of thermal efficiencies are due to lower pressures/temperatures at the inlet to a turbine ($4.5–7.8 MPa/257–293°C$). However, modern combined-cycle power plants (Brayton gas-turbine cycle and subcritical-pressure steam Rankine cycle, fueled by natural gas) and supercritical-pressure coal-fired power plants have reached gross thermal efficiencies of 62% and 55%, respectively. Therefore, next generation or Generation IV NPPs with water-cooled reactors should have thermal efficiencies as close as possible to those of modern thermal power plants. A significant increase in thermal efficiencies of water-cooled NPPs can be possible only due to increasing turbine inlet parameters above the critical point of water, i.e., supercritical water-cooled reactors (SCWRs) have to be designed. This path of increasing thermal efficiency is considered as a conventional way that coal-fired power plants followed more than 50 years ago. Therefore, an objective of the current paper is a study on neutronics and thermalhydraulics characteristics of a generic $1200-MWel$ pressure-channel (PCh) SCWR. Standard neutronics codes DRAGON and DONJON have been coupled with a new thermalhydraulics code developed based on the latest empirical heat-transfer correlation, which allowed for more accurate estimation of basic characteristics of a PCh SCWR. In addition, the computational fluid dynamics (CFD) Fluent code has been used for better understanding of the specifics of heat transfer in supercritical water. Future studies will be dedicated to materials and fuels testing in an in-pile supercritical water loop and developing passive safety systems.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2015;1(1):011007-011007-7. doi:10.1115/1.4026394.

The fluoride-salt-cooled high-temperature reactor (FHR) is an advanced reactor concept that uses high-temperature tristructural isotropic (TRISO) fuel with a low-pressure liquid salt coolant. Design of the fluoride-salt-cooled high-temperature test reactor (FHTR) is a key step in the development of the FHR technology and is currently in progress both in China and the United States. An FHTR based on pebble-bed core design with a coolant temperature of 600–700°C is being planned for construction by the Chinese Academy of Sciences’ (CAS) Thorium Molten Salt Reactor (TMSR) Research Center, Shanghai Institute of Applied Physics (SINAP). This paper provides preliminary thermal-hydraulic transient analyses of an FHTR using SINAP’s pebble-bed core design as a reference case. A point kinetic model is implemented using computer code by coupling with a simplified porous medium heat transfer model in the core. The founded models and developed code are applied to analyze the safety characteristics of the FHTR by simulating several transient conditions including the unprotected loss of flow, unprotected overcooling, and unprotected transient overpower accidents. The results show that SINAP’s pebble-bed core is a very safe reactor design.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2015;1(1):011008-011008-8. doi:10.1115/1.4026395.

The appropriate description of heat transfer to coolants at the supercritical state is limited by the current understanding. Thus, this poses one of the main challenges in the development of supercritical-fluids applications for Generation-IV reactors. Since the thermodynamic critical point of water is much higher than that of carbon dioxide ($CO2$), it is more affordable to run heat-transfer experiments in supercritical $CO2$. The data for supercritical $CO2$ can be later scaled and used for supercritical water-based reactor designs. The objective of this paper is, therefore, to discuss the basis for comparison of relatively recent experimental data on supercritical $CO2$ obtained at the facilities of the Korea Atomic Energy Research Institute (KAERI) and Chalk River Laboratories (CRL) of the Atomic Energy of Canada Limited (AECL). Based on the available instrumental error, a thorough analysis of experimental errors in wall- and bulk-fluid temperatures and heat transfer coefficient was conducted. A revised heat-transfer correlation for the CRL data is presented. A dimensional criterion for the onset of the deteriorated heat transfer in the form of a linear relation between heat flux and mass flux is proposed. A preliminary heat-transfer correlation for the joint CRL and KAERI datasets is presented.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2015;1(1):011009-011009-8. doi:10.1115/1.4026401.

Molten salt reactors (MSRs) are promising advanced nuclear reactors for closure of the fuel cycle. This paper discusses the core design of graphite-moderated MSRs, thanks to a parametric study of the fuel and moderator lattice. The study is conducted at equilibrium of the thorium-uranium fuel cycle for several fuel channel radius and moderator block size combinations. The equilibrium composition for each studied configuration is derived with the help of an in-house MATLAB code, EQL0D, which uses the Serpent 2 Monte Carlo neutronics code for the calculation of reaction rates. The results include excess reactivity at equilibrium, mirroring the breeding gain, and the actinide vector composition for each configuration. Moreover, the occurence of an optimum of the excess reactivity per percent uranium-233 was observed. The investigations showed that it is systematically seen at an interchannel distance equal to the neutron slowing-down length in graphite for each configuration and does not depend on the salt channel radius beyond a certain size, which is given by the thermal fission rate reaching the levels of the fast fission rate. In this way, an exotic energy and spatial distribution of the neutrons are attained. The investigations highlight the potential attractiveness, from a neutronics/fuel cycle point of view, of both large fuel channels and moderators with a shorter neutron slowing-down length.

Commentary by Dr. Valentin Fuster

### Technical Brief

ASME J of Nuclear Rad Sci. 2015;1(1):014501-014501-2. doi:10.1115/1.4026388.

In this study, a note on mixture density using the Shannak definition of the Froude number is presented (Shannak, B., 2009, “Dimensionless Numbers for Two-Phase and Multiphase Flow,” Proceedings of the International Conference on Applications and Design in Mechanical Engineering (ICADME), Penang, Malaysia, Oct. 11–13, 2009). From the definition of the two-phase Froude number, an expression of the two-phase density is obtained. The definition of the two-phase density can be used to compute the two-phase frictional pressure gradient using the homogeneous modeling approach in circular pipes, minichannels, and microchannels. We cannot have $gas density≤two-phase density≤liquid density$ for $0≤mass quality≤1$. Therefore, attention must be paid when using the obtained expression of the two-phase density in this note at any $x$ value.

Commentary by Dr. Valentin Fuster