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In Memoriam

ASME J of Nuclear Rad Sci. 2019;5(4):040101-040101-2. doi:10.1115/1.4043661.

Professor Geoffrey Hewitt died on the Jan. 18, 2019, after a life as husband, father, and a giant of the field of multiphase flow and heat transfer. He was a passionate and dedicated teacher and researcher and leaves an incredible legacy to the profession.

Commentary by Dr. Valentin Fuster

Research Papers

ASME J of Nuclear Rad Sci. 2019;5(4):041201-041201-11. doi:10.1115/1.4042116.

The simple cycle recuperated (SCR) and intercooled cycle recuperated (ICR) are highly efficient Brayton helium gas turbine cycles, designed for the gas-cooled fast reactor (GFR) and very-high-temperature reactor (VHTR) generation IV (Gen IV) nuclear power plants (NPPs). This paper documents risk analyses, which consider technical and economic aspects of the NPP. The sensitivity analyses are presented that interrogate the plant design, performance, and operational schedule and range from component efficiencies, system pressure losses, operating at varied power output due to short-term load-following or long-term reduced power operations to prioritize other sources such as renewables. The sensitivities of the economic and construction schedule are also considered in terms of the discount rates, capital and operational costs, and increased costs in decontamination and decommissioning (D&D) activity due to changes in the discount rates. This was made possible by using a tool designed for this study to demonstrate the effect on the “noncontingency” baseline levelized unit electricity cost (LUEC) of both cycles. The SCR with a cycle efficiency of 50% has a cheaper baseline LUEC of $58.41/MWh in comparison to the ICR (53% cycle efficiency), which has an LUEC of $58.70/MWh. However, the cost of the technical and economic risks is cheaper for the ICR resulting in a final LUEC of $70.45/MWh (ICR) in comparison to the SCR ($71.62/MWh) for the year 2020 prices.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(4):041202-041202-10. doi:10.1115/1.4043295.

Recently, the supercritical carbon dioxide Brayton (SCO2) cycle gained a lot of attention for its application to next-generation nuclear reactors. Turbine is the key component of the energy conversion in the thermodynamic cycle. Transonic centrifugal turbine has advantages of compatibility of aerodynamic and geometric, low cost, high power density, and high efficiency; therefore, it has opportunity to become the main energy conversion equipment in the future. In this paper, a transonic nozzle and its corresponding rotor cascade of the single-stage centrifugal turbine were designed. In addition, the three-dimensional (3D) numerical simulation and performance analysis were conducted. The numerical simulation results show that the predicted flow field is as expected and the aerodynamic parameters are in good agreement with one-dimensional (1D) design. Meanwhile, the off-design performance analysis shows that the transonic centrifugal turbine stage has wide stable operation range and strong load adaptability. Therefore, it can be concluded that the proposed turbine blade has good performance characteristics.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(4):041203-041203-7. doi:10.1115/1.4042852.

Removal of decay heat with nonavailability of active systems is a safety issue especially during station blackout (SBO) in a light water reactor. Passive systems are being incorporated in the new designs of nuclear reactors for this purpose. Some of the advanced reactors such as Indian advanced heavy water reactor (AHWR) have dedicated isolation condensers (ICs) which are submerged in large water pool called gravity driven water pool (GDWP). These ICs remove decay heat from the core by natural circulation cooling and dissipate it to the GDWP by natural convection. There is a concern that cracks may develop in the GDWP if a large seismic event similar to Fukushima type occurs. In that case, the pool water is lost and it can threaten the core coolability because of loss of heat sink. In AHWR, the cracks in the water pool leads to the relocation of the water of the pool to the reactor cavity. Feeders of AHWR are positioned in the reactor cavity. Thus, the water relocated in the cavity, will eventually submerge the feeders and these submerged feeders have the potential to remove the decay heat of the core. However, the feeders are located at a lower elevation as compared to the core, and hence, there is concern on the heat removal capability by the submerged feeders by natural convection. To understand this aspect and to establish the core coolability under the above-mentioned conditions, experiments were performed in a full-scale test facility of AHWR. Experiments showed that the decay heat can be safely removed in natural circulation mode of cooling with heat sink located at lower elevation than the heat source.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(4):041204-041204-9. doi:10.1115/1.4042500.

MCrAl (M = Fe, Ni, or Co) alloys have exceptional corrosion and oxidation resistance and can be used as both oxidation resistant structural materials and coatings. As coatings, they protect high temperature steels or Ni based alloys by forming a dense alumina layer on the surface and thus impeding further oxidation. In order to assess its potential usage as an overlay coating on components used in supercritical water-cooled nuclear reactors (SCWRs), an Fe-2 3Cr-5Al alloy in the form of wire was tested under two different super-heated steam (SHS) conditions (625 °C and 800 °C) and also in supercritical water (SCW) (625 °C and 26 MPa), for 500 h. The corrosion behavior of samples was assessed by measuring the weight change per unit surface area and by examining the surface, cross-sectional microstructure and the phase compositions using scanning electron microscopy (SEM) and X-ray diffraction (XRD). The tested samples showed different oxidation behavior after exposure to these three conditions. SEM and XRD results showed that FeCrAl has the ability to form protective Al- and Cr-containing oxide(s) under all three conditions. Based on the findings, it is concluded that the oxidation behavior of Fe–23Cr–5Al is highly influenced by pressure and temperature within the range of testing conditions. SHS exposure at low temperature led to greater weight gain while that in SCW resulted in weight loss. Overall, its performance is better under SHS conditions compared to CoCrWC S16 but worse under the SCW condition.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(4):041205-041205-9. doi:10.1115/1.4043198.

Decay heat removal for prolonged period of station blackout (SBO) is a safety concern of the nuclear reactors. Aftermath of Fukushima, safety evaluation (performance under severe conditions: stress test) of the reactors was carried out worldwide. It includes establishment of grace period of the reactors. Similar exercises for advanced heavy water reactor (AHWR) were also performed and the design of AHWR was established for its robustness against such events. Decay heat removal during extended SBO is such a condition to be qualified. In this regard, experiments in the integral test loop (ITL), a full scale test facility of AHWR, were conducted for continuous 7 days of extended SBO. Experiment was started with 6.8 MPa as the initial reactor pressure and decay heat removal was demonstrated for 7 days of SBO by passive means. It is observed that the pressure falls down to 1 MPa in 3 h. The design of AHWR was evaluated from safety critical aspects during such an event experimentally. During this event, the clad surface temperature was found to be well within safe limits of operations. As a result of this experiment, it can be concluded that the design of AHWR is capable to remove decay heat for 7 days of SBO with sufficient safety margins.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(4):041206-041206-7. doi:10.1115/1.4043106.

To minimize the potential risk of design extension conditions (DEC) with core meltdown, some advanced reactors employ ex-vessel core catchers which stabilize and cool the corium for prolonged period by strategically flooding it. This paper describes the coolability of the melt pool and ablation process in a scaled down ex-vessel core catcher employing sacrificial material which reduces the specific volumetric heat, temperature, and density of the melt pool. To understand these phenomena, a simulated experiment was carried out. The experiment was performed by melting about 500 kg of corium simulant using thermite reaction at about 2500 °C. The bricks of oxidic sacrificial material were arranged in the core catcher vessel which was surrounded by a tank filled with water up to a certain level. After the time required for melt inversion, water was introduced to flood the test section from the top. The melt pool temperatures were monitored at various locations using “K” and “C” type thermocouples to obtain ablation depth at different elevations with time. The results show that the coolability of the molten pool in the presence of water for the present geometry is achievable with outside vessel temperatures not exceeding 100 °C. A ceramic stable crust was observed at the top surface of the melt pool, which prevented water ingression into the molten corium. The ablation rate was found to be maximum at the lower corners of the brick arrangement with the maximum value being 0.75 mm/s. An average rate of about 0.18 mm/s was obtained in the brick matrix.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(4):041207-041207-11. doi:10.1115/1.4043681.

This paper presents the modeling approach of a multipurpose simulation tool called gas turbine Arekret-cycle simulation (GT-ACYSS); which can be utilized for the simulation of steady-state and pseudo transient performance of closed-cycle gas turbine plants. The tool analyzes the design point performance as a function of component design and performance map characteristics predicted based on multifluid map scaling technique. The off-design point is analyzed as a function of design point performance, plant control settings, and a wide array of other off-design conditions. GT-ACYSS can be a useful educational tool since it allows the student to monitor gas path properties throughout the cycle without laborious calculations. It allows the user to have flexibility in the selection of four different working fluids, and the ability to simulate various single-shaft closed-cycle configurations, as well as the ability to carry out preliminary component sizing of the plant. The modeling approach described in this paper has been verified with case studies and the trends shown appeared to be reasonable when compared with reference data in the open literature, hence, can be utilized to perform independent analyses of any referenced single-shaft closed-cycle gas turbine plants. The results of case studies presented herein demonstrated that the multifluid scaling method of components and the algorithm of the steady-state analysis were in good agreement for predicting cycle performance parameters (such as efficiency and output power) with mean deviations from referenced plant data ranging between 0.1% and 1% over wide array of operations.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(4):041301-041301-12. doi:10.1115/1.4042120.

The clearance between fuel rods is maintained by spacer grid or helical wire wrap. Thermal-hydraulic characteristics inside fuel rod bundle are strongly influenced by the spacer grid geometry and the bundle pitch-to-diameter (P/D) ratio. This includes the maximum fuel temperature, critical heat flux, as well as pressure drop through the fuel bundle. An understanding of the detailed structure of flow mixing and heat transfer in a fuel rod bundle geometry is therefore an important aspect of reactor core design, both in terms of the reactor's safe and reliable operation, and with regard to optimum power extraction. In this study, computational fluid dynamics (CFD) simulations are performed to investigate isothermal turbulent flow mixing and heat transfer behavior in 4 × 4 rod bundle with twist-vane spacer grid with P/D ratio of 1.35. This work is carried out under International Atomic Energy Agency (IAEA) co-ordinated research project titled as “Application of Computational Fluid Dynamics (CFD) Codes for Nuclear Power Plant Design.” CFD simulations are performed using open source CFD code OpenFOAM. Numerical results are compared with experimental data from Korea Atomic Energy Research Institute (KAERI) and found to be in good agreement.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(4):041302-041302-7. doi:10.1115/1.4042364.

China Fusion Engineering Test Reactor (CFETR) is a test tokamak reactor being designed in China to bridge the gap between International Thermonuclear Experimental Reactor (ITER) and future fusion power plant. As one of the candidates, a water-cooled solid breeder blanket based on pressurized water and supercritical water conditions were proposed. In the concept, multiplying layers separated by three breeding layers were designed and optimized for higher tritium breeding ratio (TBR) and uniform heat distribution. This blanket adopts the Li2TiO3 lithium ceramic pebbles as the breeder, while beryllium pebbles as the neutron multiplier. In this paper, the thermal and fluid dynamic analyses of the optimized blanket on both water conditions were performed by numerical simulation, to discuss thermo-hydraulic performance of the blanket using pressurized water/supercritical water as its coolant. At first, the neutronic analysis was performed and based on the typical outboard equatorial blanket. Then, thermal and fluid dynamic analysis of the 3D model was carried out by CFX with fluid–solid coupling approach. It was found that the blanket can be effectively cooled on both water conditions, certified the feasibility of the blanket design with pressurized/supercritical water conditions. It indicated that supercritical water blanket had smaller safety margin than pressurized water blanket, but supercritical water blanket would lead to higher outlet temperature, thermal conductivity, and heat exchange efficiency also. In addition, the beryllium fraction was considered as one of the dominant factor, which leading a higher TBR in our simulations.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(4):041303-041303-12. doi:10.1115/1.4043098.

Recuperator is one of the most important components in supercritical carbon dioxide (S-CO2) Brayton cycle, and the printed circuit heat exchanger (PCHE) has been considered as a promising candidate due to its high efficiency and compactness. The airfoil fin (AFF) PCHE has higher thermal-hydraulic performance than conventional zigzag channel PCHE. However, it also suffers from serious local flow resistance caused by the impact area. Two types of new slotted fins (SFs) based on AFFs including longitudinal slot fins (LSFs) and herringbone slot fins (HSFs) are proposed to release the effect of the impact area. The results show that both LSFs and HSFs can significantly reduce the flow resistance in the channel. Meanwhile, the SFs also show higher thermal performance due to the heat transfer area enhancement by the slots. The LSF channel can be considered as a promising candidate in some energy conversion systems due to its good hydraulic performance, while the HSF channel would behave more efficiently such as in refrigeration cycles due to its high thermal performance. Finally, the field synergy principle is employed to discuss the flow drag reduction in SF channels.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(4):041401-041401-10. doi:10.1115/1.4043294.

A study of transverse buckling effect on the characteristics of nuclides burnup wave in multiplying media (cylindrical geometry) has been carried out. The burnup wave is characterized in terms of velocity of propagation, transient length (TL), and transient time (TT) in establishing the burnup wave and full width at half maximum (FWHM) in the established region of the wave. The uranium–plutonium fuel cycle is considered. The sensitivity of the results is studied for different radial buckling led leakage of neutrons. It is discovered that the velocity of the wave increases with the increase in the radius of the cylinder (i.e., reduction in the transverse buckling and hence increase in radial neutron leakage). FWHM is relatively insensitive to radial neutron leakage. The transient time and transient length are very large for smaller radius; these decrease with the increase in radius. The study provides insight on the build-up of burnup wave in the neutron multiplying media and brings out the importance of transverse buckling led radial neutron leakage on the characteristics of fuel burnup wave in multiplying media.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(4):041402-041402-10. doi:10.1115/1.4043100.

Methodological processes for nuclear power plant (NPP) pressure vessels' (PV) neutron fluence rate determination take the form of experimental measurement or theoretical calculations. However, the process of experimental measurement takes longer periods, as it requires the incorporation of surveillance capsules into a PV system undergoing normal NPP operation. Therefore, strong reliance on computation and modeling of radiation-induced degradation is given much attention. In this work, the VENUS-3 benchmark has been analyzed using SuperMC code, with the intention of validating SuperMC for accurate reactor neutronics; dosimetry response calculations for in-core/ex-core structural components, particularly with respect to the VENUS-3 configuration type pressurized water reactors (PWRs). In this work, complete three-dimensional (3D) geometry including the source modeling for VENUS-3 facility has been developed with SuperMC. Neutron transport and calculations of equivalent fission flux for the experimental target quantities, 115In (n, n′), 58Ni (n, p), and 27Al (n, α), are also achieved. The calculation results show good agreement with the experimental measurement. The greater majority of the calculated values (C/E) were within the required accuracy of ±10% for reactor components' dosimetry calculations. Most of the calculated values were contained within 5% deviation from the experimental data. Additional calculations and detailed analysis for fast neutron flux distribution and iron displacement per atom rate (dpa/s), including the characteristic effect of partial length shielded assembly (PLSA) on VENUS-3 core barrel, are also discussed. It is therefore evidenced that the effectiveness of SuperMC code for in-core/ex-core reactor neutronics computations has been convincingly demonstrated through the VENUS-3 benchmark testing.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(4):041601-041601-8. doi:10.1115/1.4043108.

In some of the older design of pressurized heavy water reactors (PHWRs), such as in Rajasthan Atomic Power Station (RAPS) and Madras Atomic Power Station (MAPS), in case of a severe accident, the debris/corium may cause failure of the dump port of calandria and relocate into the dump tank. The sensible and decay heat of debris/corium makes the heavy water in dump tank to boil off leaving the dry debris in dump tank. The dry debris remelt with time and the molten corium, thus, formed has the potential to breach the dump tank and move into the containment cavity, which is highly undesirable. Hence, as an accident management strategy, water is being flooded outside the dump tank using fire water hook-up lines to remove the heat from corium to cool and stabilize it and terminate the accident progression, similar to in vessel retention. However, the question is “is the molten corium coolable by this technique.” The coolability of the molten corium in dump tank as in the reactor is assessed by conducting experiments in a scaled facility using a simulant material having comparable thermophysical properties with that of corium. Melting of dry debris resting on dump tank bottom marks the beginning of the experimental investigation for present analysis. Decay heat is simulated by a set of immersed heaters inside the melt. Temperature profiles at different locations in dump tank and in the melt pool are obtained as a function of time to demonstrate the coolability with decay heat. Large temperature gradient is observed within the corium, involving high melt center temperature and low tank wall temperature suggesting formation of crust which insulates the dump tank wall from hot corium.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(4):041602-041602-6. doi:10.1115/1.4043847.

Prediction of evaporation rates from spent fuel pools of nuclear power plants in normal and postaccident conditions is of great importance for the design of safety systems. A severe accident in 2011 Fukushima nuclear power plant caused failure of cooling systems of its spent fuel pools. The postaccident evaporation from the spent fuel pools of Fukushima units 2 and 4 is compared to a model based on analogy between heat and mass transfer which has been validated with a wide range of data from many water pools including a spent fuel pool. Calculations are done with two published estimates of fuel decay heat, one 25% lower than the other. The model predictions are close to the evaporation using the lower estimate of decay heat. Other relevant test data are also analyzed and found in good agreement with the model.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(4):041701-041701-8. doi:10.1115/1.4042796.

The low-pressure depressurization system (LPDS) of advanced passive pressurized water reactors (PWRs) is designed to provide depressurization of the reactor coolant system during a small break loss-of-coolant accident (LOCA). The liquid entrainment to the LPDS is important for the safety case of the advanced passive PWRs due to the significant increase of the pressure loss and the depressurization rate versus mass loss characteristics. The existing experimental researches on the liquid entrainment at LPDS have been reviewed, and the intermittent entrainment mechanism and the continuous entrainment mechanism are identified. The intermittent entrainment is closely related to the flow regime transition from the horizontal stratified flow to the intermittent flow in the hot leg where the LPDS port is located. The horizontal stratification model previously developed for the FULL SPECTRUM LOCA evaluation model has been assessed against the entrainment onset data in the available LPDS entrainment experiments, i.e., the ATLATS air–water experiment, the ADETEL air–water and steam–water experiments, and the full-scale FATE air–water experiment. The prediction matches the measure data well especially in the full-scale FATE experiments. The comparison results also confirmed the scalability of the horizontal stratification model with the applicability of the horizontal stratification criterion to the full-scale PWR condition. The uncertainty factors that impact the depressurization system entrainment onset are discussed for the future improvements. This work provides the direction to accurately model the entrainment onset for LPDS and improve the simulation of LOCA in advanced passive PWRs.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(4):042001-042001-10. doi:10.1115/1.4042117.

The thin shell design code RCC-MR is used for sodium-cooled fast breeder reactor components operating at high temperatures. Thin shells from such applications can be designed using linear elastic buckling analysis, following procedures given in RCC-MR. For human safety, such procedures can and should be examined by the broader scientific community. Among such procedures, RCC-MR provides three alternative methods to quantify an imperfection value; and that value is used in subsequent calculations to determine safe loads. Of these methods, the third seems potentially nonconservative for some situations. Here, we examine that third method using detailed numerical examples. These examples, found by trial and error, are the main contribution of this paper. The first example is a nonuniform cylindrical shell closed with a spherical endcap under external pressure. The second is a cylinder with an ellipsoidal head under internal pressure. The third is an L-shaped pipe with an end load. In all three cases, the new computed imperfection quantity is found to be surprisingly small compared to the actual value used for computations (e.g., 25 times smaller), and in two cases, the result is insensitive to the actual imperfection. We explain how the three examples “trick” the imperfection quantification method in three different ways. We suggest that this imperfection quantification method in RCC-MR should be re-examined. The primary value of our paper lies not in new mechanics, but in identifying unexpected ways in which a particular step in shell design using RCC-MR could be potentially nonconservative.

Topics: Stress , Buckling , Shells , Design
Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(4):042201-042201-13. doi:10.1115/1.4042710.

Three-dimensional (3D) modeling of magneto-inertial fusion (MIF) is at a nascent stage of development. A suite of test cases relevant to plasma liner formation and implosion is presented to present the community with some exact solutions for verification of hydrocodes pertaining to MIF confinement concepts. MIF is of particular interest to fusion research, as it may lead to the development of smaller and more economical reactor designs for power and propulsion. The authors present simulated test cases using a new smoothed particle hydrodynamic (SPH) code called SPFMax. These test cases consist of a total of six problems with analytical solutions that incorporate the physics of radiation cooling, heat transfer, oblique-shock capturing, angular-momentum conservation, and viscosity effects. These physics are pertinent to plasma liner formation and implosion by merging of a spherical array of plasma jets as a candidate standoff driver for MIF. An L2 norm analysis was conducted for each test case. Each test case was found to converge to the analytical solution with increasing resolution, and the convergence rate was on the order of what has been reported by other SPH studies.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(4):042202-042202-8. doi:10.1115/1.4042707.

The present work includes thermal hydraulic modeling and analysis of loss of heat sink (LOHS) accident for the ITER divertor cooling system. The analysis is done for the new design of full tungsten divertor. The new design is also analyzed for different local heat loads ranging from 10 MW/m2 to 20 MW/m2 (while maintaining the total heat load 200 MW) under the steady-state fluid conditions. The LOHS event is selected since divertor is the most sensitive component to loss or reduction in coolability of divertor primary heat transport system (DV-PHTS) loop as it receives large heat flux from plasma. The main objective of this analysis is to find margins to unwanted conditions like overstress temperatures of structure and elevated water level in the pressurizer. The analysis is done by modified thermal hydraulic code RELAP/SCDAPSIM/MOD 4.0. The results obtained are compared with the results of old divertor design which uses carbon fiber composite (CFC) layer to show that how the new design of divertor behaves compared to the older design under the accident scenario. A detailed model of DV-PHTS loop and its ancillary system is presented. The model includes promotional integral differential (PID) controller for controlling the pressurizer heater and spray system. A detailed pump model is also included in the present analysis which was previously used as a time-dependent junction. The analysis shows that under the accident scenario, (a) the divertor structure temperature at the critical sites (inner vertical target (IVT) and outer vertical target (OVT)) is always within the design limit and does not affect the structural integrity of the divertor. (b) The water level in the pressurizer increases moderately and finely controlled by the PID controller, and pressurizer safety valve does not open.

Commentary by Dr. Valentin Fuster

Technical Brief

ASME J of Nuclear Rad Sci. 2019;5(4):044501-044501-7. doi:10.1115/1.4043814.

We have calculated the gamma and X-ray shielding parameters such as mass attenuation coefficient, half value layer (HVL), tenth value layer (TVL), specific gamma ray constant, effective atomic number, and buildup factors in various steels. By studying these X-ray and gamma interaction parameters, we have selected the best steel which can be used for the X-ray and gamma shielding material. The steel type 20Mo-4 is having higher values of mass attenuation coefficient, specific gamma ray constant, effective atomic number, and buildup factor and smaller values of HVL and TVL. A detail analysis of X-ray/gamma interaction in the different steels reveals that the steel type (S15) 20Mo-4 is good absorption of both X-ray/gamma radiations.

Commentary by Dr. Valentin Fuster

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