ASME Conference Presenter Attendance Policy and Archival Proceedings

2017;():V009T00A001. doi:10.1115/ICONE25-NS9.

This online compilation of papers from the 2017 25th International Conference on Nuclear Engineering (ICONE25) represents the archival version of the Conference Proceedings. According to ASME’s conference presenter attendance policy, if a paper is not presented at the Conference by an author of the paper, the paper will not be published in the official archival Proceedings, which are registered with the Library of Congress and are submitted for abstracting and indexing. The paper also will not be published in The ASME Digital Collection and may not be cited as a published paper.

Commentary by Dr. Valentin Fuster

Student Paper Competition

2017;():V009T15A001. doi:10.1115/ICONE25-66001.

This paper analyzes world steam generator (SG) operational experience (OPEX), research and development (R&D) for operational CANDU SGs. Emphasis is placed on design configuration, materials, and chemistry. SG OPEX from various databases is recounted to identify performance trends, active degradation mechanisms and common failure modes. A summary of CANDU and pressurized water reactor (PWR) SG OPEX including tube plugging statistics by material and degradation mechanism is provided. A discussion of ongoing SG-related research initiatives is also included. It was found that SGs with Monel-400 tubing and carbon steel tube support would perform well if properly maintained although Incoloy-800 tubing and stainless steel tube support is the current preferred design for CANDU, regarding Inconel-600 as an interim solution. All volatile treatment has also been adopted as the preferred SG chemistry for the CANDU fleet. Lessons learned from OPEX are highlighted and further international collaboration on the study of plant ageing, SG integrity, and nuclear safety is encouraged.

Topics: Boilers
Commentary by Dr. Valentin Fuster
2017;():V009T15A002. doi:10.1115/ICONE25-66009.

Due to great deal of operation experience and technology accumulation, sodium cooled fast reactor (SFR) is the most promising among the six Generation IV reactors, which has advantages of breeding nuclear fuel, transmuting long-lived actinides and good safety characteristics. Thermal-hydraulic computer codes will have to be developed, verified, and validated to support the conceptual and final designs of new SFRs. However, work on developing thermal hydraulic analysis code for SFR is very limited in China, while the common software RELAP5 MOD3 is unable to analyze liquid metal systems. So the modified RELAP5 MOD3.2 is being considered as the thermal-hydraulic system code to support the development of the SFRs.

The thermodynamic and transport properties of sodium liquid and vapor have been implemented into the RELAP5 MOD3.2 code, as well as the specific heat transfer correlations for liquid metal. The sodium liquid properties use polynomial equations based on data obtained from Argonne National Laboratory, and the vapor is assumed to be perfect gas. The property equations are acceptably accurate for analysis of SFR, especially for single-phase liquid. New files are added to the fluids directory to generate property tables for new working fluid, which are similar to the table interpolation subroutines for light and heavy water in the original file directory. The method of code modifications are universal for other working fluids and will not affect the code original performance. Some basic verification work for the modified code are carried out. The steam generator of CEFR is analyzed to verify the modified code. The calculated results show that all the water will boil off in the evaporator and the calculated results are in good agreement with the design values. By using modified RELAP5 to model the primary loop of EBR-II fast reactor, the SHRT-17 PLOF test was analyzed. The results show that the natural circulation can be established in the EBR-II primary system after main pumps off to remove the core decay residual heat effectively, and the peak temperature under the safety limits. Moreover, the results computed in this work compared well with the test experimental data for the steady state condition. During the transients, the changing trends of temperature and pressure are similar to experimental data. The discrepancies between calculation and experiment are considered acceptably which need to be improved in the future work. Our work could demonstrate the capability and reliability of the modified RELAP5 for the analysis of SFRs further.

Commentary by Dr. Valentin Fuster
2017;():V009T15A003. doi:10.1115/ICONE25-66031.

Steam generator (SG) water level system is a highly complex nonlinear time-varying system. It is complicated at low power levels due to shrink and swell phenomena which must be considered for plant safety and availability. To improve the transient performance of the SG level subject to power adjustments, an innovative set-point function method is put forward in this paper. The set-point functions based on the inverse-control theory and the swell and shrink effect which generate a desirable reference input to the widespread cascade Proportional Integral Derivative (PID) controller of the level control system respectively. The set-point function can apply appropriate control to the feed-water flow rate duly depended on the pivotal time between the power adjustment decision and the real start time of adjustment. Finally, comparative simulation is carried out under the same condition of power adjustment. The simulation results demonstrate that the water level control system added set-point functions can restrain the disturbance and improve the transient performance effectively. The method added the Inverse Control-Based Set-Point (ICSP) function can achieve better control performances than the swell-based set-point (SBSP) function.

Topics: Boilers , Water
Commentary by Dr. Valentin Fuster
2017;():V009T15A004. doi:10.1115/ICONE25-66039.

The function of steam separator is to remove the small droplets carried by the vapor stream and to provide qualified saturated vapor for the steam turbine in the nuclear power station. The separating characteristics of the steam-water separation plant are of vital importance to the safe operation, economy as well as reliability of the power station. In order to satisfy the requirement of power increase of large nuclear power station as well as space compaction of the vessel power plant, the steam vapor quality must be improved, which requires that the steam-water separator has better separating function to make sure that it can provide the qualified steam on the condition of higher steam pressure, power load as well as circulating ratio. There are many complex phenomena when the droplet moves in the steam-water separating plant, including the droplet emergence, the droplet moving with steam vapor, the collision between droplets and with solid wall, evaporation. It is a good way to study the steam-water separating characteristics for the microcosmic behavior of the droplet. Thus, in order to know the droplet evaporation characteristics in the steam-water separator, the static droplets phase transformation model under the pressure variation condition is built according to the physical phenomenon description and mechanism comprehension when the droplet moves with the steam vapor in the steam-water separation plants. This model is solved by the typical four steps Runge-Kutta method and validated by comparing with the experimental results. Then, the influence of working pressure as well as pressure difference between the droplet surface and the environment on the static droplet evaporation characteristics is conducted. The simulation results show that the working pressure and pressure difference have great impact on the static droplet evaporation characteristics. With the increase of the working pressure, the droplet evaporation rate becomes slower, that is because the physical property parameters of the water vapor and water become closer to each other and the self-diffusion coefficient of the water vapor as well as the evaporation condensation coefficient become smaller, which results in the droplet evaporation rate becomes slower. When the pressure difference between the droplet surface and the environment rises, the droplet evaporates faster, that is because the vapor velocity around the droplet becomes larger and the droplet evaporates faster. These results of the simulation can lay the foundation for subsequent study of the droplet evaporation characteristics when the droplet moving in the separating plants and for the droplet fast evaporation characteristics when the environment pressure changes fast.

Topics: Pressure , Drops , Evaporation
Commentary by Dr. Valentin Fuster
2017;():V009T15A005. doi:10.1115/ICONE25-66191.

In this paper, the capability of uncertainty propagations of the nuclear-data to the reactor-physics calculations has been implemented in our home-developed code NECP-UNICORN based on the statistical sampling method (SSM). The “two-step” scheme has been applied in NECP-UNICORN to perform the uncertainty analysis for the reactor-physics calculations. For the lattice calculations, the nuclear-data uncertainties are propagated to the few-group constants; then for the core simulations, the uncertainties of the multiplication factor and power distributions introduced by the few-group constants’ uncertainties can be quantified. Applying the NECP-UNICORN code, uncertainty analysis has been performed to the BEAVRS benchmark problem at the Hot Zero Power (HZP) conditions, with situations of All Rod In (ARI) and All Rod Out (ARO). From the numerical results, it can be observed that for the multiplication factors of the core simulations, the relative uncertainties are about 5.1‰ for the ARO situation and 5.0‰ for the ARI situation, which are the same magnitude of the relative uncertainties of the fuel assemblies; for the radial power distributions, the relative uncertainties can up to be 4.27% as the maximum value and 2.08% as the RMS value for the ARO situation, and 6.03% as the maximum value and 2.37% as the RMS value for the ARI situation.

Commentary by Dr. Valentin Fuster
2017;():V009T15A006. doi:10.1115/ICONE25-66250.

The management of long-lived radionuclides in spent fuel is a key issue to achieve the closed nuclear fuel cycle and the sustainable development of nuclear energy. Partitioning-Transmutation is supposed to treat efficiently the long-lived radionuclides. Accordingly, the study of transmutation for long-lived Minor Actinides (MAs) is a significant work for the post-processing of spent fuel.

In the present work, the transmutations in Pressurized Water Reactor (PWR) Mixed OXide (MOX) fuel are investigated through the Monte Carlo based code RMC. Two kinds of MAs are incorporated homogeneously into two initial concentrations MOX fuel assembly.

The results indicate an overall nice efficiency of transmutation in both initial MOX concentrations, especially for two MAs primarily generated in the UOX fuel, 237Np and 241Am. In addition, the inclusion of 237Np has no large influence on other MAs, while the transmutation efficiency of 237Np is excellent. The transmutation of MAs in MOX fuel depletion is expected to be an efficient nuclear spent fuel management method.

Commentary by Dr. Valentin Fuster
2017;():V009T15A007. doi:10.1115/ICONE25-66334.

In the numerical analysis of sodium fire event following coolant leakage in a sodium-cooled fast reactor, sufficient understanding is needed for the liquid jet atomization to estimate the combustion rate accurately. In the present work, a liquid jet was emanated vertically downward from a circular nozzle onto a liquid film formed on a horizontal plate. The droplets produced at the impact point were investigated. The splash ratio (the ratio of the mass of the splashed droplets to the jet flow rate) was measured under varied experimental conditions of nozzle diameter, fall height and flow rate of liquid jet. The experimental result depended significantly on the morphology of liquid jet upon impact. The splashing rate was negligibly small when the liquid jet impinged as the continuous jet but a significant amount of liquid was splashed when the liquid jet impinged as the broken jet. Thus, we developed a method to estimate the impact frequency. It was shown that the splash ratio can be correlated well if the impact frequency is included in the correlation.

Commentary by Dr. Valentin Fuster
2017;():V009T15A008. doi:10.1115/ICONE25-66391.

The continuous generation of graphite dust particles in the core of a High Temperature Reactor (HTR) is one of the key challenges of safety during the operation. The graphite dust particles emerge from relative movements between the fuel elements or from contact to the graphitic reflector structure and could be contaminated by diffused fission products from the fuel elements. They are distributed from the reactor core to the entire reactor coolant system. In case of a depressurisation accident, a release of the contaminated dust into the confinement is possible. In addition, the contaminated graphite dust can decrease the life cycle of the coolant system due to chemical interactions.

On the one hand, the knowledge of the behaviour of graphite dust particles under HTR conditions using helium as the flow medium is a key factor to develop an effective filter system for the discussed issue. On the other hand, it also provides a possibility to access the activity distribution in the reactor. The behaviour can be subdivided into short-term effects like transport, deposition, remobilization and long-term effects like reactions with material surfaces.

The Technische Universität Dresden has installed a new high-temperature test facility to study the short-term effects of deposition of graphite dust particles. The flow channel has a length of 5m and a tube diameter of 0.05m. With helium as the flow medium, the temperature can be up to 950 °C in the channel center and 120 °C on the sample surface, the Reynolds number can be varied from 150 up to 1000. The particles get dispersed into the accelerated and heated flow medium in the flow channel. Next, the aerosol is passing a 3 m long adiabatic section to ensure homogenous flow conditions. After passing the flow straightener, it enters the optically accessible measurement path made from quartz glass.

In particular, this test facility offers the possibility to analyse the influence of the thermophoretic effect separately. For this, an optionally cooled sample can be placed in the measuring area. The thickness of the particle layer on the sample is estimated with a 3D Laser scanning microscope. The particle concentration above the sample is measured with an aerosol particle sizer (APS). Particle Image Velocimetry (PIV) detects the flow-velocity field and provides data to estimate the shear velocity. In combination with the measured temperature-field, all necessary information for the calculation of the particle deposition and particle relaxation time are available.

The measurements are compared to results of theoretical works from the literature. The experimental database is relevant especially for CFD-developers, for model development, and model verification. A wide range of phenomena like particle separation, local agglomeration of particles with a specific particle mass and selective remobilization can be explained in this way. Thus, this work contributes to a realistic analysis of Nuclear Safety.

Commentary by Dr. Valentin Fuster
2017;():V009T15A009. doi:10.1115/ICONE25-66490.

A new method has been presented in this study, for evaluating the functional reliability of passive systems based on the fuzzy method. A case study has been presented, in order to analyze the passive residual heat removal system (PRHRS) of AP1000. The RELAP5 is used to analyze the thermal-hydraulic behavior. However, the failure criterion may not be suitable, since the local saturated boiling temperature is not only related to the pressure, but also related to the local surface conditions. In addition, it is not reasonable to categorize the failure, based on a single numerical value. There are no essential differences among these values which are near the critical numerical value. Hence, it is different to use the classical theory for calculating the failure probability. However, the fuzzy method has been founded to be appropriate for such a case. The failure criteria parameter is regarded as a fuzzy number, which can be described by a membership function. Finally, the failure probability of PRHRS is obtained as 7.26×10−8 which is well within the range of classical failure probability.

Topics: Reliability
Commentary by Dr. Valentin Fuster
2017;():V009T15A010. doi:10.1115/ICONE25-66495.

More and more attentions have been paid to the impact of the earthquake on the reactor thermal hydraulics after the Fukushima accident. In order to figure out the seismic effect on the flow in primary loop, the natural circulation in primary loop of CPR1000 influenced by earthquake is modeled by learning from the study on natural circulation under ocean conditions. Natural circulation ability at earthquake, which includes longitudinal earthquake and transversal earthquake, is dynamically simulated by a program written with MATLAB, respectively. Besides, the safety of the system is discussed based on the calculated results.

The research results show that, once the longitudinal earthquake occurs, in the earlier stage of earthquake, the peak and valley values of the coolant mass flow rate decrease gradually and the amplitude will be a constant as time goes on because of flow fluctuation in seismic condition. On the contrary, as for transversal earthquake, the peak and valley values of the coolant mass flow rate increase gradually in the earlier stage of earthquake. However, the amplitude will be a constant as time goes on, just like that in the longitudinal earthquake.

For both of the two conditions, they have the same characteristics. The constant mentioned above grows up with the increase of the earthquake acceleration peak value and the period; the fluctuation period of mass flow rate is generally the same as the period of earthquake acceleration and the mass flow rate changes steadily; the average mass flow rate decreases slightly with the increase of the earthquake acceleration peak value and the period.

Both of the vibration period and the peak of seismic wave have disturbing effect on reactor coolant system natural circulation. And the increase of the peak of the earthquake acceleration as well as the period of the earthquake acceleration will reduce the natural circulation ability. In the range of research parameters, the system is considered to be safe according to the calculation results of core coolant temperature.

Topics: Modeling , Earthquakes
Commentary by Dr. Valentin Fuster
2017;():V009T15A011. doi:10.1115/ICONE25-66511.

PWR core phenomena can be simulated and predicted more precisely and in more details with high-fidelity neutronics and thermal-hydraulics coupling calculations. An internal coupling between a newly developed high-fidelity neutronics code NECP-X and the sub-channel code SUBSC has been realized. In order to verify the NECP-X/SUBSC coupling system, another high-fidelity neutronics and thermal-hydraulics coupling system OpenMC/SUBSC was developed through external coupling method. Both coupling systems were applied to a simplified PWR 3×3 pin cluster case. The numerical result shows good agreement in both eigenvalue and normalized axial power distribution for a selected pin, demonstrating the success of the internal coupling of NECP-X and SUBSC.

Commentary by Dr. Valentin Fuster
2017;():V009T15A012. doi:10.1115/ICONE25-66525.

During off-normal conditions, the temperature transient experienced by a CANDU fuel bundle may lead to deformation of the bundle via fuel element bowing between the spacer pads due to sub-channel temperature gradients, or fuel element sagging under gravity. The resulting deformation could impact the coolant flow distribution through the bundle, causing further degradation of the fuel cooling. Moreover, the deformation of the fuel elements may also lead to contact with the pressure tube, leading to localized heating that may compromise the integrity of the pressure tube. The objective of this work is to create a 3D finite element model to analyze bundle deformation behaviour under transient conditions. The methodology and preliminary results are presented herein.

Commentary by Dr. Valentin Fuster
2017;():V009T15A013. doi:10.1115/ICONE25-66528.

The paper presents results of a study on flow and temperature fields in bare tubes cooled with SuperCritical Water (SCW). This study is based on a Computational Fluid Dynamics (CFD) simulation with the FLUENT code for upward flows in vertical tubes with heated length of 4 m and an inner diameter of 10 mm. Operating conditions were: Mass flux – G ≈ 500 and 1000 kg/m2s; heat flux – q = 189 – 826 kW/m2; and inlet coolant temperature – Tin = 320–360°C.

CFD predictions were compared with experimental data in this study. All three heat-transfer regimes: 1) normal heat transfer; 2) improved heat transfer; and 3) deteriorated heat transfer; were considered. The obtained results show that within normal and improved heat transfer CFD predicts experimental values reasonably well. However, within conditions of deteriorated heat transfer CFD predictions are less satisfactory.

The CFD outcomes of the heat flux effect on the flow and heat transfer of SCW are presented. Specifics of flow within the pseudocritical region (i.e., approximately ±25°C around a pseudocritical point) are discussed. The buoyancy effect is investigated by axial velocity profiles at the medium mass flux of 500 kg/m2s and heat flux of 287 kW/m2.

Commentary by Dr. Valentin Fuster
2017;():V009T15A014. doi:10.1115/ICONE25-66542.

In order to explore and analyze the heat transfer characteristics in narrow rectangular channel, experiments on local single-phase heat transfer of natural circulation in a one-side heating narrow rectangular channel have been conducted under vertical and inclined condition. The thermotechnical parameters such as inlet temperature, heat flux and inclination angle varies during the experiments. The width of the flow channel is 40 mm and the narrow gap is 2 mm. It is heated from one side with a homogeneous and constant heat flux and the working medium is deionized water.

Based on the experimental results, under vertical condition, the driving force in the loop goes up and the Reynolds number also increases when the inlet temperature is elevated, which causes an increase in local Nusselt number. When the heat flux rises, the local Nusselt number increases and the heat transfer temperature difference increases. The local Nusselts number is influenced by entrance effect and the entrance region length is computed for laminar and turbulent flow. Under inclined condition, with the inclination angle from −30° to 30°, it is found that when the inclination angle is positive, the local Nusselt number in fully developed region is larger than that under vertical condition and increases with the angle value, even though the Reynolds number decreases by the effect of incline. This phenomenon is explained by giving an analysis of the natural convection, which is characterized by the normal Grashof number, in the direction perpendicular to the heating plat. Moreover, the variation of heat transfer is also interpreted on the basis of field coordination principle. However, when the inclination angle is negative, the heat transfer shows no obvious difference between vertical condition and inclined condition.

Topics: Heat transfer
Commentary by Dr. Valentin Fuster
2017;():V009T15A015. doi:10.1115/ICONE25-66563.

An experiment study is conducted to investigate the effect of transverse power distribution on the Onset of Nucleate Boiling (ONB) through a one-side heated narrow rectangular channel. Two test section are used to perform the experiment; uniform and non-uniform heated suction. The demineralized water is flowing in upward direction through the coolant channel with a thickness of 2.35 mm, a width of 54 mm, and a length of 300 mm. The experiment is carried out under different thermal power (0.5 kW – 6.5 kW) for the both test section. As well as, a wide variety of inlet subcooling and flow velocity are used as; 65−35 °C and 0.1–1.0 m/s, respectively. The wall temperature distribution of the heated plate is measured by 10 TCs for the uniformly heated test section, and 20 TC for the non-uniformly heated section. On the other hand, the ONB location is visualized via high speed camera, in which the ONB occurs near the edges for the non-uniformly heated section and occurs at the center of the heated surface for the uniformly power distribution. The results of the ONB heat flux and temperature in the non-uniformly heated section are compared against the one in the uniformly heated power. The results show the variety of the ONB location, ONB heat flux with the different power distribution. With the increase of the power, the ONB is shifted toward the inlet. On the other hand, the ONB for the non-uniform power distribution occurs near the edges at power lower than that the one in the uniformly power distribution. Also, the results are compared against the available correlations, such as Bergles and Rohsenow (1965), Jens and Lottes (1951), and Thom et al. (1965), as well as other experimental results done by several research institutes.

Commentary by Dr. Valentin Fuster
2017;():V009T15A016. doi:10.1115/ICONE25-66602.

A primary pipe rupture accident is one of the design-basis accidents of a Very-High-Temperature Reactor (VHTR). When a primary pipe rupture accident occurs, air is expected to enter into the reactor pressure vessel from the breach and oxidize in-core graphite structures. Therefore, it is important to understand the mixing processes of different kinds of gases in the stable and unstable stratified fluid layers. In particular, it is also important to examine the influence of localized natural convection and molecular diffusion on the mixing process from a safety viewpoint. Therefore, in order to predict or analyze the air ingress phenomena during a pipe rupture accident, it is important to develop a method for the prevention of air ingress during an accident.

We carried out experiments to obtain the mixing process of two-component gases and flow characteristics of localized natural convection. This study also investigated a control method for the natural circulation of air through the injection of helium gas. An experiment has been carried out to investigate a control method of natural circulation of air by injection of helium gas. The experimental apparatus consists of a reverse U-sharped vertical slot and a storage tank. One side-slot consists of the heated and cooled walls. The other side-slot consists of the two cooled walls. The dimensions of the vertical slots are 598 mm in height, 208 mm in depth, and 70 mm in width. Each two vertical slots were connected and were a reverse U-shaped passage. The dimensions of the connecting passage were 16 mm in height, 106 mm in depth, and 210 mm in length. The storage tank was connected to the lower part of the reverse U-shaped passage. The dimensions of the storage tank were 398 mm in length, 398 mm in depth, and 548 mm in width. The reverse U-shaped passage and the storage tank were separated by a partition plate. The wall and gas temperature were measured by a K-type thermocouple.

Experimental results regarding mixing process of two component gases in vertical fluid layer were as follows. The heavy gas was transported to the slot by the molecular diffusion and natural convection. As time elapses, natural circulation of heavy gas suddenly occurs through the reverse U-shaped slot. As a result of experiments, the onset time of natural circulation is affected by not only molecular diffusion coefficient but also the strength of natural convection. When the helium gas is injected into the channel, it is possible to control the natural circulation of air. The onset time of the reproduction of the natural circulation can be varied by changing the injection rate of the helium gas.

Commentary by Dr. Valentin Fuster
2017;():V009T15A017. doi:10.1115/ICONE25-66616.

Under normal refueling and emergency full-core offload condition, the fuel assembly is removed to the Spent Fuel Pool (SFP). Decay heat produced by the spent fuel is carried out by the cooling system. Active cooling method is adopted by the traditional PWR nuclear power plants, which means decay heat is taken away depending on forced circulation of the pump. However, the spent fuel pool, under accident condition, will lost the forced circulating cooling capacity, which will be a threat of for the fuel building safety. To study the thermal-hydraulic characteristics in the SFP missing the forced cooling, through CFX methodology and experiment, change of temperature and heat transfer coefficient of the wall of the heating tube at different heights were discussed, meanwhile the streamline chart and temperature contour were obtained as well. The present result indicated that under different power conditions, different height of water temperature increased at first and then trend to stable at saturation temperature. For a single 9*9 spent fuel assembly, water temperature at the higher height is higher than the lower at the same time, and water temperature at higher location reached a stable value more quickly. In addition, power value had a significant impact on the time of reaching saturation temperature, for example, 7000s is needed to reach saturation under 8.68KW condition while only 3000s under 16.12KW, which illustrates that fuel unload power is crucial to the SFP safety. Based on the experiment data and single phase calculation, heat transfer coefficient at different height of the heating tube decreased slowly at first, and then increased. Especially, heat transfer coefficient at the highest test point rapidly decreased at one point because of boiling crisis.

Commentary by Dr. Valentin Fuster
2017;():V009T15A018. doi:10.1115/ICONE25-66744.

After Fukushima Daiichi Nuclear Power Plant accident, alternative fuel-design to enhance tolerance for severe accident conditions becomes particularly important. Silicon carbide (SiC) cladding fuel assembly gain more safety margin as novel accident tolerant fuel. This paper focuses on the neutron properties of SiC cladding fuel assembly in pressurized water reactors. Annular fuel pellet was adopted in this paper. Two types of silicon carbide assemblies were evaluated via using lattice calculation code “dragon”. Type one was consisted of 0.057cm SiC cladding and conventional fuel. Type two was consisted of 0.089cm SiC cladding and BeO/UO2 fuel. Compared the results of SiC cladding fuel assembly neutronic parameters with conventional Zircaloy cladding fuel assembly, this paper analyzed the safety of neutronic parameters performance. Results demonstrate that assembly-level reactivity coefficient is kept negative, meanwhile, the numerical value got a relatively decrease. Other parameters are conformed to the design-limiting requirement. SiC kinds cladding show more flat power distribution. SiC cases also show the ability of reducing the enrichment of fuel pellets even though it has higher xenon concentration. These types of assembly have broadly agreement neutron performance with the conventional cladding fuel, which confirmed the acceptability of SiC cladding in the way of neutron physics analysis.

Commentary by Dr. Valentin Fuster
2017;():V009T15A019. doi:10.1115/ICONE25-66805.

Alpha radiation emitting radon (Rn) gas seepage into homes in the USA leads to over 21,000 annual lung cancer deaths (according to the US-Environmental Protection Agency, EPA) leading to mandatory monitoring for Rn throughout the USA. In the nuclear industry alpha emitting radionuclides in air (e.g., in spent fuel reprocessing) also constitute a major safety and security-safeguards related issues. Purdue University, along with Sagamore Adams Laboratories LLC, is developing the tensioned metastable fluid detector (TMFD) technology for general-purpose alpha-neutron-fission spectroscopy. This paper focuses on rapid, high-efficiency detection of Rn and progeny in air using the novel TMFD technology; Rn and progeny isotopes in air are sparged through the TMFD detection fluid (to entrap the radioactive gas), which is then placed under a metastable state. Through tailoring the metastable fluid state, an audible and visible cavitation detection event is created and readily detected from transient bubble formation. Changing the tensioned state allows for the spectroscopic differentiability of Rn and its daughters which can be used to actively measure the equilibrium between the parent and daughter products. Such a technique can also be used to monitor the atmosphere in critical nuclear facilities for contamination from other alpha emitting isotopes (e.g., Pu, Cm, U...). TMFDs offer the unique ability for high intrinsic efficiency (>95%) alpha-neutron-fission fragment detection, while remaining blind to background beta and gamma radiation (qualified to >3.8×108 Bq m−3 using a dissolved 32P beta source, and also via gammas from a 53 R/hr 137Cs gamma source). Immunity to beta and gamma is beneficial for the discrimination of buildup of beta-emitting Thoron and Rn progeny in the detection fluid allowing for reusability. This paper will discuss the research results pertaining to detection of Radon and progeny in air, for concentrations between 74 Bq m−3 (2 pCi/L) and 740 Bq m−3 (20 pCi/L). The system measures a radon concentration between these levels to within ±15% intrinsic relative error (IRE) within 24 hours meeting the standards outlined by the American Association of Radon Scientists and Technicians-National Radon Proficiency Program (AARST-NRPP) Device Evaluation Program. Precision evaluation was also performed and the relative standard deviation defined by the AARST-NRPP was <5% exceeded the requirement of 25%. Ambient temperature effects were assessed at 10 °C and at 27 °C, which revealed a large increase in collection efficiency with decreasing sampling temperature and slight increase with increasing sampling temperature. Temperature effects on sensitivity thresholds and volumetric expansion were measured and used to compensate for variability in temperature over time. Blind testing with the help of Bowser-Morner Radon Reference Laboratory was performed and succeeded in accurately determining the Rn in air concentration to within 20% within only 6h of sampling. Finally, a 48-hour based collection time has also been developed for use in dwellings where Rn in air concentrations may vary in a day. Overall, the reproducibility and precision of TMFD technology is found to allow for an efficient, cost-effective, reliable, and environmentally friendly means of Rn and progeny detection, and by extension for use for general actinide in air monitoring for the nuclear industry.

Commentary by Dr. Valentin Fuster
2017;():V009T15A020. doi:10.1115/ICONE25-66807.

The variation law of motor vibration with load under saturation of magnetic circuit is one of the focus of motor vibration research. T-type equivalent circuit method is shown to be a suitable engineering analysis way. In this paper, the variation of magnetic flux density under saturation state was analyzed by using a small induction motor, and a physical model of magnetic saturation and motor vibration was established by the way of T-shaped equivalent circuit. Based on the physical model, the vibration mechanism of the motor under the saturated magnetic circuit was analyzed and the motor vibration simulation with different load conditions was carried out by the finite element method. Finally, a vibration test rig for motor under load was built and the simulation results were verified by experiments. The results show that when the motor load is less than the rated load, the amplitudes of the electromagnetic force in the middle and high frequency band change little. When the load is close to or higher than the rated value, the harmonic components of electromagnetic force in the middle and high frequency bands increase, and the amplitudes gradually increases due to the saturation of the magnetic circuit. Therefore, in the low-noise motor design, motor torque performance and vibration noise performance should be considered to optimize the saturation control of the magnetic circuit.

Commentary by Dr. Valentin Fuster
2017;():V009T15A021. doi:10.1115/ICONE25-66819.

Traditionally, the flow regime in two phase flow in rod bundles are considered in a global sense by the visualization. However, a sub-channel flow regime is required to understand and model the two phase flow structures in rod bundles. In this work, a sub-channel impedance meter was designed to get the dynamic feature in the sub-channels, which was applied to identify the sub-channel flow regime by the fast search and finding peaks of the cumulative probability distribution functions (CPDFs) objectively. In the present study, five flow regimes, namely bubbly flow, quasi-cap bubbly flow, quasi-slug flow, cap-turbulent flow and churn-turbulent flow were defined and recognized. The sub-channel flow regimes at the same cross section were compared to each other, which show similar feature with the local flow regimes in pipe. It is possible to identify nine different global flow regime configurations by combining the corner, side and inner sub-channel flow regime at the same cross section, which was drawn in the 2D sub-channel flow regime.

Commentary by Dr. Valentin Fuster
2017;():V009T15A022. doi:10.1115/ICONE25-66832.

The Supercritical CO2 Direct Cycle Gas Fast Reactor (SC-GFR) is a Generation IV reactor. The thermal properties of supercritical carbon dioxide are different from helium gas. Therefore, it is necessary to develop its dynamic model and study its control system. A mathematical model is developed for the SC-GFR. One prompt neutron group point kinetics equations with six groups of delayed neutrons is applied in the model. The core consists of an assembly of hexagonal fuel elements: the innovative Tube-in-Duct (TID) fuel assembly. Steady-state calculation is performed and the results are compared with the design data. The transient results are analyzed and the responses are reasonable in theory. The transient results show that the model can properly predict the SC-GFR dynamics. The study showed that it is not only feasible to build a numerical model of the SC-GFR, but also the control system can satisfy the design purposes.

Commentary by Dr. Valentin Fuster
2017;():V009T15A023. doi:10.1115/ICONE25-66843.

Heat transfer enhancement by the motion of the bubbles sliding along the heating surface are wildly reported by many researchers, thus it is of great importance to quantitatively investigate the characteristics of sliding bubbles. A visualization study of subcooled flow boiling of water in a vertical single face-heated narrow rectangular channel under a series of natural circulation working conditions was conducted. Pictures of the bubble sliding behaviors were captured by a high speed camera simultaneously with thermal data. A sequence of digital image processing algorithms were applied to the original picture to extract bubble shape and location information, which post-processing methods were adopted to obtain characteristic sliding parameters (including the distribution of the equivalent sliding bubble diameter and velocity, number density of the sliding bubbles).

It is found that bubbles can be able to nucleate and grow while sliding on the heating plate after the ONB point; the bubble number density, average bubble sliding velocity and the average sliding diameter continue to increase along the test section; heat transfer in the flow channel are significantly enhanced along flow direction with relatively low local void fraction. The average bubble sliding velocity near the inlet is significantly smaller than the sectional average velocity of single-phase fluid of the flow channel, and then it exceeded near the outlet of the test section. The average bubble sliding velocity and diameter increase with increasing heat flux and decreeing local subcooling degree. The equivalent diameter of sliding bubbles and the bubble sliding velocity approximately follow normal distribution. The distribution of the bubble diameter and velocity both cover a wild range. The standard deviations of the probability density function of the sliding bubble diameter and velocity increase with increasing heat flux and decreasing subcooling degree.

Commentary by Dr. Valentin Fuster
2017;():V009T15A024. doi:10.1115/ICONE25-66881.

Plate-type reactor fuel is getting increasing attentions as it features excellent heat transfer ability and compact structure. Turbulence mixing accompanied with momentum and mass exchange is typical phenomena in the converging region downstream of parallel plane fuels. To deepen understanding turbulence mixing characteristics and obtain more benchmark data for CFD, an experimental study was conducted by 2D particle image velocimetry (2D-PIV). Ensemble average profile of velocity, vorticity, turbulence intensity, as well as Reynolds stress are analyzed, respectively.

Results reveal that two kinds of merging points (mp) were found, i.e., flows from two side jets merge at y/d = 3 (mp1), and flows from middle three jets merge at y/d = 7.6(mp2). The decay of vertical velocity decreases fast primarily until y/d = 6, then slightly increases until y/d = 8.8, and finally decreases gradually. Decay of velocity magnitude of present study decreases sharply until y/d = 6.4, then decreases gradually. This tendency of five-parallel jets is similar to the result of two-parallel jets. From flow visualization, the vortex scale decreases once the vortex formed. The vorticity reaches its maximum at about y/d = 1.3 and then decreases gradually. Consistently with the indicated of velocity distribution, the vorticity distribution tends to mild with the flow developing downstream. However, the location of vortex center is affected by spanwise momentum exchange. Vortices started moving closer around the mp2 indicating a strong combination of activities. Spanwise turbulence intensity shows strong fluctuations existed around mp1 implying that the main momentum transfer happened in the merging region.

Topics: Turbulence , Jets
Commentary by Dr. Valentin Fuster
2017;():V009T15A025. doi:10.1115/ICONE25-66914.

The objective of this paper is to act as a collection of multiple different heat-transfer correlations and to check their accuracy when compared to experimental data obtained in supercritical-pressure refrigerants (R-22 and R-134a). This paper is also intended to collect as much relevant data of heat transfer in supercritical refrigerants as possible for future research. The experimental data have been retrieved from graphs within a wide range of operating parameters. This study is in support of potential use of supercritical refrigerants as modeling fluids instead of supercritical water. The use of refrigerants as modelling fluids instead of water will allow to decrease costs and technical difficulties during experiments at supercritical pressures and widen operating ranges, because the critical parameters of refrigerants are significantly lower than those of water.

The research was completed by collecting graphed data from several different experimental series using both R-22 and R-134a data. The advantage of comparing different refrigerants for determining correlation accuracy is to increase the predictability for other potential experiments using refrigerants. All data are taken from bare-tube experiments to produce a relative baseline for heat-transfer characteristics. These experiments have been performed within the following range: Inner tube diameter ranging between 4.4 mm to 13 mm, pressure ranging between 4.3 MPa to 5.5 MPa, and at a number of various mass and heat fluxes. Sixteen potential heat-transfer correlations have been selected and used in this assessment. The correlation by Watts and Chou [1] and Cheng et al. [2] were shown to have the lowest root-mean-square error. Other correlations with the reasonable accuracy include Mokry et al. [3] and Swenson et al. [4] correlations. However, it was decided to develop a new correlation based on these refrigerant data in an attempt to increase the prediction accuracy.

Therefore, based on the Mokry et al. [3] correlation a modified correlation was developed, which generalized the experimental Freon data with higher accuracy than the know correlations. This correlation is intended to create a basis for further study on the use of refrigerants as modeling fluids. While Freon has similar properties to water at supercritical conditions, the different molecular properties causes factors to affect each fluid differently. For refrigerants at supercritical conditions, the factors that seem to have the most effect are the dynamic viscosity and density of a fluid.

Commentary by Dr. Valentin Fuster
2017;():V009T15A026. doi:10.1115/ICONE25-66925.

Currently, there is a number of Generation-IV SuperCritical Water-cooled nuclear-Reactor (SCWR) concepts under development worldwide. These high temperature and pressure reactors will have significantly higher operating parameters compared to those of current water-cooled nuclear-power reactors (i.e., “steam” pressures of about 25 MPa and “steam” outlet temperatures up to 625 °C). Additionally, SCWRs will have a simplified flow circuit in which steam generators, steam dryers, steam separators, etc. will be eliminated, as the steam will be flowing directly to a steam turbine.

In support of developing SCWRs studies are being conducted on heat transfer at SuperCritical Pressures (SCPs). Currently, there are very few experimental datasets for heat transfer at SCPs in power-reactor fuel bundles to a coolant (water) available in open literature.

Therefore, for preliminary calculations, heat-transfer correlations developed with bare-tube data can be used as a conservative approach. Selected empirical heat-transfer correlations, based on experimentally obtained datasets, have been put forward to calculate Heat Transfer Coefficients (HTCs) in forced convective in various fluids, including water at SCPs.

The Mokry et al. correlation (2011) has shown a good fit for experimental data at supercritical conditions within a wide range of operating conditions in Normal and Improved Heat-Transfer (NHT and IHT) regimes. However, it is known that a Deteriorated Heat-Transfer (DHT) regime appears in bare tubes earlier than that in bundle flow geometries. Therefore, it is important to know if bare-tube heat-transfer correlations for SCW can predict HTCs at heat fluxes beyond those defined as starting of DHT regime in bare tubes.

The Mokry et al. (2011) correlation fits the best SCW experimental data for HTCs and inner wall temperature for bare tubes at SCPs within the NHT and IHT regimes. However, this correlation might have problems with convergence of iterations at heat fluxes above 1000 kW/m2.

Commentary by Dr. Valentin Fuster
2017;():V009T15A027. doi:10.1115/ICONE25-66934.

A wireless tomography system has been developed to measure the real-time behavior of solid-liquid two phase flow in the centrifuge for controlling a rotational velocity and particle supply rate.

The purpose of this study is development of real-time behavior of solid-liquid two phase flow measurement technique and image conductivity distribution in rotating body because a technique for measuring the behavior of solid-liquid two phase flow in the centrifuges has not existed yet. In this study, a device to measure wirelessly behavior of solid-liquid two phase flow in stationary body has developed as a preliminary to measure in rotating body.

A centrifugation technology for industry process should be improved to obtain more effective separation, shorten process time and save energy. These requirements are achieved by optimizing rotational velocity and particle supply rate. The real-time measurement of behavior of solid-liquid two phase flow for controlling the rotational velocity and the particle supply rate is essential. In other words, the real-time behavior of solid-liquid two phase flow measurement and the rotational velocity control become innovative technologies. Typical behavior of solid-liquid two phase measurement technics with easy handling in particles liquid two-phase flow uses process tomography because the process tomography is suitable for real-time measurement in a rotating body. Process tomography has high temporal resolution. This detector was used with the wireless because electrical cables are not available for centrifuges under high speed rotational condition. This wireless tomography system was used for a lab scale rotating machinery measurement experiment.

Consequently, we could wirelessly measure the behavior of solid-liquid two phase flow in stationary body in real time and get data to image the behavior of solid-liquid two phase flow. This result images the behavior of solid-liquid two phase flow. Observing it enable to control a rotational velocity and particle supply rate.

So this wireless tomography system can separate solid-liquid two phase flow efficiently.

Commentary by Dr. Valentin Fuster
2017;():V009T15A028. doi:10.1115/ICONE25-66937.

Steam Injector (SI) is a passive pump activated by steam and water, and it does not require any external power supplies or rotating machineries. Moreover, SI has high ability as a heat exchanger by undergoing direct contact condensation mechanism. From these characteristics, SI has a capability to be applied as a passive safety system in the nuclear power plant. In the present study, experiments targeting the operating range and pump performance were carried out to obtain SI’s detailed characteristics under following supply conditions; inlet steam pressure was 0.02 ∼ 0.81MPaG, inlet water flow rate was 0.21 ∼ 0.80 kg/s. In the former experiment, operating range was investigated by changing inlet conditions, and the influence of steam inlet pressure and water inlet flow rates on SI operating range was tested. In the latter experiment, pump performance of SI was evaluated by investigating the maximum discharge pressure in each inlet condition by using back pressure valve. From the results of experiments, it was confirmed the operating range of SI was limited by supplied steam pressure and supplied water flow rate, and some clear trends were found in the operation boundary map. In addition, SI could discharge water at least 1.2 times higher than inlet steam pressure under above-mentioned conditions, which verified SI’s capability to be applied for the nuclear power plant as a core cooling system. And furthermore, existing 1D analysis model’s predictive capability was tested based on these experimental results.

Topics: Ejectors
Commentary by Dr. Valentin Fuster
2017;():V009T15A029. doi:10.1115/ICONE25-66953.

SuperCritical Water-cooled Reactor (SCWR) as one of the six Generation-IV nuclear-power-reactor concepts will have increased thermal efficiency compared to that of current Nuclear Power Plants (NPPs) equipped with water-cooled reactors by operating the reactor coolant at supercritical conditions: Coolant pressure of about 25 MPa, inlet temperatures between 300–350°C, and outlet temperatures between 550–625°C. The major flow geometry inside the reactor core is the bundle flow geometry. For safe and efficient operation of an SCWR heat transfer coefficients should be calculated with minimum uncertainties.

Unfortunately, the vast majority of experimental datasets were obtained in vertical bare tubes cooled with SCW. Experiments in a bundle flow geometry are even more complicated and expensive compared to that in bare tubes. Due to this very few experiments have been performed in bundles.

According to the abovementioned, the vast majority of heat-transfer correlations are based on bare-tube data, and only one currently known correlation is based on a 7-element bundle cooled with SCW (the so-called, Dyadyakin and Popov correlation (1977)). Rods in this bundle are equipped with four helical ribs to enhance the heat transfer. However, the authors have not provided any dataset(s) associated with this bundle and correlation.

In the current paper a number of bare-tube heat-transfer correlations obtained in SCW and the Dyadyakin and Popov correlation were compared with two datasets obtained in an annular channel with the heated central rod and 3-element bundle. The central rod in this annular channel and rods in the 3-element bundle have the same heated length as those in the 7-element bundle tested by Dyadyakin and Popov in 1977, and are also equipped with four helical ribs.

The comparison showed that the Jackson correlation (2002) is the most accurate one in predicting Heat-Transfer-Coefficient (HTC) profiles in the annular channel at normal heat-transfer regime. The Dittus and Boelter correlation (1930) is the most accurate in predicting HTC profiles in the 3-element bundle at normal heat-transfer regime. No one correlation is capable to follow closely HTC profiles at the deteriorated heat-transfer regimes in both flow geometries. Aloo, it should be mentioned that bare-tube heat-transfer correlations, which have thermophysical properties based on bulk-fluid and wall temperatures, might have problems with convergence at high heat fluxes, i.e., above the heat flux at which the deteriorated heat-transfer regime starts in bare tubes.

Topics: Heat transfer , Water
Commentary by Dr. Valentin Fuster
2017;():V009T15A030. doi:10.1115/ICONE25-66981.

As an important kind of energy source, radioisotope batteries are attracting more and more academic researchers and people from industry due to the high power density, long lifetime (equal to half life of the radioisotope source), outstanding reliability, without maintenance, miniaturization and wide application compared with traditional dry batteries, chemical batteries, fuel batteries and solar batteries. Radioisotope batteries have been developed for more than 15 species since the first β battery invented by Henry Mosley in 1913. Based on a Brayton cycle Radioisotope Power System and a PZT-5H (Pb(ZrxTi1-x)O3, 0≤x≤1) unimorph, the piezoelectric nuclear battery driven by the jet-flow (PNBJ) is demonstrated in this work. The PZT-5H unimorph replaces turbine and utilizes high speed nitrogen jet-flow heated by the decay energy of radioisotope to output electrical energy. Over 0.34% energy conversion efficiency for the PNBJ is obtained at the flow of 2.26 × 10−3 m3/s and room temperature on half plane. The PNBJ can be used in low power microelectronics and microsystems, like electronic watch, AC-LED (alternating current light-emitting diode), and sensors. We believe that the researches and applications of PNBJ will be much attractive with the breakthroughs of preparation technology made by academic and industrial world.

Commentary by Dr. Valentin Fuster
2017;():V009T15A031. doi:10.1115/ICONE25-67009.

From the viewpoint of an importance of safety, the nuclear power plant should be managed to prepare severe accidents. The performance of safety dropped by an accident is strongly to be minimized during the situation of station blackout. The installation of a steam injector (SI) into the nuclear power plant has long been expected.

In the SI, the steam condenses due to the direct contact at the surface of water jet, resulting in the force attracting water. The force drives the circulation of an amount of coolant water. SI also works as a reactor condenser thanks to its high efficient performance during the condensation. Because any external forces to circulate water and steam are not required, SI can be operated without the electric powers. The structure of SI is similar to a convergent-divergent nozzle. After the flow acceleration at a throat, the discharged pressure is expected to exceed the inlet pressure. Owing to its quite simple structure, the reduced cost of installation and maintenance is also expected.

The following previous studies for four cases of throat diameter clarified two-phase flow structures and heat transfer characteristics in water jet and performance of SI: (i) Narabayashi et al. (2000) examined for 5.5 and 6.5 mm in diameter; (ii) Osakabe et al. (2004) for 3.4 mm; (iii) Koizumi et al. (2006) for 4 mm; (iv) Abe et al. (2014) for 4, 6.5, and 8 mm. Although these clarified the operative state which formed a water jet, operative condition was not elucidated. Furthermore, the scale effect for various diameters of SI has not been discussed in detail.

The aim of this study is to clarify scale effect of a test section on operating criteria and performance. Experiment was performed to clarify the scale effect by using three types of throat diameters: 4, 6.5, and 8 mm.

As a result, three formations of a water jet were observed: (i) formation, (ii) incomplete formation, and (iii) no formation. We proposed a classification which enables us to categorize complex flow patterns into five regimes. We clarified the operating criteria of them by comparing water flow rate with steam flow rate. SI did not form a water jet on the condition with low steam flow rate. The suppling water was stopped, and only steam was supplied to the test section for the condition that steam latent heat was larger than subcooled water enthalpy.

Commentary by Dr. Valentin Fuster
2017;():V009T15A032. doi:10.1115/ICONE25-67021.

The phenomenon of secondary droplet production during single drop impingement onto a liquid film is encountered in many industrial situations. Typical examples in the field of nuclear engineering are the spray cooling of hot surface and the atomization of radioactive liquids in severe accident. Therefore, the prediction of the onset of secondary droplet production is very important. It is known that the two types of droplet splashing mechanisms are present: the prompt splash and the late splash. The main purpose of this research is to determine the splashing limit separately for the prompt splash and the late splash. It is expected that the splashing limits are expressed using the three dimensionless numbers: the Weber number, the Ohnesorge number, and the dimensionless film thickness. Experiments were hence carried out using pure water and silicone oil as the working liquid. The experimental ranges were 129–606 for the Weber number, 0.00183–0.00300 for the Ohnesorge number, and 0.13–3.0 for the dimensionless film thickness. It was found that the occurrence of splashing can be predicted more accurately if the splashing limit is evaluated separately for the prompt splash and the late splash.

Commentary by Dr. Valentin Fuster
2017;():V009T15A033. doi:10.1115/ICONE25-67065.

Instrumentation and Control (I&C) systems for Nuclear Power Plants (NPP) are exceedingly complicated electronic solutions that include thousands of different components such as microcontrollers, Field-Programmable Gate Arrays (FPGAs), integrated circuits etc. Deployment of such safety-critical systems cannot be performed without complex safety and reliability assessment, verification and validation (V&V) activities that are addressed to exposing of overlooked faults. The examples of such activities are Fault Tree Analysis (FTA), Failure Modes and Effects Analysis (FMEA), Fault Injection Testing (FIT). Due to complexity of NPP I&C systems in most cases the process of assessment is very time consuming and the results mostly depend on experts’ qualification.

Traditional safety and reliability assessment methods are being constantly modified and enhanced so as to comply with increasing demands of national and international standards and guidance, as well as to be applied for I&C systems that contain number of complex components like FPGA.

Although much work related to analysis of FPGA-based systems has been performed, there is a lack of detailed technique for FPGA-based I&C systems failure identification that considers probability of several faults at the same time (multi-faults), development of preventive strategies for controlling or reducing of the risk related to such failures, as well as automation of this technique so as to make it utilizable for real NPP industry tasks.

FIT as verification for Failure Modes, Effects and Diagnostics Analysis (FMEDA) was used during Safety Integrity Level 3 (SIL3) certification process of RadICS NPP I&C platform, while the parts of proposed technique were used as internal verification and validation activities applied on several modules of the platform.

Commentary by Dr. Valentin Fuster
2017;():V009T15A034. doi:10.1115/ICONE25-67120.

Nowadays cyber security assurance is one of the key challenges of safety critical software based NPP I&C (Nuclear Power Plants Instrumentation and Control) systems requirements profiling, development and operation. Any I&C system consists of a set of standard software (SW), hardware (HW) and FPGA components. These components can be selected and combined in different ways to address the particular control and safety assurance related tasks. Some of them are proprietary software (PS) and commercial off-the-shelf (COTS) components developed previously. Application of such components reduces the level of safety and cyber security, because they can contain vulnerabilities that were created intentionally. In this case, targeted attacks can lead to a system failure.

National Vulnerability Database (NVD) and other open databases contain information about vulnerabilities which can be attacked by insiders or other intruders and decrease cyber security of NPP I&C systems.

In this paper, we propose a safety assessment technique of NPP I&C systems, which consists of the following procedures:

1. Analysis of I&C architecture to assess influence of OTS component failures on dependability (reliability and safety) of the system. For that purpose, FMEDA or similar techniques can be applied.

As a result, three-dimension criticality matrixes (CM) (with metrics of detection, probability and severity) are developed for different components (SWFCM and HW/FPGAFCM).

2. The IMECA-based assessment of OTS components and their configuration. In this case, CMs (SWICM and HW/FPGAICM) describe the degree of failure component influence on cyber security.

3. Joining of criticality matrixes (SWFCM and HW/FPGAFCM, SWICM and HW/FPGAICM), impact analysis of components depending on degree of influence on cyber security and safety as a whole.

4. Developing of Security Assurance Case and selecting of countermeasures according to safety (cyber security)/costs criteria.

The developed tool supports creation of criticality matrixes for each analyzed component of the system and I&C as a whole. Joining of criticality matrixes allows creating common matrix for system cyber security and functional safety. The tool supports decision making to optimize choice of countermeasures according to criterion of safety and security/cost criterion.

Commentary by Dr. Valentin Fuster
2017;():V009T15A035. doi:10.1115/ICONE25-67126.

In severe accidents of nuclear power plants such as the Fukushima Daiichi Unit 1 accident, aerosol containing massive amount of Fission Products (FP) could be created. Pool scrubbing is an effect of capturing these FP inside the water during the flow of aerosol. Such a situation may occur at the vent of containment vessel in the BWR as well as the heat-transfer-tube disrupted accident of steam generator in the PWR. The amount of FP released into the atmosphere could be reduced by the effect of pool scrubbing. Hence, the removal of aerosol by pool scrubbing is expected to evaluate the source terms stated below.

Although the importance of pool scrubbing has been recognized, the study for its mechanisms from the viewpoint of particle-bubble and two phase flow had not been carried out intensively. Furthermore, lack of experimental data for various conditions should be improved. Such a state leads to prevent us from analyzing the efficiency of pool scrubbing when advancing the model and testing the validation of it.

The aim of this study is to clarify the transfer of aerosol and bubble experimentally. The data from the experiment evaluates the influence of each parameter to remove the FP. The analysis code presently used adopts the model for the velocity which affect the particle to go out from the bubble. Each of them are written as (i) centrifugal deposition velocity, (ii) gravitational velocity, (iii) incoming vapor velocity, (iv) Brownian diffusion velocity. We choose our parameter which correspond to each of the velocities. Remarking that the experiment also include conditions e.g. the temperature stratification, the types of particle and gas, which is not been described in the manual. Based on the experimental result, our goal is to test validation and improve the model appropriate to the severe accidents predictive code such as the MELCOR by reflecting the study of flow mechanism and the effect of various conditions. To the end, this paper focuses on the rotation and its movement of aerosol particles inside the bubble.

This study focuses on the behavior of a single air bubble including aerosol by visualizing the inner flow of bubble containing particle. As a result for visualization of a single rising bubble, we were able to take a film of an area inside the particle where information of particle movement is seen. We also succeed in observing a clockwise rotation flow inside bubble just above the nozzle. As to avoid the refraction between air-water, we also made an experiment against rising oil droplets. From the series of experiment, the growth of upward flow from desorption to the stationary rising mode was been seen inside the droplet. As the flow grew, a top-to-bottom rotation was seen as it was mentioned in previous models.

Commentary by Dr. Valentin Fuster
2017;():V009T15A036. doi:10.1115/ICONE25-67167.

In this paper, we first propose a novel composite nuclear fuel of UO2-GaN, which has never been reported before, and then its fully coupled multiphysics fuel performance is investigated using the CAMPUS code developed by ourselves. We propose two different fabrication methods to obtain the UO2-GaN fuel, which are Green Granule/Slug Bisque and Spark Plasma Sintering, respectively, resulting in different fuel thermal conductivities. By comparing two kinds of UO2-GaN fuel which are fabricated by two methods, we found that fuel fabricated by Green Granule/Slug Bisque possesses high thermal conductivity and performs well during the reactor operation. The gap width, gap conductance, fission gas release, plenum pressure, deviation of oxygen to metal ratio and displacement are all studied in this work. The performance of this novel fuel is also compared with the traditional UO2 fuel. The UO2-GaN enhanced thermal conductivity composite fuel shows the potential of decreasing the fuel temperature, and improving fuel performance and reactor safety. This makes GaN a good candidate to fabricate composite fuel with UO2 from the thermal standpoint. However, this work is to conduct an exploratory approach to the effect for the GaN addition to UO2 fuel with very limited data. So, further studies are still needed on GaN’s compatibility with UO2, neutronic behavior, fission product retention capabilities and irradiation performance, both on experimental measurements and numerical simulations.

Commentary by Dr. Valentin Fuster
2017;():V009T15A037. doi:10.1115/ICONE25-67168.

The pressurizer is the fundamental equipment in nuclear power plant, maintaining the pressure in the primary side. A U-shaped tube filled with water, as the water seal structure, was installed in front of the safety valve with the purpose of reducing the non-condensable gas leaked through the safety valve. When the safety valve opens, water slug in the U-shaped tube moves through the safety valve, causing a large number of thermal-hydraulic loads on the safety valve and downstream line. In order to reduce the thermal-hydraulic loads on the safety valve system, the U-shaped tube water seal structure was placed inside the pressurizer dome and replaced by a scoop-shaped structure. Thermal-hydraulics characteristics of the water seal structure are simulated based on the geometric model with a scale ratio of 1:1 to investigate the condensation and formation of water seal. The key parameters in the water seal formation process are investigated under different pressure and heat dissipating capacity. The species transport model is utilized to describe the impact of non-condensable gas under various mass fractions. Three-dimensional distributions of pressure and temperature are obtained from the calculation by using the CFD code ANSYS FLUENT. The water seal formation time is calculated by using the condensation rate and the geometric model. The result reveals that water seal formation can be completed within the required time, even under a high mass fraction of non-condensable gas. Water seal formation time reduces when the system pressure increases. The temperature difference across the water seal is lower than 20K.

Commentary by Dr. Valentin Fuster
2017;():V009T15A038. doi:10.1115/ICONE25-67239.

The water condensation is important for wide range of industrial systems such as condensers and heat exchangers of steam power plants and refrigerators. The condensation generally has two patterns; filmwise condensation (FWC) and drop-wise condensation (DWC). DWC has one-tenth higher heat transfer coefficient than that of FWC. It has been pointed out by many investigators that DWC occurs on the hydrophobic surface and FWC occurs on hydrophilic surface. However, the durability of those hydrophobic effects was not clear enough. In order to maintain a sufficiently long DWC, it is important to understand the effect of the surface property and structure on the condensation surface in more detail. The recent advancement of MEMS (Micro Electro Mechanical System) technology enables us to change the physical nature the surface in the micro scale. It is expected that the hydrophobic surface by the MEMS technology may kept DWC for a longer time. In the present paper, we experimentally investigate the effects of thin metal film and micro structured surface on condensation pattern. Especially, our condensation experiments were performed with the micro structured surfaces by using etching and the metal thin film surfaces by sputtering for approximately 24 hours. Silicon (Si) wafer was used as a basic surface. For the metal thin films surface, we used sputtered Lead (Pb) and Titanium (Ti) on Si surface. For the micro structured surfaces, micro-structured grid was etched on Si surface with several conditions. In order to obtain the relation between the condensation pattern and surface condition, the surface conditions were measured by laser micro-scope, contact angle meter and atomic force microscope (AFM).

For the metal thin films surface, condensation patterns on thin Pb film surface showed DWC. Meanwhile, condensation patterns on thin Ti surfaces showed FWC. From our results, the adsorption forces decreased with increasing contact angle on DWC for Pb. On the other hand, the adsorption forces increased with decreasing contact angle on FWC for Ti. For the micro structured surfaces, condensation pattern was FWC and contact angle decreased in our experimental results. This is because that the condensed water is accumulated in the groove on the micro structure surface.

Commentary by Dr. Valentin Fuster
2017;():V009T15A039. doi:10.1115/ICONE25-67245.

The continuing search for a long-term storage for highly-active nuclear waste in Germany requires a prolonged intermediate storage period of spent fuel in dry storage casks at the power plant sites. Currently, it is not sufficiently clear if there might be a loss of integrity of the fuel rods within such long periods, e.g. due to rising pressure from gaseous products of nuclear decay. Regarding a final evaluation, extrapolative modelling of the radiochemical and thermomechanical material behavior is challenging and not suitable for predictions on the condition of storage container inventory after the intermediate storage period. Therefore, it is of public interest to find measurement principles or methods which can provide information about the condition of the storage container inventory.

In line with a cooperative project (project partners: Technical University Dresden, Zittau/Görlitz University of Applied Sciences) different measurement principles and methods (radiation emission, muon transmission, thermography, acoustical spectrometry) for non-invasive condition monitoring of the storage container inventory in case of prolonged intermediate storage are going to be investigated and evaluated. The results shall help to determine suitable methods for the identification of both changes of the spent fuel and inner container structure over long periods without opening the container and would be a significant contribution for the long-term safety of intermediately stored highly radioactive waste. Furthermore, suitable methods would provide information about the transport and conditioning ability of the waste before transfer to the repository.

This paper deals with the content of the subproject of Zittau/Görlitz University of Applied Sciences as well as with the approach for project realization. A further main part of this paper is the development of experimental infrastructure to support the investigations.

Commentary by Dr. Valentin Fuster
2017;():V009T15A040. doi:10.1115/ICONE25-67252.

The amount of spent fuel and high-level waste already available, and which will be produced by the future NPPs operation, calls for the evaluation of any possible technological solution that could minimize the burden of their disposal: reduction of Minor Actinide (MA) content, in addition to the radiotoxicity and radioactivity, and of the generated thermal load (decay heat). In this context, R&D efforts currently focus on the development of methodologies and technical solutions for Partitioning and Transmutation.

MAs and long-lived fission products are in fact the main contributors to the long-term radiotoxicity of spent nuclear fuel, and their transmutation to short-lived fission products, in fast spectrum nuclear reactors, in transmuters or in Accelerator Driven Systems (ADS), by neutron irradiation of dedicated fuels/targets, is a promising and widely investigated option.

In order to provide substantial input for the safety assessment of innovative nuclear fuels dedicated to MA transmutation, several irradiation tests are being carried out. In some options, the investigated fuels/targets are uranium free, or of low uranium content, to improve the transmutation performance and contain high concentrations of MA and plutonium compounds.

Two molybdenum based CER-MET fuels, called ITU-5 & ITU-6, were prepared at JRC Karlsruhe for the irradiation experiment FUTURIX-FTA (FUel for Transmutation of transURanium elements in phenIX/Fortes Teneurs en Actinide). The experiment performed from 2007 to 2009 in the Phénix reactor, France, in cooperation with CEA. The experiment ended after 235 equivalent full power days (EFPD) at a Linear Heat Rate of circa 130 W/cm and reached burn-ups of 18 %FIHMA and 13 %FIHMA, respectively. Afterwards, the pins were transported to the Hot Cells of JRC Karlsruhe for Post Irradiation Examination.

After a short summary describing the fuel preparation and irradiation conditions of the FUTURIX FTA irradiation experiment, the paper will give an overview on the current status and further planning of the Post Irradiation Examinations of ITU-5 & ITU-6 at JRC Karlsruhe. Finally, the results of the characterisations will be discussed and conclusions on the irradiation performance will be drawn. The results of this experiment will help to increase the knowledge and understanding of the irradiation behaviour of metal based transmutation targets and the qualification and validation of models developed to predict fuel safety performance.

Topics: Nuclear fuels
Commentary by Dr. Valentin Fuster
2017;():V009T15A041. doi:10.1115/ICONE25-67282.

In order to understand airflow dynamics through small openings encountered in containment enclosures used for nuclear decommissioning operations, the results of experimental and numerical investigations are analyzed.

The main purposes of this work are to identify the required conditions likely to generate flow inversions at the studied opening which lead to pollutant leakage outside depressurized enclosures, and also to verify the ability of CFD1 simulations to predict these flow inversions by using U-RANS2 and LES3 approaches. All along this work, we tried to reproduce the conditions of leakage occurring at the opening in terms of aerodynamics and openings geometries.

Laser flow visualizations and CFD results show that an additional flow, such as a turbulent jet in competition with the directional flow and a disturbed level of pressure inside the enclosure are among the main causes leading to the leakage through the opening.

Commentary by Dr. Valentin Fuster
2017;():V009T15A042. doi:10.1115/ICONE25-67284.

As a critical component in the advanced non-light water nuclear power plants with indirect power cycles, an intermediate heat exchanger (IHX) is used to transfer the heat from the primary coolant system to the secondary system. Generally, such reactors are operated at very high temperatures and pressures. Therefore, IHXs are required to be able to provide high thermal effectiveness while withstanding demanding operating pressure and temperature combinations. A suitable IHX candidate needs to possess high mechanical integrity.

One of the promising IHX candidate is Printed Circuit Heat Exchangers (PCHEs). The compact diffusion-bonded PCHEs are characterized by mini- or micro-flow channels photo-chemically etched on flat metal plates. These etched plates are stacked up and diffusion-bonded to form a monolithic block. Therefore, the mechanical integrity of PCHEs are centered on the mechanical design of these etched flow channels.

To date, there are several flow channel geometries developed for PCHEs. Straight and zigzag channels, the first two designs, can be often seen in Heatric-manufactured PCHEs in compliance with ASME codes to ensure the mechanical integrity. To improve the fluid flow and heat transfer capability and reduce the associated pressure drop, straight and zigzag channels evolve to other two variants: S-shaped fin and airfoil fin channels. However, since the surface geometry of the flow channels become significantly complicated, the ASME code is not applicable any longer. Numerical simulation is then required to analyze the mechanical strength of the evolved flow channels under extremely harsh operating conditions.

This paper provides a review on the mechanical design of PCHEs with straight and zigzag channels, which refers to the ASME Boiler and Pressure Vessel Codes. In addition, numerical simulations of the mechanical stress for a case study of S-shaped fin will be presented with Inconel 617 as the construction material. Research plans at the Ohio State University are also presented. The aim of this paper is to summarize the common mechanical design method for PCHEs as well as introduce some of the ongoing research work related to the mechanical strength analysis of complicated flow channels with the aid of computer simulation.

Commentary by Dr. Valentin Fuster
2017;():V009T15A043. doi:10.1115/ICONE25-67287.

In the present experimental study an attempt has been made to study the boiling heat transfer characteristics of variety of enhanced surfaces. Three different copper test surfaces: polished copper and two structured surfaces were used in the present investigation. The heat transfer performance of each surface is studied under saturated pool boiling conditions at atmospheric pressure by using water and isopropyl as pool liquid. The effect of intersecting tunnel geometry with 0.5 mm and 1 mm depth on heat transfer performance has been studied. The comparison of heat transfer coefficient indicates that the intersecting tunnel structure enhanced the boiling heat transfer performance and reduced the wall superheat at given heat flux inputs.

Commentary by Dr. Valentin Fuster
2017;():V009T15A044. doi:10.1115/ICONE25-67293.

The present study investigates the influence of liquid inlet modelling on the development of liquid waves in isothermal churn flow of air and water in a vertical pipe. The porous wall liquid inlet section, commonly used in experiments, is modelled as a simple inlet flow area in our simulation. Using the liquid mass flow rate from experiment, the magnitude of the wall normal velocity component is determined by the inlet area which is used as a modelling parameter. This parameter significantly affects the calculated liquid wave frequency. The inlet liquid velocity profile was not measured in available experiments and thus presents a major source of uncertainty in simulations. The parametric analysis shows that a suitable liquid inlet area can be determined over the range of liquid flow rates, leading to good agreement of simulated and measured wave frequencies. A three-dimensional simulation was performed using the multiphase solver interFoam from the open-source code OpenFOAM.

Commentary by Dr. Valentin Fuster
2017;():V009T15A045. doi:10.1115/ICONE25-67327.

Passive safety systems represent one field of research concerning the safety-related enhancement of nuclear power plants. Passive safety systems can ensure the safe removal of decay heat without an input of electrical or mechanical energy for commissioning or operation. The heat removal chain is guaranteed on the basis of the physical principles condensation, heat conduction, boiling and natural circulation. The thermal hydraulic processes in passive safety systems disagree with the plant-specific thermal hydraulics because of different operating conditions. Since the established system codes are validated for the plant-specific conditions, the operational behavior of passive safety systems is currently not sufficiently predictable.

On this account, the German Federal Ministry of Education and Research initiated the joint project PANAS to investigate the decay heat removal by passive safety systems on the basis of experimental analyses, modelling and validation. Object is the heat removal chain in advanced boiling water reactors consisting of emergency condensers (EC; heat transfer from reactor core to core flooding pools) and containment cooling condensers (CCC; heat transfer from the containment to the shielding/storage pool).

At Technische Universität Dresden, the test facility GENEVA was constructed for the experimental investigation of the operational behavior of the CCC. GENEVA models the CCC concerning the original thermal hydraulic conditions of the heat source and heat sink as well as the tube geometry for the heat transfer. In this way, the comparability of the thermal hydraulic phenomena is given. Previous experiments focused on the stability analysis of the natural circulation in the test facility.

The focus of PANAS is on the condensation process of saturated steam at the outside of the slightly inclined tubes and the convection respectively boiling of both a stable and an unstable two-phase flow inside these tubes. For a detailed analysis, condensation rates at the outside as well as the flow structure inside have to be investigated experimentally.

Therefore, the instrumentation in the heat transfer section of GENEVA is considerably enhanced. This enhancement comprises an optical measuring system for the film thickness or droplet size of the condensate, a tipping scale for the condensate mass flow, void probes for the steam void fraction and more than 100 thermocouples outside and inside the tubes for temperature profiles in axial, radial and azimuthal direction. By reference to these parameters, it is possible to examine the thermal hydraulic models for the heat transfer.

The paper outlines the available models in system codes regarding condensation and boiling concerning the operating conditions of the CCC. Since dropwise condensation could be observed in previous experiments and the condensation models in system codes focus on film condensation, the review is extended beyond native models. A sensitivity analysis of the reviewed models regarding condensation shows huge differences concerning the value of the heat transfer coefficient. Furthermore, the courses of the condensation models present different dependencies regarding the heat transfer coefficient and the wall temperature. Due to this, the necessity of the experimental investigation and later the revision of the condensation models in system codes is confirmed. The comparison of the reviewed models with first experimental results outlines the tendency for the numerical description of the condensation process.

Based on the investigation and validation of models concerning the heat transfer processes in the CCC, the operational behavior will be accurately predictable by established system codes, which enhances the safety investigation and the licensing. Although the conception of this investigation is founded on the CCC, the adapted models will be able to characterize the heat transfer processes boiling and condensation for saturation conditions at a relatively low pressure (maximum 4 bar) and for natural convection in general.

Commentary by Dr. Valentin Fuster
2017;():V009T15A046. doi:10.1115/ICONE25-67371.

Choking flow plays an integral part not only in the engineered safeguards of a nuclear power plant (NPP), but also to everyday operation. Current NPP steam generators operate on the leak-before-break approach. The ability to predict and estimate a leak rate through a steam generator tube crack is an important safety parameter. Knowledge of the maximum flow rate through a crack in the steam generator tube allows the coolant inventory to be designed accordingly while limiting losses during loss of coolant accidents. Here an assessment of the choking flow models in thermal-hydraulics code RELAP5/MOD3.3 is performed and its suitability to predict choking flow rates through small axial cracks of the steam generator tubes is evaluated based on previously collected experimental data. Three sets of the data were studied in this work which corresponds to steam generator tube crack sample 1, 2, and 3. Each sample has a wall thickness, channel length (L), of 1.285 mm to 1.3 mm. Exit areas of these samples are 5.22 mm2, 9.05 mm2, and 1.72 mm2 respectively. Samples 1 and 2 have the same flow channel length to hydraulics diameter ratio (L/D) of 2.9 whereas sample 3 has a L/D of 6.5. A pressure differential of 6.8 MPa was applied across the samples with a range of subcooling from 5 °C to 60 °C. Flow rates through these samples were modeled using the thermal-hydraulic system code RELAP5/MOD3.3. Simulation’s results are compared to experimental values and modeling techniques are discussed. It is found that both the Henry-Fauske (H-F) and Ransom-Trapp (R-T) models better predict choking mass flux for longer channels. As the channel length decreases both models’ predictions diverge from each other. While RELAP5/MOD3.3 has been shown to predict choking flow in large scale geometries, further investigation of data sets need to be done to determine if it is suited well for small channel lengths.

Commentary by Dr. Valentin Fuster
2017;():V009T15A047. doi:10.1115/ICONE25-67373.

In order to investigate the structure parameters evolution characteristics during flow regime transition process in slug flow, the vertical upward slug flow experiment in a wide range of liquid superficial velocity (0.03 < jl < 1.6 m/s) were conducted in a tubular test section with the inside diameter of 25.4 mm. Impedance void meters were employed to measure the void fraction of separated two parts corresponding to Taylor bubble and liquid slug. The present research studied the evolution of length ratio and void fraction in slug unit by keeping the liquid superficial velocity constant while increasing gas flow rate. New structure of slug unit in strong relation with transition process was observed. In specific, it was realized that the proportion of such special structure unit played an important role in transition from slug flow to churn-turbulent flow. The existing transition criteria from slug flow to churn-turbulent flow in upward two-phase flow (entrance effects model, flooding model, wake effects model, bubble coalescence model and Helmholtz instability model) were compared with the experimental identified results obtained by a new objective flow regime identification method, ReliefF-FCM algorithm. The results indicate that the transition model based on the wake effects could be the most appropriate choice to describe the mechanism of transition from slug flow to churn-turbulent flow in present experimental conditions.

Commentary by Dr. Valentin Fuster
2017;():V009T15A048. doi:10.1115/ICONE25-67438.

The modular High-Temperature Gas-cooled Reactor (HTGR) is one of the six generation IV advanced nuclear reactors. With the final purpose of operator training and licensing, the engineering simulation system (ESS) has been studied to model the pebble-bed type reactor core and has been successfully implemented into the full scope simulator of HTR-PM.

As stated in corresponding industrial standards, one important feature of the nuclear power plant simulator is real-time calculation, and the other one is simulation results with high fidelity (compared to design parameters or operational data in different stages). In ESS, each macro cross-section was in the form of polynomial by function of several variables (like burn-up, buckling, temperatures), the expression of which was finalized by multivariate regression analysis from large scattered database generated by the VSOP. Since the polynomial is explicit and prepared in advance, the macro cross-sections are quickly calculated in running ESS. However, some variables (such as temperature) in HTGR are in larger scope so that the polynomial is not easy to meet full range accuracy. One normal idea is to optimize the expression of polynomial, while another means was proposed and tested in present paper.

Other than focusing on the polynomials, a new method, called the fast searching, was described to significantly improve the accuracy of macro cross-section calculation while it was also fast to maintain the real-time feature. Instead of setting up a regression polynomial from the large cross-section database, the fast searching method treated the database as scatted points in the multi-dimension space, and aimed to locate the target position of unknown macro cross-section by fast searching and interpolating. Searching was to find the neighbouring database points around the target point in the multi-dimension space, which naturally improved the accuracy. While interpolating was to predict the macro cross-section of target point based on those neighbouring database points. To keep the searching and interpolating fast, the original database of macro cross-sections was analysed. A series of searching and interpolating methods have been described, programmed, tested and compared to find appropriate methods to calculate all the macro cross-sections in limited time cost. Finally, the fast searching method and its program was implemented into ESS to show better performances.

Commentary by Dr. Valentin Fuster
2017;():V009T15A049. doi:10.1115/ICONE25-67575.

Nuclear power plants (NPPs) typically have four classes of electrical supply systems to provide redundant and resilient power to ensure safe operation, cooling, and shutdown of NPPs. Class III emergency power traditionally uses standby diesel generators to fulfill equipment electrical demands. This requires a reliable source of diesel energy and reliance on one particular kind of fuel source for backup power.

During the North American blackout of 2003, hospitals had to run on backup generators that were mostly diesel fired. A prolonged blackout could have negatively affected hospital infrastructure and patient health. With current diesel generators, capable of supporting critical systems for up to 72 hours (for most NPPs), a prolonged loss of power or transient could similarly also have adverse effects on NPPs.

A Combined Heat and Power (CHP) system is a power generation system that generates electricity, in addition to concurrently having heating and cooling capabilities. Often, gas or steam turbines are used to generate electricity. CHP heating and cooling capabilities can be met via absorption coolers and heat pumps. Low-grade exhaust heat from turbines could be used for the absorption processes. Absorption coolers/heaters produce cold/warm fluid using a heat source via the vapor compression cycle, taking advantage of high affinity fluids. The absorbent allows for the refrigerant to boil at lower operating conditions, which allows for heat transfer.

CHPs’ that are capable of accepting natural gas as a form of fuel will increase resiliency of NPP power supply. This ensures a reduced risk of running out of fuel during prolonged transients due to continual supply via pipelines already in place.

The objective of this work is to generate a conceptual design of a multi-fuel source CHP system that is capable of at least accepting natural gas as an alternative to Class III diesel generators for a NPP. The system will be capable of supplying Class III power to Small Modular Reactor (SMRs), as well as commercial NPPs, such as CANDU, PWR, and BWR types.

Current state of the art Class III backup power systems, CHP systems, CHP thermodynamic cycles, multiple compatible fuels, as well as absorber heaters and coolers have been investigated. Ranking systems were used to determine the top three designs. The parameters included turbine type, efficiency, flow rates, operating temperature/pressure, fuel type, fuel to energy ratio, fuel availability, absorber chiller/heater fluid efficiency, operating temperature and thermal conductivity.

Based on the ranking system parameters, a thermodynamic model including mass, energy and entropy balance of an 8 MWe CHP with heating and cooling capability between the ranges of 15°C to 27°C was conceptually designed.

Commentary by Dr. Valentin Fuster
2017;():V009T15A050. doi:10.1115/ICONE25-67579.

With the purpose of providing preliminary assessments of specific locations in their ability to support a nuclear power plant using a quantitative method, a tool has been developed to accept input from users and provide an output of an evaluative score. With a focus on nuclear power plants that use water-based cooling and heat transfer methods, the tool considers environmental factors of the site and how they relate to nuclear power plant supportability; it reviews the socioeconomic factors of the region surrounding the site; and it considers the factors of the location that can impact the operational technical efficiency of a nuclear power plant. Using both spectral and binary scales, where a spectrum is assigned a minimum, maximum, and most preferred value, and the binary inputs allow for yes or no responses, as well as a weighting scale for each individual input factor and high level evaluation category, the tool provides a final score indicating how well the location should be capable of supporting a nuclear power plant.

Commentary by Dr. Valentin Fuster
2017;():V009T15A051. doi:10.1115/ICONE25-67589.

Faraday Cups (FCs) have been used since a long time to measure the beam current in particle accelerators. The charge collected by the beam stopper is used to measure the intensity of the beam current by an ammeter. Here, at IPR (Institute for Plasma Research) the Faraday cup is used to measure the deuteron beam current in 14 MeV neutron generator. The FC presently used is for 300 KeV beam and 500μA beam current whereas the new design is proposed for deuteron beam ranging from 300–400 KeV and 20mA beam current. The energy of Ion beam is usually very high in comparison to the work function of the material resulting in the emission of the secondary electrons which can escape from the cup and effect the current measurement. Simulations for the suppressor voltage required to suppress all the secondary electrons has been carried out using SIMION8.0 and LORENTZ. Materials like Copper, Molybdenum, Tantalum etc are commonly used in manufacturing of FCs and the present paper shows the inter-comparison between the use two FC material namely Copper and Molybdenum. The heat load deposited from the beam has been analytically simulated and proper cooling system has been suggested. Simulations for selection of FC material and heat load are needed to be carried out for the new FC and its cooling system, using analysis tools like SRIM, ANSYS WORKBENCH etc. The results of these analysis are reported in this paper. The final fabrication of the FC will be based on these simulations and analysis.

Topics: Simulation , Design
Commentary by Dr. Valentin Fuster
2017;():V009T15A052. doi:10.1115/ICONE25-67618.

In recent years, use of radiation beams or particle beams have been put to practical use for cancer therapy. In cancer radiation therapy, visualization of radiation patterns is absolutely necessary for precisely evaluating the dose distribution. Therefore, gel type or Fricke type dosimeters [1] are considered useful for visualization. In this study, we developed a new type of gel dosimeter using a doped polyvinyl alcohol (PVA) based solution. This gel uses a red color based chemical reaction that occurs when the active agent is separated. Irradiation of the gel with X-rays is sufficient to break the chemical bonds of the active agent. We irradiated different gel samples with X-rays from a Hitachi MBR-1520R-3 source under different configurations to test the gel performance. We used UV-VIS spectrometry to measure the absorbance of transmitted light through the gel. For the active agent, the absorbance is at a peak wavelength of 490 nm. The amount of absorbance is proportional to the number of interactions with X-rays. We irradiated the gel between 0.5Gy-10Gy with visualization of the gel by photography and spectrometry between each irradiation. The spectrometry was performed using a StellarNet Black Comet system observing the absorbance between 300nm and 600nm. The results show that as the X-ray dose increases, the gel transitions from a clear gel to a light pink gel and then to a red gel. All colors are translucent and allow for the passage of light. The first samples were done in clear plastic containers of 250 ml size. The containers were filled with gel to eliminate air and possible oxygen contamination. The second set of experiments repeated the first study but used metallic coins as X-ray shields. The regions covered by the coins were protected from the dose and remained clear. A sharp edge was observed at the edge of the coin. This implies that the gel does not diffuse and hence can represent a dose distribution as long as it is not mixed. The third samples were placed in disposable cells for measurement of absorbance. The absorbance had a peak in the vicinity wavelength of 490nm. The results confirmed the absorbance to be proportional for increasing applied dose. In summary, a color transition gel was developed for use in detecting irradiation dose from X-rays. This technique has potential application for visualization of dose during medical procedures.

Topics: X-rays , Dosimeters
Commentary by Dr. Valentin Fuster
2017;():V009T15A053. doi:10.1115/ICONE25-67619.

Coiled tube air heaters (CTAH) are heat exchangers currently under investigation at the University of California, Berkeley for application to Reheat Air-Brayton Combined Cycles (RACC). In a CTAH, molten fluoride salt on the tube side heats air on the shell side in a cross-counterflow arrangement, leading to high heat exchange effectiveness, while the coiled geometry provides good transient mechanical performance during thermal transients.

This paper describes the important phenomenology involved in CTAH design and operation, such as salt freezing, flow induced vibration, start-up and shut-down control, and air circulating power requirements. The THEEM simulation tool used in the design of CTAHs and its validation against simple water to air test sub-bundles is presented. Finally, an initial design for a demonstration 370 kWth CTAH closed air test loop is presented. This test loop design, when constructed, would enable demonstration of salt-to-air heating under conditions relevant to RACC power conversion (air pressures from 4 to 20 atm and air temperatures from 400 to 650°C).

Topics: Design
Commentary by Dr. Valentin Fuster
2017;():V009T15A054. doi:10.1115/ICONE25-67647.

Within the field of research and development of innovative nuclear reactors, in particular Generation IV reactors and Small Modular Reactors (SMR), the design and the improvement of safety systems play a crucial role. Among all the safety systems high attention is dedicated to passive systems that do not need external energy to operate, with a very high reliability also in the case of station blackout, and which are largely used in evolutionary technology reactors.

The aim of this work is the experimental and numerical analysis of a passive system that operates in natural circulation in order to study the mechanism and the efficiency of heat removal. The final goal is the development of a methodology that can be used to study this class of systems and to assess the thermal-hydraulic code RELAP5 for these specific applications. Starting from a commercial size system, which is the decay heat removal system of the experimental lead cooled reactor ALFRED, an experimental facility has been designed, built and tested with the aim of studying natural circulation in passive systems for nuclear applications. The facility has been simulated and optimized using the thermal-hydraulic code RELAP5-3D. During the experimental tests, temperatures and pressures are measured and the experimental results are compared with the ones predicted by the code.

The results show that the system operates effectively, removing the given thermal power. The code can predict well the experimental results but high attention must be dedicated to the modeling of components where non-condensable gases are present (condenser pool and surrounding ambient). This facility will be also used to validate the scaling laws among systems that operate in natural circulation.

Commentary by Dr. Valentin Fuster
2017;():V009T15A055. doi:10.1115/ICONE25-67714.

The Design Point (DP) performance of a Nuclear Power Plant (NPP) is fairly straightforward to establish for a given mass flow rate, turbomachinery compressor Pressure Ratio (PR) and reactor Core Outlet Temperature (COT). The plant components are optimum for that point but this is no longer the case if the plant’s operating conditions are changed for part-load performance. Data from tests or previous operating experiences are useful in determining typical part load performance of components based on characteristic maps. However, when individual components are linked in a plant, the range of operating points for part load performances are severely reduced. The main objective of this study is to derive Off-Design Points (ODPs) for the Simple Cycle Recuperated (SCR) and Intercooled Cycle Recuperated (ICR) when considering a temperature range of −35 to 50°C and COTs between 750 to 1000°C, using a modelling & performance simulation tool designed specifically for this study, which calculates the best operational equilibrium ODPs that are critical to the economics of the NPP. Results show that the recuperator High-Pressure (HP) side and reactor pressure losses alter the actual operating parameters (mass flow rate and compressor PR). The SCR yielded a drop in plant cycle efficiency of 1% for a 4% pressure loss in comparison to the ICR (5%) for the same amount of recuperator HP losses. Other parameters such as the precooler and recuperator Low-Pressure (LP) losses still retain the same operating inlet PRs and mass flow rates regardless of the magnitude of the losses. In the absence of characteristic maps in the public domain, the ODPs have been used to produce characteristic trend maps for first order ODP calculations. The analyses intend to aid the development of cycles for Generation IV NPPs specifically Gas Cooled Fast Reactors (GFRs) and Very High Temperature Reactors (VHTRs), where helium is the coolant.

Commentary by Dr. Valentin Fuster
2017;():V009T15A056. doi:10.1115/ICONE25-67715.

The Intercooled Cycle (IC) is considered as a viable alternative to the Simple Cycle Recuperated (SCR) and the Intercooled Cycle Recuperated (ICR), based on recent studies in a Nuclear Power Plant (NPP) configuration, which showed plant efficiencies of above 45%. The main difference in configuration is it does not utilise a recuperator. For part load performance, it is expected that the components of the IC will not operate at optimum conditions as the characteristics change. Thus the ability to demonstrate viable part load performance becomes an important requirement for the IC. The main objective of this study is to derive Off-Design Points (ODPs) from a known Design Point (DP) for a temperature range of −35 to 50°C and COTs between 750 to 1000°C. The ODPs have been calculated using a modelling & performance simulation tool designed specifically for this study and aim to provide a set of points that give operational equilibrium, which is critical to the economics of the plant. Results show that the intercooler alters the actual mass flow rate and compressor pressure ratio but the delta across an analysed range of 1 to 5% pressure loss shows a change of ∼9% in plant cycle efficiency, in comparison to the ICR (6%). Furthermore, the reactor pressure losses for IC has the lowest effect on plant cycle efficiency in comparison to the SCR and ICR. Characteristic trend maps have also been produced for the intercooler operation and the reactor and are applicable for NPP first order calculations. To that effect, it is also proposed to consider the intercooler pressure loss as a handle for ODP performance calculations. The analyses intend to bring further attention to the IC an alternative to current cycle configurations and to aid the development of cycles for Generation IV Nuclear Power Plants specifically Gas Cooled Fast Reactors (GFRs) and Very High Temperature Reactors (VHTRs), where helium is the coolant.

Commentary by Dr. Valentin Fuster
2017;():V009T15A057. doi:10.1115/ICONE25-67744.

Every plant aimed for energy production on large scale mandatorily needs to include systems able to ensure a proper release of the residual heat into the external environment; in the case of nuclear power plants, cooling towers are the most popular answer to this compelling task. The development of a dedicated numerical model can be a powerful tool for the simulation of the tower performance, but for quantifying the amount of heat released into the atmosphere, experimentally measured data are needed, in order to properly characterize the external environment in terms of outlet air temperature, outlet water temperature and outlet air relative humidity. Material properties, correlations and boundary conditions are among the many model parameters which can influence the model’s responses; variations of the computed quantities of interest (or “model responses”) induced by variations in the values of the model parameters can be expressed in terms of the sensitivities (i.e., functional derivatives) of the model responses with respect to the model parameters. The methodology applied in this paper in order to compute these sensitivities is the general adjoint analysis methodology (ASAM) for nonlinear systems; with this procedure, the exact values of the sensitivities of the model responses to all of the 47 model parameters are calculated exactly and efficiently by means of a single adjoint computation. The sensitivities can then be used within the “predictive modeling for coupled multi-physics systems” (PM_CMPS) methodology, aimed at yielding best-estimate predicted nominal values and uncertainties for model parameters and responses.

Commentary by Dr. Valentin Fuster
2017;():V009T15A058. doi:10.1115/ICONE25-67765.

Mixed neutron and gamma radiations require different shielding materials as their interaction with materials is different. Composites were developed in order to combine the shielding capabilities of different materials. However, their homogeneity is difficult to be assured which can lead to pinholes where radiation can penetrate. To avoid this problem, several materials arranged in layers can be used to shield against mixed radiations. Since the multilayer shielding can be made from any material in many configurations, the ant colony optimization (ACO) is a promising method because it deals with combinatorial optimization problems. The candidate materials are HDPE, boron, cadmium, gadolinium, tungsten, bismuth, and iron. Preliminary MCNP simulations were done to observe the effect of arrangements, thicknesses, and types of materials on the radiation spectrum. It was found that: (1) the final layer should be made of high density material, (2) an increase beyond certain thicknesses did not result in a significant increase in attenuation, and (3) there should be an optimum combination of material that can effectively shield against both neutrons and gamma rays.

Commentary by Dr. Valentin Fuster
2017;():V009T15A059. doi:10.1115/ICONE25-67797.

This work aims to develop an uncertainty analysis methodology for the propagation and quantification of the effects of nuclear cross-section uncertainties on important core-wide attributes, such as power distribution and core critical eigenvalue. Given the computationally taxing nature of this endeavor, our goal is to develop a methodology capable of preserving the accuracy of brute force sampling techniques for uncertainty quantification while realizing the efficiency of deterministic techniques. To achieve that, a reduced order modeling (ROM) approach is proposed to deal with the enormous size of the uncertainty space, comprising all the cross-section few-group parameters required in core-wide simulation. The idea is to generate a compressed representation of the uncertainty space, as represented by a covariance matrix, that renders sampling techniques computationally a feasible option for quantifying and prioritizing the various sources of uncertainties.

While the proposed developments are general to any reactor physics computational sequence, we customize our approach to the NESTLE [1]-TRITON [2] computational sequence, which will serve as a demonstrative tool for the implementation of our approach. NESTLE is a software used for core wide simulation, which relies on the few-group cross-sections to calculate core wide attributes over multiple cycles of depletion. Its input cross-sections are generated using a matrix of conditions evaluated using a lattice physics code, which in our implementation is done using the TRITON software of the ORNL’ SCALE suit. This manuscript presents one of the early steps towards this goal. Specifically, we focus here on the development of the algorithms for determining the reduced dimension of covariance matrix. Numerical experiment using the TRITON software is employed to demonstrate how the reduction is achieved.

Commentary by Dr. Valentin Fuster
2017;():V009T15A060. doi:10.1115/ICONE25-67844.

Tritium control is potentially a critical issue for Fluoride salt-cooled High-temperature Reactors (FHRs) and Molten Salt Reactors (MSRs). Tritium production rate in these reactors can be significantly higher compared to that in Light Water Reactors (LWRs). Tritium is highly permeable at high temperatures through reactor structures, especially. Therefore, heat exchangers with large heat transfer areas in FHRs and MSRs provide practical paths for the tritium generated in the primary salt migrating into the surroundings, such as Natural Draft Heat Exchangers (NDHXs) in the direct reactor auxiliary cooling system (DRACS), which are proposed as a passive decay heat removal system for these reactors.

A double-wall heat exchanger design was proposed in the literature to significantly minimize the tritium release rate to the environment in FHRs. This unique shell and tube heat exchanger design adopts a three-fluid design concept and each of the heat exchanger tube consists of an inner tube and an outer tube. Each of these tube units forms three flow passages, i.e., the inner channel, annular channel, and outer channel. While this type of heat exchangers was proposed, few such heat exchangers have been designed in the literature, taking into account both heat and tritium mass transfer performance.

In this study, a one-dimensional heat and mass transfer model was developed to assist the design of a double-wall NDHX for FHRs. In this model, the molten salt and air flow through the inner and outer channels, respectively. A selected sweep gas acting as a tritium removal medium flows in the annular channel and takes tritium away to minimize tritium leakage to the air flowing in the outer channel. The heat transfer model was benchmarked against a Computational Fluid Dynamics (CFD) code, i.e., ANSYS Fluent. Good agreement was obtained between the model simulation and Fluent analysis. In addition, the heat and mass transfer models combined with non-dominated sorting in generic algorithms (NSGA) were applied to investigate a potential NDHX design in Advanced High-Temperature Reactor (AHTR), a pre-conceptual FHR design developed by the Oak Ridge National Laboratory. A double-wall NDHX design using inner and outer fluted tubes was therefore optimized and compared with a single-wall design in terms of performance and economics.

Commentary by Dr. Valentin Fuster
2017;():V009T15A061. doi:10.1115/ICONE25-67890.

In the decommissioning of nuclear power plants, the long-term management of radioactive waste of fuel debris, etc. are necessary. In the process, hydrogen which is the flammable gas is generated by the decomposition of water by radiation. Therefore, it is important to ensure the safety of the waste storage container to reduce the concentration of hydrogen gas, and to keep below the explosion limit (4%). Consequently, a basic experiment to investigate the effectiveness of the waste storage container with the flammable gas concentration reduction mechanism using the passive autocatalytic recombiner (PAR) has been planned. The present study describes the research plan to use a small modeled experimental apparatus.

In addition, in order to clarify quantitatively natural convection behavior of hydrogen gas due to the decay heat of radioactive materials in the long-term waste storage container, preliminary analyses were performed on the system of a small-scale experimental apparatus in which specification of the long-term radioactive waste storage container is simply simulated. From the present results, the perspective which can predict natural convection phenomena in the long-term waste storage container numerically was obtained.

Commentary by Dr. Valentin Fuster
2017;():V009T15A062. doi:10.1115/ICONE25-67895.

Clarifying thermal-hydraulic characteristics in a nuclear reactor core is important in particular to enhance the thermo-fluid safety of nuclear reactors. Spacers installed in subchannels of fuel assemblies have the role of keeping the interval between adjacent fuel rods constantly. Similarly, in case of PWR the spacer has also the role as the turbulence promoter. When the transient event occurs, two-phase flow is generated by boiling of water due to heating of fuel rods. Therefore, it is important to confirm the two-phase flow behavior around the spacer. So, the effect of the spacer affecting the two-phase flow was investigated experimentally at forced convective flow condition. Furthermore, in order to improve the thermal safety of current light water reactors, it is necessary to clarify the two-phase flow behavior in the subchannels at the stagnant flow condition. So, the bubbly flow data around a simulated fuel rod were obtained experimentally at the stagnant flow condition. A wire-mesh sensor was used to obtain a detailed two-dimensional void fraction distribution around the simulated spacer and fuel rod. As a result of this research, the bubbly behavior around the simulated spacer and fuel rod was qualitatively revealed and also bubble dynamics in the sub-channels at the conditions of forced convective and stagnant flows were evaluated. The present experimental data are very useful for verifying the detailed three-dimensional two-phase flow analysis codes.

Commentary by Dr. Valentin Fuster
2017;():V009T15A063. doi:10.1115/ICONE25-67927.

Two-phase flow through porous media should be well understood to develop a severe accident analysis code not only for light water reactor but also sodium cooled fast reactor (SFR). When a core disruptive accident occurs in SFR, the fuel inside the core may become melted and interacts with the coolant. As a result, gas-liquid two-phase flow will be formed in the debris bed, which may have porous nature depending on the cooling process. Thus, as first step, present work focuses on the characteristics of pressure drop in single- and two-phase flows in different porous media conditions (porous size, liquid and gas flow velocity). In addition, in order to construct an experimental database, the measured pressure drop under different conditions was compared with existing correlations.

Topics: Pressure drop
Commentary by Dr. Valentin Fuster
2017;():V009T15A064. doi:10.1115/ICONE25-67931.

When a severe accident occurs, decommissioning work becomes important task. In the decommissioning work after the severe accident, establishing the way to estimate the sedimentation place of molten debris is important. However, the technique to estimate exactly sedimentation place has not been enough. Therefore, the detailed and phenomenological numerical simulation code named JUPITER for predicting the molten core behavior is under development. The comparison between experimental and numerical results is necessary to clarify the validity of the numerical analysis code. This study provides the experimental data for a BWR to examine the numerical simulation code in order to contribute to progress of the decommissioning work.

Commentary by Dr. Valentin Fuster
2017;():V009T15A065. doi:10.1115/ICONE25-67936.

Two new multi-objective differential evolution (DE) algorithms are used to optimize heterogeneous low-enriched uranium + mixed oxide fuel assemblies for use in a pressurized water reactor. The objectives were to maximize plutonium content and minimize the power peaking factor. A performance comparison to a genetic algorithm is used to evaluate the applicability of DE algorithms to nuclear fuel assembly design optimization problems. Results show that DE performs highly competitively against a more established algorithm and can arguably better represent the trade-off between both objectives through greater variety in the number of different pin arrangements explored and a higher reliability in finding the ‘true’ Pareto-front.

Commentary by Dr. Valentin Fuster
2017;():V009T15A066. doi:10.1115/ICONE25-67937.

The critical heat flux (CHF) during saturated pool boiling of water was investigated experimentally using a honeycomb porous plate attached to a heated surface. In a previous study, we demonstrated that the CHF by a honeycomb porous plate was enhanced to more than twice compared with that of a plain surface. According to the proposed capillary limit model, the CHF can be increased by decreasing the thickness of the honeycomb porous plate. However, the CHF could not be greatly enhanced when the thickness of the honeycomb porous plate was comparable to the thickness of the thin liquid film (the macro-layer thickness) formed beneath coalescent vapor bubbles. The results of this study showed that honeycomb porous plates for CHF enhancement in saturated pool boiling should be constructed by the superposition of two kinds of porous materials and that each of the honeycomb porous plates must fulfill two conditions. First, a honeycomb porous plate simply attached to a heated surface should have very fine pores to supply water to the heated surface due to strong capillary action, and the honeycomb porous plates should be as thin as possible to decrease the pressure drop caused by internal water flow. Second, the other honeycomb porous plate, stacked on top of the thin honeycomb porous plate, needs to be structured to hold a sufficient amount of water in order to prevent the inside of the honeycomb porous plate from drying out during the bubble hovering period on the plate.

Commentary by Dr. Valentin Fuster
2017;():V009T15A067. doi:10.1115/ICONE25-67984.

In this study we describe an experimental system designed to simulate the conditions of transient freezing which can occur in abnormal behaviour of molten salt reactors (MSRs). Freezing of coolant is indeed one of the main technical challenges preventing the deployment of MSR. First a novel experimental technique is presented by which it is possible to accurately track the growth of the solidified layer of fluid near a cold surface in an internal flow of liquid. This scenario simulates the possible solidification of a molten salt coolant over a cold wall inside the piping system of the MSR. Specifically, we conducted measurements using water as a simulant for the molten salt, and liquid nitrogen to achieve high heat removal rate at the wall. Particle image velocimetry and planar induced fluorescence were used as diagnostic techniques to track the growth of the solid layer. In addition this study describes a thermo-hydraulic model which has been used to characterise transient freezing in internal flow and compares the said model with the experiments. The numerical simulations were shown to be able to capture qualitatively and quantitatively all the essential processes involved in internal flow transient freezing. Accurate numerical predictive tools such the one presented in this work are essential in simulating the behaviour of MSR under accident conditions.

Commentary by Dr. Valentin Fuster
2017;():V009T15A068. doi:10.1115/ICONE25-67985.

Idaho State University (ISU), with support from Idaho National Laboratory, is actively engaged in enhancing nuclear power plant risk modeling. The ISU team is significantly increasing the understanding of non-containment, nuclear power plant component performance under flooding conditions. The work involves experimentation activities and development of mathematical models, using data from component flooding experiments. The research consists in developing experimentation procedures that comprised small scale component testing, followed by simple and then complex full scale component testing. The research is taking place in the Component Flooding Evaluation Laboratory (CFEL). Tests in CFEL will include water rise, spray, and wave impact experiments on passive and active components.

Initial development work focused on small scale components, radios and simulated doors, that served as a low-risk and low-cost proof-of-concept options. Following these tests, full-scale component tests were performed in the Portal Evaluation Tank (PET). The PET is a semi-cylindrical 7500-1 capacity steel tank, with an opening to the environment of 2.4 m. × 2.4 m. The opening allows installation of doors, feedthroughs, pipes, or other components. The first set of experiments with the PET were conducted in 2016 using hollow doors subjected to a water rise scenario. Data collected during the door tests is being analyzed using Bayesian regression methods to determine the parameters of influence and inform future experiments.

A practical method of simulating full scale wave impacts on components and structures is also being researched to further enhance CFEL capabilities. Early on, the team determined full scale wave impacts could not be simulated using traditional wave flumes or pools; therefore, closed conduit flow is under consideration. Computational fluid dynamics software is being used to simulate fluid velocities associated with tsunami waves of heights up to 6-m, and to design a wave impact simulation device capable of accurately recreating a near vertical wave section with variable height and fluid velocity.

The component flooding simulation activities associated with this project involve use of smoothed particle dynamics codes. These particle-based simulation methods do not require a mesh to be applied to the fluid, which allows for more natural flows to be simulated.

Finally, CFEL can be described as a pioneering element, comprised of several ongoing research and experimental projects, that are vital to the development of risk analysis methods for the nuclear industry.

Commentary by Dr. Valentin Fuster
2017;():V009T15A069. doi:10.1115/ICONE25-68001.

The RBWR-TR is a thorium-based reduced moderation BWR (RBWR) with a high transuranic (TRU) consumption rate. It is charged with LWR TRU and thorium, and it recycles all actinides an unlimited number of times while discharging only fission products and trace amounts of actinides through reprocessing losses. This design is a variant of the Hitachi RBWR-TB2, which arranges its fuel in a hexagonal lattice, axially segregates seed and blanket regions, and fits within an existing ABWR pressure vessel. The RBWR-TR eliminates the internal axial blanket, eliminates absorbers from the upper reflector, and uses thorium rather than depleted uranium as the fertile makeup fuel. This design has been previously shown to perform comparably to the RBWR-TB2 in terms of TRU consumption rate and burnup, while providing significantly larger margin against critical heat flux.

This study examines the uncertainty in key neutronics parameters due to nuclear data uncertainty. As most of the fissions are induced by epithermal neutrons and since the reactor uses higher actinides as well as thorium and 233U, the cross sections have significantly more uncertainty than in typical LWRs. The sensitivity of the multiplication factor (keff) to the cross sections of many actinides is quantified using a modified version of Serpent 2.1.19 [1]. Serpent [2] is a Monte Carlo code which uses delta tracking to speed up the simulation of reactors; in this modified version, cross sections are artificially inflated to sample more collision, and collisions are rejected to preserve a “fair game.” The impact of these rejected collisions is then propagated to the multiplication factor using generalized perturbation theory [3]. Covariance matrices are retrieved for the ENDF/B-VII.1 library [4], and used to collapse the sensitivity vectors to an uncertainty on the multiplication factor. The simulation is repeated for several reactor configurations (for example, with a reduced flow rate, and with control rods inserted), and the difference in keff sensitivity is used to assess the uncertainty associated with the change (the uncertainty in the void feedback and the control rod worth). The uncertainty in the RBWR-TR is found to be dominated by the epithermal fission cross section for 233U in reference conditions, although when the spectrum hardens, the uncertainty in fast capture cross sections of 232Th becomes dominant.

Commentary by Dr. Valentin Fuster
2017;():V009T15A070. doi:10.1115/ICONE25-68010.

Measurement accuracy of flow velocity by phased array UVP method was confirmed in order to develop a basic telemetry system for detections of leakage points. The basic telemetry system which consists of a robot arm for moving ultrasonic sensor and a phased array sensor for wide range area measurement of flow velocity was developed. Additionally, flow mapping of water flow around imitated leakage was conducted by the developed system and PIV method. As a result, the velocities by both methods were in agreement with 20 %. The possibility of measuring two-dimensional velocity of flow near the leakage point by the developed telemetry system was verified. Consequently, the basic telemetry system for the detection of the leakage was established.

Commentary by Dr. Valentin Fuster

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