In Memoriam: Professor Pavel Leonidovich Kirillov (20.08.1927 08.10.2021)

Professor Pavel L. Kirillov has died on October 8th, 2021, on his 95th year after a life as a husband, father, and an internationally renowned scientist, researcher, and educator in the field of nuclear engineering, thermalhydraulics, heat transfer, and two-phase flow. He was a passionate and dedicated in everything, what he has done and leaves an incredible legacy to the profession. He was born on August 20th, 1927 in Russia, and received his M.A.Sc. degree in thermal physics in 1950 (Moscow Power-Engineering Institute (MPEI) (МосковскийЭнергетическйИнститут (МЭИ)), Faculty of Physics and Power Engineering (Физико-ЭнергетическийФакультет), Ph.D. and Doctor of Technical Sciences degrees in 1959 and 1969, respectively. Professor Pavel Leonidovich Kirillov was a respected technical leader, mentor, and friend to innumerable students, researchers, scientists, and engineers, and he will be sadly missed by all, who had the privilege to know him. He was an outstanding contributor in every aspect of his prolific work and career in the true traditions of technical excellence and critical thinking, and his irreplaceable loss is deeply felt worldwide.

A Short Inroad to Optimized Compactification of Composite Neutronic Shields

Compact neutronic shields for mobile nuclear reactors or accelerator-based neutron beams are known to be optimized multilayered composites. This paper is a simplified short inroad to the complex problem of optimizing the design of such shields when they attenuate a neutron beam to extremise certain quality criteria, in plane geometry, subject to equality and inequality constraints. In the equality constraints, the interfacial polychromatic neutron fluxes are solutions to course-mesh finite-difference holonomic state equations. The set of these interfacial fluxes act as state variables,while the set of layer thicknesses, or their poisoning (by added neutron absorbers) concentrations are decision variables. The entire procedure is then demonstrated to be reducible to standard Kuhn-Tucker semi-linear programming that may also lead robustly to an optimal sequence for these layers.

Experimental Investigation of Melt Coolability Behaviour in an Ex-Vessel Core Catcher - the Effect of Flooding Time

Some of the advanced nuclear reactors employ an ex-vessel core catcher to mitigate core melt scenarios by stabilizing and cooling the corium for prolonged period by strategically flooding it. The side indirect cooling with top flooding strategy described in this study may lead to water ingression either through the melt crust which may lead to interaction between un-oxidised metal in the melt and water leading to hydrogen production. In order to avoid this deleterious scenario, water ingression into the bulk of the melt should be avoided. The studies described in this manuscript show that water ingression depends on the flooding strategy, i.e. the time delay between top flooding and melt relocation. Two experiments under identical conditions of simulant temperature, melt material and test section geometry were conducted with simulated decay heat of 1 MW/m3. Sodium borosilicate glass was used as the corium simulant. In the first experiment, water was flooded onto the top of melt pool soon after melt relocation. In the second experiment, water flooding at the top of melt pool was made after 30 minutes of the melt relocation. The results show that a finite time delay of introduction of water onto the top of the melt pool is paramount to engender the development of a stable crust around the melt and therefore eliminating water ingression into melt pool and ensuring controlled coolability of the melt.

Thermal Stratification in a Pool with Submerged Heater Under Low Frequency Excitation

Thermal stratification has potential applications in the nuclear and solar industries. Thermal performance of passive residual heat removal systems and solar heaters is affected by the thermal stratification in a pool. Under the seismic condition, thermal stratification behavior of liquid in the pool has never been studied and reported in the literature. The present work focuses on the experimental investigation of thermal stratification in a pool under the seismic condition with the horizontally mounted heater simulating heat exchanger. Effect of heater submergence depth, frequency of excitation and amplitude of displacement on the thermal stratification has been studied. It was observed that the heater submergence depth significantly influences the thermal stratification in a pool. When a pool is subjected to an external excitation, the pool water separates into two zones; convective and impulsive. If the heater submergence depth in the impulsive zone, excitation effects are not found. If heater submergence depth is close to convective zone, significant effects are observed. However, it was observed that only first mode of excitation with large amplitude helps to achieve complete thermal mixing and higher modes of excitation have the minimal on the mitigating of thermal stratification. Non-dimensional stratification number has been evaluated to explain the mitigation of thermal stratification with seismic excitation.

Uncertainty Quantification in Support of Severe Accident Analysis Code User Confidence Using MELCOR-DAKOTA

Uncertainty and Sensitivity analysis methods are often used in severe accident analysis for validating the complex physical models employed in the system codes that simulate such scenarios. This is necessitated by the large uncertainties associated with the physical models and boundary conditions employed to simulate severe accident scenarios. The input parameters are sampled within defined ranges based on assigned probability distribution functions (PDFs) for the required number of code runs/realizations using stochastic sampling techniques. Input parameter selection is based on their importance to the key FOM, which is determined by the parameter identification and ranking table (PIRT). Sensitivity analysis investigates the contribution of each uncertain input parameter to the uncertainty of the selected FOM. In this study, the integrated severe accident analysis code MELCOR was coupled with DAKOTA, an optimization and uncertainty quantification tool in order to investigate the effect of input parameter uncertainty on hydrogen generation. The methodology developed was applied to the Fukushima Daiichi unit 1 NPP accident scenario, which was modelled in another study. The results show that there is approximately 22.46% uncertainty in the amount of hydrogen generated as estimated by a single MELCOR run given uncertainty in selected input parameters. The sensitivity analysis results also reveal that MELCOR input parameters; COR_SC 1141(Melt flow rate per unit width at breakthrough candling) , COR_ZP (Porosity of fuel debris beds) and COR_EDR (Characteristic debris size in core region) contributed most significantly to the uncertainty in hydrogen generation.

Testing the Modified Subchannel TEMPA-SC Code in Comparison with Experiments and Other Computer Codes

The paper describes a modified version of the TEMPA-SC computer program designed to calculate temperature fields in bundles of rods cooled by a supercritical pressure medium. This version of the program is based on the subchannel method that was used in the TEMPA-1F program, developed earlier in the OKB "GIDROPRESS" for calculating heat and mass transfer in the core of VVER-type reactors cooled by single-phase water at subcritical pressure. As the relations that close the system of equations of mass, momentum, and energy conservation, the new version of the program includes correlations for calculating heat transfer and friction resistance, taking into account the strong dependence of the properties of the coolant on temperature and pressure. In particular, the use of the universal calculation model of heat transfer, developed by the authors of this paper, allows us to perform calculations in a wide range of flow parameters of various media, including the modes of normal, improved and deteriorated heat transfer. The results of tests of the TEMPA-SC program are presented in comparison with the available experimental data for water and modeling media (carbon dioxide, freons R-12 and R-134a) at supercritical pressures, as well as with the published data of calculations by using similar subchannel programs (COBRA-SC, ASSERT-PV) and CFD codes. A satisfactory agreement between the calculated and experimental data is shown.

Prediction Method for Condensation Heat Transfer in the Presence of Non-condensable Gas for CFD Applications

The objective of this study was to present a prediction method for condensation heat transfer in the presence of non-condensable gas (air or nitrogen) for CFD (computational fluid dynamics) analyses, where physical quantities in the computational cells in contact with the structural wall are generally used. First by using existing temperature distributions T(y) in the turbulent boundary layer along a flat plate as functions of the distance y from the condensation surface, we evaluated the distribution of condensation heat flux qc,pre(y) from the gradient of steam concentration, we derived a modification factor η(y+) as a function of the dimensionless distance y+ to obtain a good agreement with qc,cal calculated by the qc correlation defined by using the bulk quantities; and we obtained qc,mod(y)/qc,cal = 0.90-1.10 for the region of y+ > 17. Second we modified the local Sherwood number Sh(x) for flat plates for the boundary layer thickness d and obtained the function Sh(d). An existing qc correlation for flat plates as a function of Sh(d) was applied to predict the distribution of the local value qc,pre(y), and qc,pre(y)/qc,cal = 0.95-1.15 in the best case was obtained for the region of y+ > 30. Finally a correlation of the local Sherwood number Sh(y) was derived from the temperature distributions T(y) as a function of the local Reynolds number Re(y).

Thermal Fatigue of Pwr Coolant System Loop Drain Lines Due to Outflow Activities

Thermal cycling induced fatigue is widely recognized as one of the major contributors to the damage of nuclear plant piping systems, especially at locations where turbulent mixing of flows with different temperature occurs. Thermal fatigue caused by swirl penetration interaction with normally stagnant water layers has been identified as a mechanism that can lead to cracking in dead-ended branch lines attached to pressurized water reactor (PWR) primary coolant system. EPRI has developed screening methods, derived from extensive testing and analysis, to determine which lines are potentially affected as well as evaluation methods to perform evaluations of this thermal fatigue mechanism for the U.S. PWR plants. However, recent industry operating experience (OE) indicate that the model used to predict thermal fatigue due to swirl penetration is not fully understood. In addition, cumulative effects from other thermal transients, such as outflow activities, may also contribute to the failure of the RCS branch lines. In this paper, we report direct OE from one of our PWR units where thermal fatigue cracking is observed at the RCS loop drain line close to the welded region of the elbow. A conservative analytical approach that takes into account the influence of thermal stratification, in accordance with ASME Class 1 piping stress method, is also proposed to evaluate the severity of fatigue damage to the RCS drain line, as a result of transients from outflow activities. Finally, recommendations are made for future operation and inspection based on results of the evaluation.

Simulated Environmental Tests for Selected ATF Cladding Solutions

Current development on advanced technology fuel (ATF) claddings is aiming at improved high temperature integrity of new cladding solutions that are based on the existing zirconium claddings. To assess their performance for commercial use, their thorough characterization is essential. The primary requirement for the cladding materials is the ability to tolerate loss of cooling for a significant period without failing. The tests in this work were performed on different types of coated Zr-alloys in a high temperature furnace in flowing steam conditions at 1100 °C/ 60 min, 1200 °C/ 30 min and 1300 °C / 5 min. In addition, exposures were performed in pressurized water reactor (PWR) water chemistry to confirm the material viability in normal light water reactor (LWR) operating conditions. After PWR and steam tests, the exposed specimens were studied using a Zeiss Crossbeam 540 field emission gunscanning electron microscope (FEG-SEM) equipped with a semi-quantitative energy dispersive X-ray spectrometer (EDS). Most of the tested specimens indicated detached coating layer. Varying amounts of cracking in the coatings were present. Some of the cracks extended into the base material. Based on this study, further development is needed.