International Standard Problem Research Articles

Using the Cuckoo Optimization to Initialize the Harmony Memory in Harmony Search Algorithm to Find New Hybrid (CSHS) Algorithm

The current paper seeks to present a hybrid harmony search algorithm, using CS and HS. The proposed algorithm takes advantage of using a method to initialize the harmony memory (HM) with the help of the cuckoo search optimization algorithm. The HS algorithm is inspired by the method used by musicians to create and enhance harmony in music. It follows 3 main principles of pitch adjustment (PA), HM consideration (HMC), and random selection (RS). The performance of HS can be affected by its poor accuracy in optimization and speed of convergence, 2 main issues with it. However, to improve the HS algorithm, Cuckoo search can be used as it can do a local search using one parameter only, other than the size of the population, making it work more efficiently. This current method has been tested for its validity and efficiency in performance by being implemented on many international standard optimization problems. The results confirm the fact that the performance of the currently proposed algorithm is better and more efficient in finding solutions compared to HS and other algorithms.

Analyses of International Standard Problem ISP-47 TOSQAN experiment with containmentFOAM

The ISP-47 TOSQAN experiment was analyzed with containmentFOAM which is an open-source CFD code based on OpenFOAM. The containment phenomena taking place during the experiment are gas mixing, stratification and wall condensation in a mixture composed of steam and non-condensable gas. The k-ω SST turbulence model was adopted with buoyancy turbulence models. The wall condensation model used is based on the diffusion layer approach. We have simulated the full TOSQAN experiment which had a duration 20000 s. Sensitivity studies were conducted for the buoyancy turbulence models with SGDH and GGDH and there were not significant differences. All the main features of the experiments namely pressure history, temperature, velocity and gas species evolution were well predicted by containemntFOAM. The simulation results confirmed the formation of two large flow stream circulations and a mixing zone resulting by the combined effects of the condensation flow and natural convection flow. It was found that the natural convection in lower region of the vessel devotes to maintain two large circulations and to be varied the height of the mixing zone as result of sensitivity analysis of non-condensing wall temperature. The computational results obtained with the 2D mesh grid approach were comparable to the experimental results.

ANDROID-BASED GOOGLE TRANSLATE TRAINING FOR MATHEMATICS SUBJECT TEACHERS' DELIBERATIONS IN UNDERSTANDING HIGH-LEVEL REASONING TEST ITEM IN SOPPENG REGENCY, SOUTH SULAWESI

English is the social language of the world, but teachers still have limitations, especially in translating English texts. This service aims to provide counselling, training and mentoring for groups of teachers involved in consulting mathematics teachers. The activity involved 28 mathematics teachers as delegates from each school or madrasa in Soppeng Regency. Each stage is carried out pre-test and post-test. Data analysis used percentages and paired t-tests with a significance level of 0.05. The results of the service showed that 1) participants' body movements before and after the activity increased in a better direction, 2) participants' understanding of translating texts in the form of words, syllables, sentences and paragraphs increased, 3) participants' skills in translating English texts increased by an average of 80%, 4) there was a significant difference in improvement before and after counselling, training, and teacher assistance in translating high-level reasoning English text sources. Therefore, improving the ability and skills to translate English texts can use three stages: counselling, training and mentoring using text sources from international journals or international standard math problems

BETHSY 9.1b test calculation with TRACE using 3D vessel component

Recently, several advanced multidimensional computational tools for simulating reactor system behavior during real and hypothetical transient scenarios were developed. One of such advanced, best-estimate reactor systems codes is TRAC/RELAP Advanced Computational Engine (TRACE), developed by the U.S. Nuclear Regulatory Commission. The advanced TRACE comes with a graphical user interface called SNAP (Symbolic Nuclear Analysis Package). It is intended for pre- and post-processing, running codes, RELAP5 to TRACE input deck conversion, input deck database generation etc. The TRACE code is still not fully development and it will have all the capabilities of RELAP5. The purpose of the present study was therefore to assess the 3D capability of the TRACE on BETHSY 9.1b test. The TRACE input deck was semi-converted (using SNAP and manual corrections) from the RELAP5 input deck. The 3D fluid dynamics within reactor vessel was modeled and compared to 1D fluid dynamics. The TRACE 3D calculation was compared both to TRACE 1D calculation and RELAP5 calculation. Namely, the geometry used in TRACE is basically the same, what gives very good basis for the comparison of the codes. The only exception is 3D reactor vessel model in case of TRACE 3D calculation. The TRACE V5.0 Patch 1 and RELAP5/MOD3.3 Patch 4 were used for calculations. The BETHSY 9.1b test (International Standard Problem no. 27 or ISP-27) was 5.08 cm equivalent diameter cold leg break without high pressure safety injection and with delayed ultimate procedure. BETHSY facility was a 3-loop replica of a 900 MWe FRAMATOME pressurized water reactor. In general, all presented code calculations were in good agreement with the BETHSY 9.1b test. The TRACE 1D calculation results are comparable to RELAP5 calculated results. For some parameters they are better, this is mostly due to better tuning of the break flow, what influences timing of the transient. When comparing TRACE 1D and TRACE 3D calculation, the latter is slightly better. One reason for comparable results is already good agreement of 1D calculations and there was not much space to further improve the results. The other reason may be that in the facility the phenomena were mostly one dimensional (for example, external downcomer was used for reactor vessel modeling). However, when 3D behavior of the heater rod temperatures was investigated, the advantage of three dimensional treatment was clearly demonstrated.

CFD analysis of water vapor condensation in large containments: Numerical model, verification and validation

In the event of a loss of coolant accident (LOCA) in a Pressurized Water Reactor (PWR), large quantities of water vapor and hydrogen might be released into the containment building. Increasing pressure and the risk of hydrogen deflagration can threaten the containment integrity. Water vapor condensation on cold containment structures modifies the pressure load on the containment structures. On the one hand, the total pressure reduces due to the decrease of the water vapor content of the containment atmosphere. On the other hand, the concentration of hydrogen increases, which can reach locally the threshold of deflagration or detonation. A Computational Fluid Dynamics (CFD) model is developed and tested in order to predict transients of the pressure and the concentrations of gaseous species in reactor containments after a LOCA. Since the free volume of the containment of typical French PWRs is between 50,000 m3 and 80,000 m3, the CFD model requires significant simplifications.The physical model and its simplifications as well the implementation in the CEA OpenSource CFD code TrioCFD are described. Model tests are presented in two steps. In a first step, verification tests are discussed to show that the condensation model is implemented correctly in the CFD code. In a second step, the model is validated against experimental data of the International Standard Problem ISP47. In this benchmark, steady state conditions between vapor injection and condensation were reached experimentally in the MISTRA test containment of the CEA. In phase A of the benchmark, an equilibrium was achieved between water vapor injection and water vapor condensation on temperature controlled cold walls. One incondensable gas, namely air, was present in the test vessel. In phase B, a second incondensable gas was added to the containment atmosphere, namely helium. Both steady state situations were analyzed with the CFD model. The calculations represent well the experiment when the predominant condensation paths are modelled. It is shown that the pressure in the containment vessel as well as the mean mass fractions of water vapor and air, as well as of helium in phase B, are calculated in accordance to the experiment. The temperature in the containment is overestimated. Measured vertical profiles of the species concentrations are reproduced correctly.

Development, Validation, and Application of the Turbulent Combustion Model for 3D CFD Code CYCAS

The CYCAS code is a 3D CFD code developed independently by the China Nuclear Power Research Institute, aiming at hydrogen safety analysis in containments. In order to equip the CYCAS code with the capability of combustion analysis, a combustion model is screened and developed, and validated with two cases from International Standard Problems. Then a combustion analysis is conducted with a containment to demonstrate the capability of CYCAS to evaluate the pressure and thermal load of combustion. Firstly, the mature and widely-used combustion model, i.e., the Burning Velocity Model (BVM) with Zimont correlation is developed for the CYCAS code in this study. Thereafter, the lamiar flame speed correlation for CYCAS is screened based on the containment atmospheric condition during the accidents. Then, two cases from the THAI facility are calculated to validate the model. The results show that the combustion model in CYCAS manages to have reasonable prediction on the slow deflagration for the gas mixture both with and without steam. Finally, a containment combustion analysis is conducted with the initial hydrogen distribution calculated from a postulated accident. The analysis shows that the CYCAS manages to calculate the pressure and thermal load of the combustion to evaluate the influence of the combustion on the containment integrity and equipment survivability.

Intelligent Recognition Method of Vehicle Path with Time Window Based on Genetic Algorithm

Based on particle filter and improved cuckoo genetic algorithm, an algorithm for intelligent vehicle path recognition with a time window is designed. Particle filter (PF) is an influential visual tracking tool; it relies on the Monte Carlo Chain framework and Bayesian probability, which are essential for intelligent monitoring systems. The algorithm first uses particle filters for visual tracking and then obtains the current operating environment of the vehicle, then performs cluster analysis on customer locations, and finally performs path recognition in each area. The algorithm not only introduces particle filters, which are advanced visual tracking, but also improves the cuckoo search algorithm; when the bird’s egg is found by the bird’s nest owner, it needs to randomly change the position of the entire bird’s nest, which speeds up the search speed of the optimal delivery route. Analyze and compare the hybrid intelligent algorithm and the cuckoo search algorithm. Finally, the international standard test set Benchmark Problems is used for testing. The experimental outcomes indicated that the new hybrid intelligent approach is an effective algorithm for handling vehicle routing tasks with time windows.

Evidence-based background for constrained uncertainty quantification in a core transient analysis

The paper discusses some topics related to a validation of multi-physics modeling. Since validation belongs to a category of decision-making processes, it should be prepared by dedicated scientific researches. In particular, the validation means a kind of characterization of the predictive capability maturity of given tools, libraries and calculational models. The compliance criteria might be expressed in terms of achievable accuracy or, inverse, in terms of somehow identified and quantified uncertainties. Despite the uncertainties, as such, might not be measured or compared with something measurable all the judgments should rely on reality, i.e., be supported by an evidence-based background. In practice it does require to eliminate subjective statements, if any, replacing them with something inferred from objective observations, including representative integral experiments. Unfortunately, in many fields like, among others, multi-physics simulations, we have not statistically sufficient number of high-fidelity and confident experiment-based benchmarks. In addition, because of technological and safety constraints, what is needed lies, largely, beyond the experimental domain. This is why, assessors have to rely on numerous, but partially representative experiments. These data could be treated using one or other Data Assimilation techniques to provide correction factors and uncertainties to single- and few-physics modules all having an evidence-based background. Then, coupling these pre-validated modules and their uncertainties, we could estimate uncertainties (and accuracies) in an application domain using one of wide range error propagation techniques. Thus, combining experiments-grounded and calibrated uncertainties, we are providing consistent, evidence-based background for validation. The last phase – constrained uncertainty propagation – was illustrated with an example of one international standard problem on transient initiated in LWR core by inadvertent Control Rod ejection.

CFD/RELAP5 coupling analysis of the ISP No. 43 boron dilution experiment

Multi-dimensional coupling analysis is a research hot spot in nuclear reactor thermal hydraulic study and both the full-scale system transient response and local key three-dimensional thermal hydraulic phenomenon could be obtained simultaneously, which can achieve the balance between efficiency and accuracy in the numerical simulation of nuclear reactor. A one-dimensional to three-dimensional (1D-3D) coupling platform for the nuclear reactor multi-dimensional analysis is developed by XJTU-NuTheL (Nuclear Thermal-hydraulic Laboratory at Xi'an Jiaotong University) based on the CFD code Fluent and system code RELAP5 through the Dynamic Link Library (DLL) technology and Fluent user-defined functions (UDF). In this paper, the International Standard Problem (ISP) No. 43 is selected as the benchmark and the rapid boron dilution transient in the nuclear reactor is studied with the coupling code. The code validation is conducted first and the numerical simulation results show good agreement with the experimental data. The three-dimensional flow and temperature fields in the downcomer are analyzed in detail during the transient scenarios. The strong reverse flow is observed beneath the inlet cold leg, causing the de-borated water slug to mainly diffuse in the circumferential direction. The deviations between the experimental data and the transients predicted by the coupling code are also discussed.

Improvement of Condensation Model With the Presence of Non-Condensable Gas for Thermal-Hydraulic Analysis in Containment

Steam condensation plays a key role in prediction of the pressure behavior and hydrogen distribution in the containment during a hypothetical loss-of-coolant accident or a severe accident in a light water nuclear reactor. The objective of this study is to evaluate and improve the condensation model in GASFLOW code. Reynolds analogy coupled with wall function and Chilton-Colburn empirical analogy is used to model heat and mass transfer in GASFLOW, which has requirements for dimensional distance of the first cell near the wall and some deficiencies in description of heat and mass transfer process in the stagnant zone. Based on the evaluation of original condensation, the results shows good agreement with COPAIN experiment cases where the mass fraction of air ranges from 76.7 to 86.4%. However, with the changes in geometry of the facility and the presence of helium, the original model has a large deviation in the prediction of pressure, temperature and gas distribution compared with MISTRA ISP47 (OECD International Standard Problem No. 47) experiment data. This work proposes a modified condensation model which uses McAdams correlation and Schlichting correlation with a weight factor to calculate natural, forced, or mixed convection heat transfer coefficient, and adopts Chilton-Colburn empirical analogy to model mass transfer. The modified model has no requirement for the dimensionless distance near the wall in heat and mass transfer calculation and improves the prediction performance of heat transfer in stagnant zone. The prediction result of the modified model shows good agreements with MISTRA ISP47 problem, and the error of it compared with COPAIN experiment data is within 25% which is the same as that predicted by the original model.

Safety evaluation of nuclear power plant based on multiscale analyses

A loss-of-coolant accident of nuclear power plant is simulated using the system analysis code. The analytical condition is based on the international standard problem 26 conducted with the integral effect test facility. The experimental analysis is also performed, and the result of plant analysis is compared with the experiment and the experimental analysis. The discharge coefficient of critical flow model is determined so as to obtain the agreement of reactor pressure between the experiment and the experimental analysis, and is used for the plant analysis. The thermal-hydraulic phenomena in the experiment such as core heat up are simulated well by the two analyses, while some problems in the experiment for simulating plant accidents are made clear. Parametric sensitivity analyses of plant accidents are performed as an example of reliable safety evaluation using the validated plant model.

Analysis of Computer Modelling Results on Fuel Rods Strength and Condition at Reduced or Absent Cooling Caused by Accident

The paper describes the phenomenology of fuel rod behaviour in severe accident. As an example, an experiment is described resulting in severe damage of 19 fuel rod assembly of VVER type; it was carried out in the CORA facility in 1993 (Research Centre, Karlsruhe, Germany). Testing conditions and results of post-test investigations of fuel assembly are given. The fuel rod code RAPTA-SFD is briefly dealt with; the code was a participant in the International Standard Problem ISP-36. The basic results are presented acquired by computer modelling CORA-W2 experiment using RAPTA-SFD code. Among the presented experimentally acquired and calculated results, the scope of the data on stainless steel component behaviour is substantial. The tested CORA-W2 fuel assembly contained a significant quantity of steel components, viz., spacer grids, a guide thimble, and a cladding of an absorber element. It is to be borne in mind that the spacer grids and a guide thimble of the updated and upgraded fuel assembly of VVER-1000 are fabricated from Zr-alloy, hence, the relative quantitative characteristics of chemical interactions between materials and stainless steel (Cr-Ni alloy) will be much lower for the up-to-date upgraded fuel assembly under identical conditions.

The reactor power effect on the consequences pressure waves after LBLOCA in VVER-1000

In this paper, the influence of the reactor power behavior on the LBLOCA hydrodynamics at the initial moment of the process (0-1 second) is considered. The calculations are carried out for the first loop in case of the beginning of loading, end of loading and constant power. The parameters of a typical reactor plant V-320 (VVER-1000) were used for the calculation, particularly, the 3rd unit of the Kalinin NPP. All initial data for the calculation were obtained from the materials of the international standard problem Kalinin-3 [1]. The fuel composition and kinetics of the reactor core was considered for the case of the end of the cycle. For the calculations, the computational best estimate code ATHLET [2.3] was used, which is included in the software package AC2 and officially acquired by the national research Nuclear University MEPhI according to the license agreement with Gesellschaft fur Anlagen-und Reaktorsicherheit (GRS) gg GmbH, Germany [3]. Code ATHLET is certified in Russia for stationary and transient conditions at reactors with water coolant [4].

Development and validation of boron diffusion model in nuclear reactor core subchannel analysis

The diffusion characteristic of borated water in nuclear reactor core is important for the reactivity control during some possible reactivity transient scenarios. The non-uniform distributions of boron concentration increase the risk of reactor accidents and lead to some reactor safety problems. Hence it is necessary to develop an effective boron diffusion model into the reactor thermal-hydraulic analysis code. In this paper, a quasi-three dimensional boron tracking model, in which the turbulent diffusion effect was considered carefully, was proposed and successfully implemented in the subchannel analysis code. The reliable CFD method with k-omega turbulent model validated against International Standard Problem (ISP No. 43) experimental data was used to verify the feasibility of proposed boron tracking model. A typical Pressurized Water Reactor (PWR) 3 × 3 fuel bundle model was established and the borated water was injected into one specified channel and pure water into other channels. The distributions and variations of boron concentration along the flow direction were achieved and analyzed. Results indicate that the proposed boron tracking model could predict the boron diffusion process in the fuel rod bundles more accurately compared with the original model. This work provides an essential supplement for the nuclear reactor subchannel analysis method.

Coupling scheme of multicomponent sectional equations and Mason equations via transition rate matrix of hygroscopic growth applied to international standard problem No. 44

Most fission products are released into the containment as aerosol particles during severe accidents. In the conventional codes, two sets of equations are used to analyze the aerosol behaviors in the containment. The first set of equations are multicomponent sectional equations, solutions of which are given with the mass concentrations in each discretized section. The multicomponent sectional equations require very large number of sections to remove the numerical diffusion when hygroscopic growth occurs with other aerosol dynamics simultaneously. For efficient calculations, Mason Equations are used to analyze the hygroscopic growth. The solutions of the equations, on the other hand, are given with the radius of an individual particle. Therefore, it is essential to couple the two equations in rigorous manner for efficient analyses on the aerosol behaviors when hygroscopic growth and other aerosol dynamics occur simultaneously. In this paper, we propose a new formulation for the coupling scheme of multicomponent sectional equations and Mason equations. In the formulation, the two equations are coupled via a transition rate matrix obtained by interpolation of the mass distribution within a section for each component in multicomponent sectional equations and the solutions of Mason equations. In order to validate the formulation, it is applied to the simplified GDE to compare with the analytic solutions. The formulation is also applied to international standard problem No. 44 for the validation with experimental data. Computation time for the formulation is also compared with those of the conventional method. With negligible computation burden as additional, the formulation would analyze the aerosol behaviors efficiently, accurately, and stably when hygroscopic growth occurs with other aerosol dynamics simultaneously.

Spatial temperature distribution of fuel assembly pre-simulation for a new simple core degradation experiment

A new simple core degradation experiment will be conducted to investigate the distribution of mass and energy during a core degradation process, and the International Standard Problem (ISP) No.31 is chosen as a pre-simulation for the new experiment. The pre-simulation has been conducted using the widely accepted severe accident analysis software MELCOR. Firstly, the numerical analysis model and oxidation model are described, and all the input parameters are in accord with experiment conditions. Then, numerical results are validated by experimental measurements and SCDAP/RELAP5 results. Simulations results agree well with measured data. It is indicated that MELCOR has the capability of predicting the behaviors of fuel elements in reflood correctly. Finally, the visually spatial temperature distribution is obtained by TECPLOT, and the behaviors of molten fuel elements are directly reflected. The evolution of peak temperature in fuel rods during the experiment period can be visible. The peak temperature firstly appeared in outer heated rods of the fourth ring and later showed in a heated fuel rod of the second ring. The behaviors of molten fuel elements are visible in the figures of spatial temperature distribution, and it does help researchers to understand the migration behaviors of molten material accompanied by effective mitigation measures for a severe accident.

Pressure waves in the primary circuit of the VVER-1000 reactor at instant stop of the MCP

The initial stage of the accident caused by the instantaneous stop of the main circulation pump (MCP) of the first loop of the VVER-1000 reactor plant (RP) is considered. The appearance of pressure waves and their propagation in the equipment of the primary circuit of the RP is shown. For the calculation, the parameters of the typical RP V-320 (VVER-1000), in particular, the 3rd unit of the Kalinin NPP, are used. All initial data for the calculation is obtained from the international standard problem Kalinin-3 [1-2]. Calculations were carried out using the computational best estimated code ATHLET [3], developed the society for reactor safety (Gesellschaft fbr Anlagen-und Reaktorsicherheit-GRS), Germany and certified in Russia for use in calculations to justify the safety of reactors with water coolant [4]. In case of an accident, the failure of the load-off and reactor power controller (LRPC), automatic reactor power controller (APC), warning protection systems is assumed. Emergency protection reactor (PR-1) only works on the 2nd signal. The considered emergency situation is included in the list of mandatory reports for different types in the justification of VVER safety [5]. The obtained results can be used as boundary conditions for the analysis of equipment strength in this transition process.

A Review of the CFD Modeling Progress Triggered by ISP-47 on Containment Thermal Hydraulics

The Organisation for Economic Co-operation and Development (OECD)/Nuclear Energy Agency International Standard Problem 47 (ISP-47) was aimed at assessing the predictive capabilities of computational fluid dynamics (CFD) and lumped-parameter codes regarding hydrogen mixing under representative thermal-hydraulic conditions of a loss-of-coolant-accident. The benchmark consisted of two systematic steps. The first step was a fundamental model assessment based on quasi-steady-state separate-effects tests in the French TOSQAN facility (7 m3, IRSN, Saclay) and MISTRA facility (100 m3, CEA, Saclay) regarding steam condensation, buoyant turbulent flows, and mixed atmospheric conditions. The second step was based on a more realistic experimental transient in the multicompartmented German Thermal-hydraulics, Hydrogen, Aerosols and Iodine (THAI) facility (60 m3, Becker Technologies, Eschborn). At that time, the blind and open analysis revealed that CFD codes needed further improvement regarding modeling of turbulence in buoyant flows, steam condensation, temperature and species concentration, and stratification buildup as well as their dissolution. This result triggered a comprehensive experimental and analytical effort, e.g., within the German national THAI, the OECD-THAI, and the OECD-SETH-1 and OECD-SETH-2 projects. Now, 10 years later, this paper aims to benchmark the state-of-the-art containment CFD model, developed at Forschungszentrum Juelich and RWTH Aachen University, and to highlight the progress made and the remaining open issues.

3D-thermo-structural simulation of pressure tube–calandria tube behaviour under accident conditions in PHWR using ABAQUS

In-depth understanding of the thermo-structural behaviour of structure system and components (SSCs) of a nuclear power plant is necessary for both safety margin assessment, and for devising appropriate prevention and accident mitigation strategies. In a pressurised heavy water reactor (PHWR), the fuel in the form of fuel bundles resides inside pressure tube (PT). The pressure tube is surrounded by a co-axial calandria tube (CT). The heat generated by nuclear fission in fuel bundle is carried away by heavy water coolant, which flows inside pressure tube. The coolant is at high temperature and high pressure inside the pressure tube. The surrounding calandria tubes are at low temperature and low pressure under normal operating conditions. Under certain accident scenario, the fuel bundles cooling may be lost. This results in heating up of the fuel bundles which leads to increase in temperature of the pressure tube. The pressure tube starts deforming as its temperature increases. The deformed pressure tube may contact the calandria tube either by sagging or ballooning. This leads to increase in the calandria tube temperature. If the temperature on the outer surface of the calandria tube exceeds the critical heat flux, film boiling may occur on the surface of the calandria tube. If the area in dry-out is sufficiently large and the dry-out is prolonged, the pressure-tube/calandria-tube combination can continue to strain radially and may challenge fuel-channel integrity. The International Collaborative Standard Problem (ICSP) on Heavy Water Reactor (HWR) Moderator Sub-cooling Requirements is organized by the IAEA to facilitate the development and validation of computer codes for the analysis of fuel channel integrity. The objective was to assess the capability of safety analysis computer codes in predicting the associated phenomena namely; radiation heat transfer from fuel to the pressure tube (PT), from PT to calandria tube (CT) heat transfer, PT deformation or failure, CT to moderator heat transfer and CT deformation or failure. The current analysis is carried out as a part of international collaborative standard problem exercise. This work simulates the heat transfer phenomena along with resulting deformation of the pressure tube and calandria tubes in a finite element framework. The insights obtained from the detailed modelling of the structural aspects of the phenomena can be used to update the system safety analysis codes. In this paper, a coupled heat transfer and structural analysis of pressure tube -calandria tube assembly is carried out using the finite element code, ABAQUS. The heat transfer analysis implements radiation heat transfer from heater to PT and from PT to CT. Contact conductance between PT-CT is modeled based on contact pressure. Convection from outer surface of CT to water is also considered. Structural analysis included three cases elastic–creep, elasto-plastic and elasto-plastic with creep of the individual PT-CT assembly under thermal and mechanical loading. It is observed that the results matched well for the case where only elastic creep constitutive model was considered. The temperature variation of pressure tube and calandria tube along with resulting deformation is obtained. The simulation is able to capture and predict the progressive contact of PT with CT and the deformation of CT along with the PT using FEM model. It was observed during the experiment and captured in the simulation that the PT-CT assembly deformation lead to removal of heat to the moderator and channel integrity was maintained for the given moderator sub cooling margin.

Validation of STRCS code for calculation of fission-product transport in reactor coolant system during severe accidents

Transport and deposition of fission products within the primary coolant circuit during severe accident condition have significant effect on the amount of released radionuclides in the containment. In this study, a new code, STRCS, compatible with RELAP5/SCDAP code, has been developed that calculates the fission products’ transport through the piping system. The homogeneous and heterogeneous nucleation, aerosol agglomeration, aerosol and vapor deposition and aerosol resuspension are considered in this code. The developed code was validated through international standard problems STROM SR11 and PHEBUS FPT-1 experiments. The STRCS predictions are in good agreement with the ISPs published results, although the obtained results deviate from the PHEBUS FPT-1 experimental data.