ASME Conference Presenter Attendance Policy and Archival Proceedings

2018;():V007T00A001. doi:10.1115/ICONE26-NS7.

This online compilation of papers from the 2018 26th International Conference on Nuclear Engineering (ICONE26) 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presenter attendance policy, if a paper is not presented at the Conference by an author of the paper, the paper will not be published in the official archival Proceedings, which are registered with the Library of Congress and are submitted for abstracting and indexing. The paper also will not be published in The ASME Digital Collection and may not be cited as a published paper.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

Decontamination and Decommissioning, Radiation Protection, and Waste Management

2018;():V007T10A001. doi:10.1115/ICONE26-81017.

In order to maintain the U.S. domestic nuclear capability, its scientific technical leadership, and to keep our options open for closing the nuclear fuel cycle, the Department of Energy, Office of Nuclear Energy (DOE-NE) invests in various R&D programs to identify and resolve technical challenges related to the sustainability of the nuclear fuel cycle. Sustainable fuel cycles are those that improve uranium resource utilization, maximize energy generation, minimize waste generation, improve safety and limit proliferation risk. DOE-NE chartered a Study on the evaluation and screening of nuclear fuel cycle options, to provide information about the potential benefits and challenges of nuclear fuel cycle options and to identify a relatively small number of promising fuel cycle options with the potential for achieving substantial improvements compared to the current nuclear fuel cycle in the United States. The identification of these promising fuel cycles helps in focusing and strengthening the U.S. R&D investment needed to support the set of promising fuel cycle system options and nuclear material management approaches. DOE-NE is developing and evaluating advanced technologies for the immobilization of waste issued from aqueous and electrochemical recycling activities including off-gas treatment and advanced fuel fabrication. The long-term scope of waste form development and performance activities includes not only the development, demonstration, and technical maturation of advanced waste management concepts but also the development and parameterization of defensible models to predict the long-term performance of waste forms in geologic disposal. Along with the finding of the Evaluation and Screening Study will be presented the major research efforts that are underway for the development and demonstration of waste forms and processes including glass ceramic for high-level waste raffinate, alloy waste forms and glass ceramics composites for HLW from the electrochemical processing of fast reactor fuels, and high durability waste forms for radioiodine.

Commentary by Dr. Valentin Fuster
2018;():V007T10A002. doi:10.1115/ICONE26-81048.

Clearance is an effective method to control radioactive waste, but some assessment of types of pollutants may need to be done before clearance in some situation. In fact, there is large volume very low level radioactive materials in nuclear wastes in nuclear power plant, a reasonable method should be built based on both of potential clearance study and clearance method optimization. Some suggestions are presented about the work of radioactive waste disposal for its present condition and development in China.

Commentary by Dr. Valentin Fuster
2018;():V007T10A003. doi:10.1115/ICONE26-81147.

Many historic nuclear facilities have reached the end of their lifetimes and are now being decontaminated and dismantled. Our institute has been working on an innovative computing solution to simulate intervention scenarios in highly radioactive environments that humans cannot enter. Various treatment techniques can be compared and workers can be trained before operating on site. For the past 10 years, our institute has been developing a Virtual Reality tool that simulates all the key elements of a nuclear project, including the remote handling, and the accessibility of the site for humans, and the dose assessment associated. This tool allows all these parameters to be computed in a single interactive environment, allowing predefined scenarios to be verified and alternative solutions to be designed. This software is presented in the following paper; we here illustrate that it is particularly well-suited to dismantling projects. We also describe a first application of the tool to investigate radiological exposure scenarios. The advantages of this software include its user-friendliness, responsiveness, speed, and usefulness for the preparation of complex dismantling operations. The perspectives of this project include the simulation in Virtual Reality of human interventions.

Commentary by Dr. Valentin Fuster
2018;():V007T10A004. doi:10.1115/ICONE26-81160.

Aerosols as the main component of radioactive products in migration performance, which is an important factor that a unclear reactor accident present strong diffusion and affects the distributions of source and dose level in reactor containment, and they are therefore expected to be deposited in liquid phase such as in suspension pool and filtered containment venting device. In this paper, the deposition characteristics of micro-nano aerosols in rising bubble under pool scrubbing condition is studied with experiment, the aerosols size in the research range from 20 nm to 600 nm, and the bubble morphology mainly concern homogeneous bubbly flow. The results show that the deposition efficiency and mechanism of aerosol closely relate to gas flow rate, liquid level, particle size and bubbles size and so on. The aerosol deposition near 85nm is proved most difficult because of the convert of deposition mechanisms. In a high liquid level condition, micro-nano aerosol filtration efficiency is enhanced but gradually gradual. Under different gas flow rate, air bubble residence time and the bubble size distributions affect the filtration efficiency of aerosols.

Topics: Aerosols , Bubbles
Commentary by Dr. Valentin Fuster
2018;():V007T10A005. doi:10.1115/ICONE26-81228.

In 2002, the International Atomic Energy Agency (IAEA) mentioned the strengthening of nuclear knowledge, technology and application. This background has that there are aging of nuclear facility and nuclear power plant staffs. In addition, it would be difficult to succession a nuclear knowledge, technology, and skills. For example, undergraduate departments of nuclear energy and science are decreasing. The IAEA discussing those situations and pointed out the importance of a nuclear knowledge management. The nuclear knowledge management (NKM) is developing a database science as management on nuclear knowledge and information. In recent years, the IAEA has also advanced knowledge taxonomies on nuclear accidents as one of a nuclear knowledge management. In Japan, this achievements of nuclear knowledge taxonomy was using in the organization of information on accidents in Fukushima. A few studies are attempts to appropriately arrange and utilize huge amounts of information.

Even in nuclear facilities in Japan, it is pointed out a veteran or expert staff retirement and loss of knowledge and skill caused by this retirement. This problem is common issue in the world. Then, we created a prototype database system to utilize past documentation of knowledge and information. The database made from semantic web technology. The semantic web is a method of preparing a frame of categorized knowledge and linking information related to it. The target is a nuclear reactor of ATR Fugen that is decommissioning from 2008. Until now, cases of decommissioning completion are 17 cases in the world. One case of JPDR in Japan. It is not enough to understand a good method of decommissioning.

In general, the decommissioning project requires many information related to dismantling and decontamination. Particular, past information is important to know a past contamination situation and so on. This study focus on an access method for past data and information. However, we need to pay attention to other side of decommissioning project. Because of history of operating reactor has different tasks that are design, construction, operation and decommissioning. It is not appropriate to use the collected information as it is. For that reason, we will continue our research on the points pointed out above.

Commentary by Dr. Valentin Fuster
2018;():V007T10A006. doi:10.1115/ICONE26-81338.

Large amounts of highly contaminated water over 800,000 m3 accumulated in the reactor, turbine building and the trench in the facility were generated from the nuclear accident of Fukushima NPS (BWR) caused by the Great East Japan Earthquake. At present, the cold shutdown is completed stably by the circulating injection cooling system (SARRY, KURION) for the decontamination of radioactive nuclides such as 134Cs and 137Cs using zeolites and crystalline silicotitanate (CST). Further, the Advanced Liquid Processing System (ALPS) is under operation for the decontamination of 62 nuclides such as 90Sr, 129I and 60Co, etc. However, the adsorption behaviors of actinoids through the decontamination systems are complicated, and especially their adsorption properties for zeolites and CST, major inorganic adsorbents, are not yet clarified. In near future, the decontamination of actinoids leached from the crushed fuel debris will be an important subject. In this study, the practical adsorption properties of U(VI) for various inorganic adsorbents were evaluated under different solution conditions.

The adsorption properties (distribution behaviors and adsorption kinetics) were evaluated by batch adsorption method; 19 kinds of inorganic adsorbents including zeolites and CST (crystalline silicotitanate) were contacted with U(VI)) solutions. The conditions of 5 kinds of U(VI) solutions were as follows;

Solution 1: [U(VI)] = 50 ppm, initial pH = 0.5 ∼ 5.5

Solution 2: [U(VI)] = 50 ppm, [NaCl] = 0.1 M, initial pH = 4.0

Solution 3: [U(VI)] = 50 ppm, [CaCl2] = 0.1 M, initial pH = 4.0

Solution 4: [U(VI)] = 4.84 mM, [NaCl] = 0.1 M, initial pH = 3.18

Solution 5: [U(VI)] = 4.86 mM, 2,994 ppm boric acid/30% seawater, initial pH = 4.25

The uptake (%) and distribution coefficient (Kd. cm3/g) were estimated by counting the radioactivity using NaI(Tl) scintillation counter and liquid scintillation counter.

In the simple Solution 1, the Kd values for zeolites increased linearly with equilibrium pH up to pH 7. The Kd value for tin hydroxide had a maximum profile around pH 7 and a relatively large Kd value above 104 cm3/g was obtained. In the presence of NaCl and CaCl2 (Solution 2 and 3), relatively large Kd values above 102 cm3/g were obtained, other than mordenite and clinoptilolite, and the effect of [Ca2+] on U(VI) uptake was larger than that of [Na+]. In Solution 4 containing high concentration of U(VI), the uptake(%) was considerably lowered, while that for zeolite A, X and Y was estimated over 20%. Similar tendency was observed in Solution 5, and, in the case of granulated potassium titanate, yellow precipitate was observed on the surface due to the increase of equilibrium pH up to 5.25.

The adsorption behavior of U(VI) on inorganic adsorbents is mainly governed by three steps; ion exchange, surface precipitation of hydrolysis species and sedimentation depending on equilibrium pH, and hence it should be noted the change of U(VI) chemical species. These basic adsorption data are useful for the selection of inorganic adsorbents in the Fukushima NPS decontamination process.

Commentary by Dr. Valentin Fuster
2018;():V007T10A007. doi:10.1115/ICONE26-81341.

Operation and decommissioning of nuclear facilities will produce radioactive waste, and different radionuclides in the waste will bring different hazards to the public and the environment. The waste would be sorted more reasonably by distinguishing different radionuclides. Yet it is still very difficult to measure directly the pure Beta radioactive waste in situ, though in situ Gamma-analytical and Alpha-waste-barrel measurement techniques have become more sophisticated. The aim is to propose a scientific technique to sort the radioactive waste in situ. This study focused on the 90Sr-contaminated material in China Institute of Atomic Energy and optimized the design of the existing solid waste disposal facilities. A novel technique to measure the radioactive waste 90Sr-90Y online was proposed, trying to sort the radioactive waste as optimally as possible to realize further separation of exemption waste. Theoretically, the exemption waste can be further sorted, and it can guide the design of radioactive waste disposal system.

Commentary by Dr. Valentin Fuster
2018;():V007T10A008. doi:10.1115/ICONE26-81388.

The broad half-life range of Activated Corrosion Products (ACPs) results in major radiation exposure throughout reactor operation and shutdown. The movement of unpredicted activity hot spots in coolant loop can bring about huge financial and dosimetric impacts. The PWR operating experience depicts that activity released during reactor operation and shutdown cannot be estimated through a simple correlation. This paper seeks to analyze buildup and decay behavior of ACPs in primary coolant loop of AP-1000 under normal operation, power regulation and shutdown modes. The application of a well-tested mathematical model is extended in an in-house developed code CPA-AP1000, to simulate the behavior of dominant Corrosion Products (CPs), by programing in MATLAB. The MCNP code is used as a subroutine of the program to model the reactor core and execute energy dependent neutron flux calculations. It is observed that short-lived CPs (56Mn, 24Na) build up rapidly under normal operation mode and decay quickly after the reactor is shutdown. The long-lived CPs (59Fe, 60Co, 99Mo) have exhibited slow buildup under normal operating conditions and likewise sluggish decay after the shutdown. To analyze activity response during reactor control regime, operating power level is promptly decreased and in response specific activity of CPs also followed decreasing trend. It is noticed that activity of CPs drops slowly during reactor control regime in comparison to emergency scram. The results are helpful in estimating radiation exposure caused by ACPs during accessibility of the equipment in coolant loop, under normal operation, power regulation and shutdown modes. Moreover, current analyses provide baseline data for further investigations on ACPs in AP-1000, being a new reactor design.

Topics: Coolants , Corrosion
Commentary by Dr. Valentin Fuster
2018;():V007T10A009. doi:10.1115/ICONE26-81428.

Nuclear Decommissioning Projects and Programmes (NDPs) are characterized by significant risks, long schedules, and high costs that keep rising. Additionally, due to the NDP complexity and variety, it is extremely hard to understand which are the NDP characteristics that are associated with the NDP performance. This research takes the project management perspective, collects empirical information on NDPs and investigates this relationship between NDP characteristics and NDP performance, both through qualitative cross-comparison and quantitative statistical analysis based on the Fisher’s Exact Test (FET). In this paper, the results from the implementation of the FET applied on a pool of European NDPs are presented and discussed. Key takeaways are that some project characteristics present a stronger relationship with the project performance, while others do not show a significant association. Ultimately, qualitative and quantitative analyses complement each other to support the development of guidelines and improve the selection, planning and delivery of future NDPs.

Commentary by Dr. Valentin Fuster
2018;():V007T10A010. doi:10.1115/ICONE26-81531.

The general context of the article is related to the development of the laser cutting technique for the fuel debris retrieval on the damaged reactors of Fukushima Dai-ichi. IRSN and CEA are involved in a project, led by ONET, to bring relevant elements to analyze the risk occurred by the dispersion of aerosols emitted by the dismantling operations. Results regarding the aerosols source term characterization emitted during laser cutting of non-radioactive fuel debris simulants were acquired during experiments undertaken on the DELIA cutting laser platform from CEA. IRSN realized aerosol sampling, aerosol size distribution measurement and CFD calculation of aerosol transport and wall deposition. The evaluations performed will enable the Japanese teams responsible for extracting corium from the damaged reactors of Fukushima Dai-ichi to define the best strategies to implement containment, and ultimately to limit the dissemination of radionuclides in the environment.

Commentary by Dr. Valentin Fuster
2018;():V007T10A011. doi:10.1115/ICONE26-81552.

Radiation safety is an important part of safety assessment of spent fuel dry storage technology. This paper describes the radiation protection design of PWR spent fuel dry storage facility for radiation safety completed by China General Nuclear Power Corporation. Considering the special site conditions, Monte Carlo method is used to complete the precise calculation of the three-dimensional radiation dose field in the spent fuel storage building. Through the spent fuel storage module and the storage building with shielding function, radiation shielding design is completed to meet China’s regulatory requirements, which ensures radiation safety for workers and the public during the transport and storage of spent fuel. It will provide a reference for construction of spent fuel dry storage facility of CPR1000 and HPR1000.

Commentary by Dr. Valentin Fuster
2018;():V007T10A012. doi:10.1115/ICONE26-81609.

An inverse source estimation method is proposed to reconstruct emission rates of multi-radionuclides using local gamma dose rate measurements under the data assimilation framework. It involves the Proper Orthogonal Decomposition (POD)-based ensemble four-dimensional variational data assimilation (PODEn4DVar) algorithm and a transfer coefficient matrix (TCM) created using FLEXPART, a Lagrangian atmospheric dispersion model. PODEn4DVar is a hybrid data assimilation method that exploits the strengths of both the ensemble Kalman filter (EnKF) and the 4DVar assimilation method. With an explicit expression of control (state) variables in the cost functional, the data assimilation process is substantially simplified than traditional 4D variational method. By setting a unit emission rate and running the ATDM model (FLEXPART in this article) driven by meteorological fields forecasted with WRF, we get the transfer coefficient matrix with the progression of nuclear accident. TCM not only acts as observation operator in PODEn4DVar, but also eliminates the control run in traditional data assimilation framework. The method is tested by twin experiments with ratios of nuclides assumed to be known. With pseudo observations based on Fukushima Daiichi nuclear power plant (FDNPP) accident, most of the emission rates were estimated accurately, except under conditions when wind blew off land toward the sea and at extremely slow wind speeds near the FDNPP. Because of the long duration of accident and variability of meteorological fields, measurements from land only in local area is unable to offer enough information to support emergency response. With abundant measurements of gamma dose rate, emission rates can be reconstructed sequentially with the progression of nuclear accident. Therefore, the proposed method has the potential to be applied to nuclear emergency response after improvement.

Commentary by Dr. Valentin Fuster
2018;():V007T10A013. doi:10.1115/ICONE26-81616.

Before the design of the nuclear facility, it is necessary to estimate the dose caused by radioactive material released into the environment during the course of the serious accidents. Semi-infinite hemisphere geometric model is established to estimate the personal external exposure dose outdoors in which the distribution of the nuclides is assumed to be uniform. The exposed staffs are in a limited cubic space when using this model to evaluate the controllability of the main control room. Thus, the volume correction factor is needed to correct the dose, whose traditional expression is f = 352/pow(V, 0.338). The formula cannot satisfy the requirement of higher accuracy due to the neglect of the influence of the shape of geometric model and γ-rays energy. Usually the actual control room is a cube and the γ-rays energies emitted from various nuclides are different.

In order to calculate the accurate volume correction factor of main control room under different geometric conditions, a finite cubic geometric model is established in this paper. The length and width of the model are between 6m and 50 m, the height is between 4m and 6m, and γ-ray energy respectively are 0.05, 0.2, 0.733, 1.2 and 3 MeV, respectively. The effective volume values for different conditions are calculated by the Monte-Carlo program, and 318 groups of results are obtained. The calculated volume dose rate of 360m × 360m × 255m (assuming semi-infinite) cube at 733keV γ-rays energy is taken as a criterion, whose ratio of the other calculation results is the new volume correction factor value. By comparing two volume correction factors, the relative discrepancies are within 3 folds, proving that the calculation result is reasonable and feasible. The new volume correction factor varies with γ-rays energy and the shape of the geometric model.

A neural network model corresponding to the volume correction factor is developed to apply to more cases. 80% of the results are randomly selected as the training set of neural network. The remaining 20% of the result as the test set of the cross-test is to predict the results of the trained neural network, whose relative errors are less than 5%. The neural network model can obtain the volume correction factor under different geometry and γ-ray energy conditions. Finally, a volume correction factor library is established, which can provide a powerful reference to obtain the volume correction factor of the limited space model such as the main control room.

Commentary by Dr. Valentin Fuster
2018;():V007T10A014. doi:10.1115/ICONE26-81754.

In this study, a Monte Carlo model has been developed for a Cerenkov-based fiber-optic gamma-ray sensor (CFOGRS) using the GEANT4 simulation toolkit. The detection material for gamma rays in CFOGRS is the transparent silica core of the optical fiber, which is also used for optical signal propagation. The model implemented with the GEANT4 includes the transport process of gamma rays, as well as the physical processes of Compton scattering, the Cerenkov effect, and optical photon propagation within the optical fiber. The model also simulated the applicability of the CFOGRS in a radiation environment by using the Monte Carlo code of GEANT4.

Commentary by Dr. Valentin Fuster
2018;():V007T10A015. doi:10.1115/ICONE26-81853.

Laser cleaning study was performed on prepared samples using a nanosecond pulsed ytterbium fiber laser. To prepare samples, AISI 304L stainless steel samples were oxidized and implemented with non-radioactive contaminants in a controlled manner. In order to validate the cleaning process for metallic equipment polluted in nuclear installations, two types of contamination with europium (Eu) and with cobalt (Co) were studied. Eu was used as a simulator-product resulting from uranium fission, while Co — as an activation-product of nickel, which is a composing element of a primary coolant system of a reactor. The oxide layers have suffered laser scanning which was followed by the furnace treatment to obtain thicknesses in the range of 100 nm to 1 μm depending on the oxidation parameters [1] with a mean weight percentage of 1% of Eu and 1 % of Co in the volume of the oxide layer. Glow Discharge Optical Emission (GD-OES) and Mass Spectrometry (GD-MS) analyses have been performed to assess the efficiency of the cleaning treatment and to follow the distribution of residual contamination with a detection limit of 0.1mg/kg of Eu and Co. Decontamination rates up to 95.5 % were obtained.

One of the identified reasons for this limitation is that the radionuclides are trapped in surface defects like micro cracks [2, 3]. Therefore, cleaning treatments have been applied on surface defects with controlled geometry and a micrometric aperture obtained by laser engraving and juxtaposition of polished sheets of AISI 304L stainless steel. The goal of this study is surface decontamination without either welding or inducing penetration of contamination into the cracks. GD-MS analysis and Scanning Electron Microscopy (SEM) were performed to analyze the efficiency of the treatment and the diffusion of contaminants in this complex geometry.

Topics: Lasers
Commentary by Dr. Valentin Fuster
2018;():V007T10A016. doi:10.1115/ICONE26-81941.

Control room habitability (CRH) shall be maintained to provide adequate protection for control room operators, such that they can remain in the control room envelope (CRE) safely for an extended period and thus control the nuclear facility during normal and accident conditions. A critical objective of CRH systems is to limit operator doses and/or exposure to toxic gases. The CRH systems does this by the combination of the intake of filtered air, isolation of outside air, recirculation systems and etc.

Among the parameters determining radioactivity in a control room (in proportion to radiation doses of operators), intake flowrate of filtered air is an important one. For different types of accident source terms, the evolution of operator doses in a control room versus intake flowrate were analyzed in this paper. It turns out that the increase of intake flowrate results in larger operator doses when inert radioactive gases are the dominant radioactive substances. On the contrary, increasing intake flowrate does good to lower the irradiation level of control room operators when radioactive aerosols dominate the source terms. The rationality behind this fact was interpreted in detail in this paper, with special attention paid to the unfiltered in-leakage rate. It can be inferred that an optimal intake flowrate probably exists leading to the minimum operator dose under an actual accident condition.

This paper then performed a calculation analysis based on design parameters and source terms of design basis accident of LOCA (a large break loss of coolant accident) accident. The evolution of operator dose was found to be a U-curve versus increasing intake flowrate, which proved the existence of the abovementioned optimal intake flowrate of filtered air for CRH systems. Furthermore, the sensitivity analysis of intake flowrate was carried out to study the effects of unfiltered in-leakage rate and filtered recirculation.

This study indicates that intake flowrate of filtered air can significantly influence the CRH. For different accidents, the intake flowrate should be properly modified rather than set as a fixed value. To optimize the radiological habitability of control rooms, the effects of unfiltered in-leakage must be taken into consideration. Besides, filtered recirculation is an effective way to control radiation exposure caused by iodine and radioactive aerosols.

Topics: Control rooms
Commentary by Dr. Valentin Fuster
2018;():V007T10A017. doi:10.1115/ICONE26-82026.

There is a large amount of tritium in tail gas emitting from the molten salt reactor. The content of tritium content in tail gas must be measured and controlled according to the national standard. Nowadays, the ionization chamber and the liquid scintillation counter are mainly used for tritium monitoring. The proportional counter required pure sampling gas, and the liquid scintillation cannot measure the radioactivity online. Only the ionization chamber could measure the activities of off-gas from molten salt reactor. In order to reduce tritium contamination in the ionization chamber, a wire type high pressure ionization chamber was used in measuring tail gas from molten salt reactor. In order to improve the ion collection efficiency of the wire type electrode ionization chamber, it is necessary to optimize the number of the electrodes. In this study, the electric field of ionization chamber with different numbers of electrodes, such as 3, 6, 12, 18, 24, 30, 36, 42, 48, were simulated by COMSOL software. The equipotential line of electric field near the collector is round. The equipotential line of electric field near the high voltage wires present as undulate. From the results, the optimum number of wires was 36. The difference of electric field between two types chambers were less than 5%. After that, collection efficiency was simulated as well. Increasing the number of the high voltage electrode at top and bottom of effective area could not improve the electron collection efficiency. Making the ionization chamber shell as an insulator can effectively improve the electrons collection efficiency.

Commentary by Dr. Valentin Fuster
2018;():V007T10A018. doi:10.1115/ICONE26-82039.

Since August 2015 a new classification of radioactive waste was issued by Italian Ministry of Economic Development, in order to adapt Italian historical classification to European standards. This new classification provides 6 categories, from exempt to high level waste, and it is based on the waste final destination: from free release to final disposal or interim storage (high level waste and intermediate level waste with α-content higher than 400 Bq/g) [1].

Nucleco is a State owned Company acting as Waste Management Organization for radioactive waste coming from hospitals, industries and research and development activities not related to electricity production by nuclear plants. Nucleco collects, safely manages and temporarily stores waste that will be sent to the National Repository (site definition phase is still ongoing), while for the Short Lived Radionuclides and Very Low Activity waste Nucleco performs all necessary operations to be compliant with the conditions of release prescribed by the Italian Control Authority.

Short Lived Radionuclides are those whose half-life is shorter than 100 days or reach the condition of non-radiological relevance in 5 years: they are mainly produced by bio-medical applications of radioactive materials. Very Low Activity waste are characterized by activity concentrations lower than 100 Bq/g (of which less than 10 Bq/g of α-emitting radionuclides) and reach the condition of non-radiological relevance in 10 years: these waste usually came from research institutions and industrial activities.

This work presents the authorized operating procedures, the radiological measurements criteria and the technical know-how put in place by Nucleco to fulfil the provisions of Italian regulations for unconditioned release of radioactive waste.

A case study of ISO 20’ containers is discussed in the current paper.

Main emphasis will be addressed to:

• gathering of historical information about the state of the material to be released and definition of the reference radiological spectrum;

• sampling procedures to ensure representativeness of the samples from homogeneous waste batch to be released and then subjected to radiological characterization;

• characterization phase consisting of the integration of several state-of-art techniques aiming to collect the most complete set of radiological data;

• data processing protocols needed for the calculation of the activity concentrations for each radionuclide of the reference spectrum (or other radionuclides eventually detected);

• evaluation of the main sources of uncertainty affecting the results;

• comparison of the activity concentration (including the uncertainty) of each radionuclide with the corresponding authorized concentration limits.

Commentary by Dr. Valentin Fuster
2018;():V007T10A019. doi:10.1115/ICONE26-82056.

Radiation dose and personnel protection are among the safety goals of geological disposal of high-level radioactive waste. The calculation of the dose field on the surface of the packaging container is of great significance for the research on the dose constraint value of the repository. This paper built model consulting the Sweden KBS-3 canister, the temporal and spatial distribution of the dose rate on canister surface was calculated by Monte Carlo method, the temporal and spatial distribution of radiation dose rate of the tunnel was obtained. The research results showed that the photon dose rate on canister surface was greater than the neutron dose rate by 4 to 6 orders of magnitude, and the dose value of repository tunnel within 100 thousand years was lower than the ICRP recommended dose limit value (0.3 mSv/a) by 5 orders of magnitude.

Commentary by Dr. Valentin Fuster
2018;():V007T10A020. doi:10.1115/ICONE26-82258.

The NSRR (Nuclear Safety Research Reactor) is a research reactor of TRIGA-ACPR (Annular Core Pulse Reactor) type, located in the Nuclear Science Research Institute (NSRI). The NSRR facility has been utilized for fuel irradiation experiments to study the behaviors of nuclear fuels under reactivity initiated accident (RIA) conditions.

Under the new regulation standards after the Fukushima Daiichi accident, the research reactors are being regulated according to the risk of the facility. Graded approach is introduced in the regulation. In order to apply the graded approach, the radiation effects of residents living around the NSRI under the external hazards were evaluated and the level of the risk of the NSRR facility was investigating. This report is summarized for the result of the evaluation in case the safety functions were lost by the tornado, earthquake and following tsunami.

As the result, the risk is confirmed to be low, since the effective dose of the residents has been below 5 mSv per event due to the loss of the safety functions by the tornado, earthquake and following tsunami.

Commentary by Dr. Valentin Fuster
2018;():V007T10A021. doi:10.1115/ICONE26-82291.

In this paper, a kind of transport container we designed was introduced. This container was designed to transport nuclear fuel pellets in different enrichment of U-235. The weight of this package is about 400kg, including the contents. One protect shell and two sealed border were designed in this container, which can ensure the contents were intact and the package has no criticality risk after transport accidents. During the design and safety analysis process, finite element analysis methods were used to improve the structure and analyze the safety performances of the container. In addition, we will test the safety performances of this container through a series of experiments in the future, including 9m drop (or crush), puncture, fire, water immersion and so on. Now, the calculate results show that this container was fit to the safety requirements in the transport accidents.

Topics: Containers , Safety , Design
Commentary by Dr. Valentin Fuster
2018;():V007T10A022. doi:10.1115/ICONE26-82468.

A salt waste generated from the pyroprocess contains residual actinides and needs to be purified for recycling of the salt and waste conditioning. A co-reduction process could be considered for removal of residual actinides from the salt waste, which contains lanthanides and residual actinides. In the study, specifically, an effect of Bi(III) ion on the electrochemical reaction of Tb(III) ion was investigated in the molten LiCl-KCl eutectic with BiCl3 and TbCl3 at 773 K using electrochemical techniques of cyclic voltammetry, square wave voltammetry and open circuit chronopotentiometry. Tb(III) has a single redox couple without Bi(III). However, the cyclic voltammograms obtained at tungsten electrode in LiCl-KCl-BiCl3-TbCl3 showed four redox couples. The square wave voltammogram in same condition also showed five reduction peaks. Cyclic voltammogram and square wave voltammogram was resolved to find the accurate peaks for redox reaction. Each peak indicates the formation of Tb-Bi intermetallic compound except Tb(III) reduction peak. From the phase diagram of Tb-Bi, it is inferred that each peak corresponds to TbBi2, TbBi, Tb4Bi3, and Tb5Bi3. The open circuit chronopotentiometry was conducted to estimate Gibbs free energy of formation of Tb-Bi intermetallic compound.

The experimental results obtained from three kind of the electrochemical techniques showed that Tb-Bi intermetallic compounds were electrochemically formed under potential of Tb(III) reduction potential by co-reduction of Bi(III) and Tb(III). These results indicate that underpotential deposition by co-reduction could be used for Tb(III) removal from the salt waste with Bi(III).

Commentary by Dr. Valentin Fuster
2018;():V007T10A023. doi:10.1115/ICONE26-82572.

Decommissioning cost including radioactive waste management for 1100 MWe nuclear power plant (BWR) was analyzed comparing multiple scenarios. The total cost of decommissioning nuclear power plant was first estimated including the radioactive waste management cost for the standard Japanese decommissioning case with 30 years of the project duration including approximately 20 years in safe storage. It showed that the cost relating to waste management accounts for more than half of the total cost. Focusing on the radioactive waste management cost, the duration of safe storage was varied as a parameter. The timing of waste disposal was a key parameter determining the waste management cost due to the decay of radioactive nuclides resulting in the decrease in the total volume of the radioactive waste, and the change in the ratio of the waste volume in the three radioactive waste categories (intermediate-level, low-level, and extremely low-level). The total cost showed the minimum value at around 60 years of the project duration balancing the waste management cost and period dependent cost for safe storage.

Commentary by Dr. Valentin Fuster
2018;():V007T10A024. doi:10.1115/ICONE26-82581.

The high-temperature gas-cooled reactor pebble-module (HTR-PM), which is under constructed in Shidao, Shandong Province of China, adopts the spherical graphite coated components as fuel elements. During the operation of the reactor, the spherical fuel elements will be reloaded into the core periodically through the refueling pipelines, which will cause radioactive irradiation to personnel and facilities in several corridors and rooms. The three-dimensional geometry model was built according to the real structure and the radiation shielding design was verified and optimized by the QAD-CGA program. In the calculation, the average gamma intensities of the spherical fuel elements of 15 different cycle times were adopted as the source term. From the results, it can be concluded that: 1) For the 20-meter corridor, the dose rate will be less than 1E−6mSv/h at 30cm outside the concrete wall of 0.6m thickness by single fuel elements; 2) For the 0-meter corridor, an extra 32.0 cm steel shield should be added to satisfy the personnel dose rate limit of 0.03mSv/h; 3) For semiconductor electrical equipment closed to the pipelines, the required steel shield thickness is 12.0 cm under the accumulative dose limit of 50Gy; 4) For non-radiation resistant cables, the required steel shield thickness is 4.5 cm under the accumulative dose limit of 1000Gy.

Commentary by Dr. Valentin Fuster
2018;():V007T10A025. doi:10.1115/ICONE26-82617.

Compacted bentonite materials are often considered as a buffer material in the geological radioactive waste disposal. This bentonite is expected to fill up the space between the waste and the surrounding ground by swelling. Therefore, understanding the surrounding ground, i.e., groundwater behavior in bentonite, as a buffer material, is essential in order to evaluate the bentonite buffer performance and guarantee long-term safety. The monitoring system of the water saturation level in compacted bentonite is required because water content in buffer material may influence its elastic properties. In this study, the correlation between water content and elasticity in unsaturated compressed bentonite was experimentally evaluated. The evaluation was done by measuring the sound velocity of both longitudinal wave and transverse wave. As a result, it can be confirmed that ultrasonic velocities could evaluate a degree of saturation and bulk modulus of compacted bentonite.

Commentary by Dr. Valentin Fuster

Mitigation Strategies for Beyond Design Basis Events

2018;():V007T11A001. doi:10.1115/ICONE26-81171.

The spray cooling and heat removal efficiency is one of the important aspect of nuclear thermalhydraulics and safety, especially for passive containment cooling after severe accidents. In order to design and optimize these systems effectively, computer modelling of the underlying mechanism of the liquid drop interaction with the hot solid surface would be necessary. Therefore, completeness, accuracy and reliability of the models that are being used in such sensitive areas are vital to the society and environment. Furthermore, the current powerful computer resources need to be fully exploited, so that the precision and the accuracy of the obtained computational results would be further enhanced. Nowadays, Volume-Of-Fluid (VOF) method is widely used in simulating the droplet dynamics, however these models provide estimations that are different in certain extents compare to the experimental results. In present work, we have used the level-set method to study the droplet dynamics and heat removal when the water droplet impact on the surface with different morphologies. The developed model which is based on the finite element method (FEM) has been benchmarked with previously performed experiments regarding the droplet bouncing on a flat hydrophobic surface; these estimations were in a good agreement with the previously published results. Moreover, hot solid surfaces with presence of micro-pillar has been considered to perform sensitivity study for different sizes of the micro-pillars and water droplets. In addition, it has been found that the heat transfer and droplet dynamic behavior would significantly vary in scenarios when the micro-pillars are presents in compare to a flat solid surface; it is observed that a better droplet spreading can be obtained with optimal size of micro-pillars that are present underneath of the droplet axial trajectory. The present study and the model would add valuable information to the field of heat transfer in aspect of spray cooling by investigating the feasibility of using the level-set method for a better estimation of fluid and heat transfer related results.

Commentary by Dr. Valentin Fuster
2018;():V007T11A002. doi:10.1115/ICONE26-81397.

Protection against large commercial aircraft crash is a new design requirement for international advanced nuclear power plants, and gradually becomes one of important advance characteristics. The study follows the principle of balance of safety, economy, and mature achievable engineering technology, conducted by design extension condition and realistic analysis inapplicable to single failure criterion, adopts Anti-Plane Crash (APC) shell protection + redundant features physical isolation to fulfill the safety acceptance criteria, and results in the HPR1000 design scheme withdraw large commercial aircraft crash. Beside the overall analysis on the impact effects of global structure, local structure, vibration and fuel, the study optimizes the design of each outside openings beyond personnel access size in APC shell to limit the mechanical damage of equipment due to local penetration and induced internal fire or explosion by leaking or pouring oil, by the design or construction scheme such as removable walls, protection covers, anti-aircraft gate, manway relocation, concrete second-pouring, and etc. In addition, the study also analyzes the hazard consequence of impact without APC shell protection, especially the influence of the safety structures or equipment damage of other safety buildings (e.g. nuclear auxiliary building) and induced fire spreading or flooding, to ensure the acceptance criteria of integrity of containment and spent fuel pool, or the cooling ability of reactor core and spent fuel pool. This outcome of the study enhances the safety and capability against external extreme hazards of HPR1000, and strengthens the public confidence on nuclear safety and anti-terrorism.

Topics: Aircraft
Commentary by Dr. Valentin Fuster
2018;():V007T11A003. doi:10.1115/ICONE26-81465.

The process of the ex-vessel molten core cooling in a pre-flooded reactor cavity during a severe accident of a light water reactor includes complicated phenomena such as melt jet breakup, debris bed formation and cooling and the molten core-concrete interaction. The melt coolability and its impact on the containment consequences are dependent on the interactions among those phenomena. A simplified parametric model, COOLAP-II, covering the melt jet breakup, debris bed formation and cooling was developed for the synthetic assessment of the ex-vessel melt coolability. The model was validated on the melt breakup and initial cooling by comparison with a full-model for fuel-coolant interactions, JASMINE.

Topics: Cooling , Vessels
Commentary by Dr. Valentin Fuster
2018;():V007T11A004. doi:10.1115/ICONE26-81633.

The primary purpose of this study is to assess the release of source terms into the environment for representative spent fuel pool (SFP) severe accident scenarios in a reference pressurized water reactor (PWR). For this, two typical accident scenarios (loss-of-cooling and loss-of-pool-inventory accidents) and two different reactor operating modes (normal and refueling modes) are considered in the analysis. The secondary purpose of this study is to assess the impact of an emergency makeup water injection strategy, which is one of representative SFP severe accident mitigation (SAM) strategies being employed after the Fukushima accident, upon the release of the radiological source terms. A total of 16 cases, consisting of four base cases and three injection cases for each base case were simulated using the MELCOR1.8.6 SFP version. The, analysis results are given in terms of (a) the key thermal-hydraulic behaviors during an accident progression and (b) releases of radiological fission products (such as Cesium and Iodine) into the environment. In terms of a release of Cesium and Iodine into the environment, the present study show that the two cases subject to a loss of pool inventory (i.e., LOPI-N-03 and LOPI-R-00) lead to the worst results with the respective release fractions of 77.5% and 59.4%.

Commentary by Dr. Valentin Fuster
2018;():V007T11A005. doi:10.1115/ICONE26-81736.

Due to the inherent safety characteristics of passive core injecting system, it is one of the most important mitigation strategies under loss of coolant accident (LOCA). However, flow instability during passive gravity-driven reflooding may occur, which can obstruct core reflooding process and reduce the heat removal rate from the reactor core. Therefore, exploring the characteristic and influencing mechanisms of flow instability during gravity-driven reflooding is necessary to analyze the effectiveness of passive core injecting system and mitigate the consequences of LOCA.

In this work, small scale experiments are performed to investigate the flow instability during gravity-driven reflooding when the open ratio of steam outlet are in the ranges of 3.33–50%, gravity driving head are 16–20kPa and system pressure are 0.1–0.3MPa. The test facility consists of a cooling water storage tank, a test section and a condensate tank. The results show that flow instability could occur under specific conditions and have a strong regularity, which is divided into three stages: cooling water initial injection, cooling water expulsion and cooling water re-injection. In addition, the effects of open ratio of steam outlet, gravity driving head and system pressure on flow instability phenomenon are investigated. Increasing the open ratio of steam outlet accelerates the discharge rate of steam and prevents the accumulation of steam at the inner cylinder, which can contribute to cooling the heated rods and restrict the flow instability. And under low gravity driving head of 16kPa and low system pressure of 0.1MPa the temperature and pressure acutely change and flow instability is more likely to occur.

Commentary by Dr. Valentin Fuster
2018;():V007T11A006. doi:10.1115/ICONE26-81788.

Hydrogen can be generated by the oxidation of fuel claddings with steam at a high temperature in a nuclear reactor. When the generated hydrogen is discharged from the reactor vessel into the containment, the integrity of the containment can be challenged by the high pressure and shock wave resulting from a hydrogen explosion similar to those during the Fukushima accident and the TMI accident. Thus, an investigation of the hydrogen behavior is required in order to assess the threat of a hydrogen explosion. The objective of this research is to conduct a preliminary 3D computational fluid dynamics (CFD) analysis to examine the hydrogen behavior in the containment. To achieve this objective, GASFLOW-MPI, a computational fluid dynamics tool, was selected given that it is specialized for analysis of hydrogen behavior in nuclear power plants. A large-break loss-of-coolant accident (LBLOCA) scenario was selected for the assessment. A guillotine break was assumed at a cold leg for the LBLOCA. The boundary condition, specifically the guillotine break, and the initial conditions were set based on calculation results from MAAP 5.03. Mitigation measures, including a spray system, passive autocatalytic recombiners (PARs), and hydrogen ignitors installed in the containment of a NPP were considered in the assessment. In this case, the mitigation measure models inserted by GASFLOW-MPI were used. Specifically, the Korean PAR was modeled by GASFLOW-MPI for the PAR installed in the OPR1000. The results of the preliminary analysis show that GASFLOW-MPI has the capacity to assess hydrogen behavior. The operation of the spray system increased the hydrogen volume fraction and halved the steam volume fraction compared to when it did not operate due to condensation.

Commentary by Dr. Valentin Fuster
2018;():V007T11A007. doi:10.1115/ICONE26-81896.

A passive endothermic reaction cooling system (PERCS) is proposed to provide reactor core cooling during a station blackout (SBO). During a SBO, a PWR in which PERCS has been installed has a peak reactor core outlet temperature remains below 640 K (692.3°F) for 30 days, which is well below the nominal accident core outlet temperature during a SBO. During a LOCA, LOFA, and LOHSA, installation of a PERCS has no significant impact on safety performance. It should be noted that the PERCS will represent a minimal heat source (unless the PERCS is very large) during DBAs as emergency systems lower the coolant temperature below the PERCS temperature.

A typical PWR with an installed PERCS is modeled using RELAP5-3D. The results of the model demonstrate the high level of passive safety afforded by the PERCS which contributes to the mitigation of SBO consequences without adversely affecting nuclear plant safety during a LOCA, LOHSA, or LOFA. Future work in validating the PERCS as a method of passive safety for existing light water reactors is underway, including the refining the physical design, determining better kinetic and thermodynamic properties for MgCO3, updating the PERCS model, and using a more robust PWR plant model.

Commentary by Dr. Valentin Fuster
2018;():V007T11A008. doi:10.1115/ICONE26-81899.

Hydrogen combustion or detonation happened in the containment within the process of the small reactor severe accident may threaten the integrity of the containment. In this paper, based on systemic design of the Small Modular Reactor (SMR) surrounded by the steel containment, an innovatory combustible gas control strategy which using the passive containment cooling system (PCCS) and passive autocatalytic recombiners (PARs) is made to control the hydrogen risk in the small steel containment. A severe accident hydrogen risk analysis model is built by the integrative severe accident analysis program MELCOR, the validity of the strategy is analyzed at a typical severe accident. With this understanding, a three-dimensional computed fluid dynamics hydrogen behavior analysis model of the small steel containment is established by GASFLOW code, and the gas distribution non-uniformity in the containment is analyzed. The result shows that the steam condensation process in the containment could be slowed down by controlling the action of PCCS, and the steam concentration in the containment could be in the range of high level, while the oxygen concentration could be in the range of low level. If the PARs were added, the PARs could consume the hydrogen and oxygen in the containment sustainedly. The containment atmosphere could be in an inerted condition during the accident process, even though the hydrogen concentration in the containment is high. The gas distribution non-uniformity analysis result shows that oxygen concentration was low in the extent of high hydrogen concentration and high steam concentration, the steam, oxygen and hydrogen distribution non-uniformity would not affect the inerted atmosphere of containment.

Commentary by Dr. Valentin Fuster
2018;():V007T11A009. doi:10.1115/ICONE26-82028.

The dry-out heat flux (DHF) for the copper sphere beds was measured using the hydrogen generated at a high potential in the copper electroplating system of mass transfer. The uniform self-heating condition was simulated by the experimental methodology. The experimental apparatus consisted of a power supply, a data acquisition system, a polycarbonate cylinder containing the copper cathode beds and copper anode submerged in an aqueous solution of H2SO4. This study reported the experimental results for the bottom flooding condition with the 6 mm copper beds. The DHF data was obtained for the bed height of 20, 40 and 60 mm. The tendency of DHF on the bed height is similar to that of existing heat transfer studies. However, the absolute DHF value was quite different. We inspected the potential of the non-heating experimental method and reviewed further studies for reducing the discrepancy.

Topics: Heating , Heat flux
Commentary by Dr. Valentin Fuster
2018;():V007T11A010. doi:10.1115/ICONE26-82029.

The explosivity of corium is evaluated from the fuel coolant interaction experimental data produced in the TROI facility.

From 62 experimental datasets used to simulate the fuel coolant interaction under a partial flooded condition of the reactor, the explosivity when observing the dynamic pressure and/or force measured in experiments is qualitatively evaluated depending on the coolant depth, the shape of the interaction chamber, the free fall height of the melt jet, and the water temperature effect. For 12 of the experimental datasets produced under the OECD/SERENA Project, the explositivity is quantitavely evaluated using the conversion ratio. The conversion ratio is also evaluated for three experimental datasets to simulate the fuel coolant interaction under reactor flooded conditions.

The corium system showed a relatively low explosivity compared to the ZrO2 or Al2O3 system. It also turned out that the flooding condition of the reactor cavity does not affect the change in explosivity.

Topics: Fuels , Coolants
Commentary by Dr. Valentin Fuster
2018;():V007T11A011. doi:10.1115/ICONE26-82074.

Steam generator tube rupture (SGTR) accident is one of important accident that has high probability of resulting in severe accidents. As a bypass scenario, fission product can be directly released to the environment during the SGTR accident. Thus, the severe accident by SGTR should be carefully managed by severe accident management guidance (SAMG). In Korea, SAMG for optimized power reactor 1000 (OPR1000) has been developed in 1999 and used to mitigate the severe accident of OPR1000 with seven mitigation strategies. Among the mitigation strategies, ‘Depressurization of reactor coolant system (RCS)’ is one of the most powerful strategies to reduce direct release of the fission product. To reduce the RCS pressure, indirect depressurization using steam generator is generally recommended. However, depending on the RCS condition, the indirect depressurization can be ineffective to reduce the RCS pressure. In this case, direct depressurization using pilot operated relief valve (PORV) should be performed as a second plan. From this point of view, sensitivity study of RCS depressurization was performed to investigate priority of depressurization in this study. The severe accident scenario initiated by SGTR accident was selected from probabilistic safety assessment (PSA) level 1 report and simulated using MELCOR 2.1. For the mitigation strategy, various timing of depressurization, the number of opening valves and flow rate of feed water were applied to simulate the possible depressurization strategies during the severe accident. The MELCOR code simulation shows that if depressurization was performed at 30 minutes after SAMG entrance, the direct depressurization was more efficient to reduce the RCS pressure and the fission product release. Therefore, it was recommended to use direct depressurization rather than indirect depressurization in certain time. The sensitivity of flow rate of feed water and different number of opening valves were insignificant for progress of the accident and fission product release. In conclusion, operators should select the way of depressurization to reduce the RCS pressure and the fission product release during the SGTR accident, considering the condition of the plant such as accident progress and availability of safety features. To suggest more proper information for depressurization, more sensitivity analysis and detailed thermal-hydraulic analysis should be performed for the future work.

Commentary by Dr. Valentin Fuster
2018;():V007T11A012. doi:10.1115/ICONE26-82161.

During a hypothesized severe accident, a containment building is designed to act as a final barrier to prevent release of fission products to the environment in Nuclear Power Plants (NPPs). However, in a bypass scenario of Steam Generator Tube Rupture (SGTR), radioactive nuclides are released to environment even if the containment is not ruptured. The radioactive nuclides are transported from primary to secondary systems through a broken steam generator tube during SGTR accident. Accordingly, the radioactive nuclides of the secondary system can be released to the environment through Main Steam Safety Valve (MSSV) or Atmospheric Dump Valve (ADV). Thus, thorough mitigation strategies are needed to prevent such unfiltered release of the radioactive nuclides during SGTR accidents. To mitigate the consequence of the SGTR accident, this study was conducted to devise In-Containment Relief Valve (ICRV) from steam generator to the free space in the containment building of the Optimized Power Reactor 1000 MWe (OPR1000). This study focuses on the conceptual development of the mitigation strategy and MELCOR code was used for the numerical simulation. The MELCOR input model of OPR1000 consists of 58 control volumes and 161 flow paths. Safety features such as Pressurizer Safety Relief Valve (PSRV), Safety Injection Tanks (SITs), and MSSV were modeled in the MELCOR model. To initiate the SGTR scenario, a flow path between secondary and primary sides of Steam Generator (SG) was modeled with a flow area of 4.49 × 10−4 m2. The safety features were assumed that a few passive systems such as PSRV, MSSV, and SIT, were available. Under this condition, the ICRV connecting the SG and the free space in the containment such as dome and Reactor Drain Tank (RDT) were modeled. Specifications of the ICRV such as length, flow area, and valve opening condition were assumed to similar to those of the MSSV. Using these paths, three cases were considered; a base case, a case of steam release to the containment dome (CNMT case), and a case of release to the RDT (RDT case). Simulation results show that in the base case released radionuclides to the environment. In the other cases, the radioactive nuclides were not released to the environment although the containment pressure increased more than the base case, which is lack of the ICRV. As a result, the ICRV prevents the radionuclides release to the environment during SGTR accidents. Further studies are needed to incorporate practical valve inputs, reactor type, and safety features to gain more feasibility.

Commentary by Dr. Valentin Fuster
2018;():V007T11A013. doi:10.1115/ICONE26-82243.

The In-Vessel Retention (IVR) strategy for Light Water Reactors (LWR) intends to stabilize and retain the core melt in the reactor pressure vessel. This type of Severe Accident Management (SAM) strategy has already been incorporated in the SAM guidance (SAMG) of several operating small size LWR (reactors below 500MWe, like VVER440) and is part of the SAMG strategies for some Gen III+ PWRs of higher power like the AP1000.

One of the main issues for the demonstration of the success of the IVR strategy lies in the evaluation of the transient heat fluxes applied by the corium pool along the vessel wall. Indeed, these transient heat fluxes, during the corium pool stratification evolution, are expected to be higher than the steady-state ones, in particular due to the concentration of the heat flux in the top metal layer when it is thin (so called focusing effect). Another issue appears when a heavy metal is initially formed and rises later to the top (inversion of stratification): in such a situation, the metal goes through the oxide phase and accumulates a significant superheat which is likely to produce a high transient heat flux. Thus, it is of primary importance to be able to evaluate the duration of these transient peaks in order to evaluate the minimal residual vessel thickness after such fast transient ablation and draw conclusions about the vessel integrity.

This paper first presents the phenomenology associated to the transient molten pool stratification and the model implemented in the severe accident integral code ASTEC (Accident Source Term Evaluation Code) to evaluate this kinetics. Then, evaluations are presented, based on a typical PWR reactor configuration. A sensitivity study is proposed to consider the impact of the main uncertainties on parameters which govern this kinetics. A particular focus is made on the physical phenomena driving the transient stratification of material layers in the corium pool and on the identification of critical situations with possible consequences in terms of vessel failure. The characteristic times of each individual process (chemistry, stratification, natural convection) are compared. In particular, the limiting cases of very fast chemistry or very slow chemistry are evaluated.

This work is performed in the frame of the European H2020 project IVMR (In-Vessel Melt Retention) coordinated by IRSN. This project has been launched in 2015 and gathers 27 organizations with, as main objective, the evaluation of feasibility of IVR strategy for LWR (PWR, VVER, BWR) of total power 1,000MWe or higher.

Topics: Vessels
Commentary by Dr. Valentin Fuster
2018;():V007T11A014. doi:10.1115/ICONE26-82248.

The In-Vessel Retention (IVR) strategy for Light Water Reactors (LWR) intends to stabilize and isolate corium and fission products in the reactor pressure vessel and in the primary circuit. This type of Severe Accident Management (SAM) strategy has already been incorporated in the design and SAM guidances (SAMGs) of several operating small and medium capacity LWRs (reactors below 500 MWe, e.g. VVER440) and is part of the SAMG strategies for some Gen III+ PWRs of higher power such as the AP1000 or the APR1400. However, the demonstration of IVR feasibility for high power reactors requires using less conservative models as the safety margins are reduced.

In Europe, the IVMR project aims at providing new experimental data and a harmonized methodology for IVR. A synthesis of the methodology applied to demonstrate the efficiency of IVR strategy for VVER-440 in Europe (Finland, Slovakia, Hungary and Czech Republic) was made. It showed very consistent results, following quite comparable methodologies. The main weakness was identified in the evaluation of the heat flux that could be reached in transient situations, e.g. under the “3-layers” configuration, where the “focusing effect” may cause higher heat fluxes than in steady-state (due to transient “thin” metal layer on top). Analyses of various designs of reactors with a power between 900 and 1300 MWe were also made. Different models for the description of the molten pool were used: homogeneous, stratified with fixed configuration, stratified with evolving configuration. The last type of model provides the highest heat fluxes (above 3 W/m2) whereas the first type provides the lowest heat fluxes (around 500 kW/m2) but this model is not realistic due to the immiscibility of molten steel with oxide melt. Obviously, there is a need to reach a consensus about best estimate practices for IVR assessment to be used in the major codes used for safety analysis, such as ASTEC, MELCOR, SOCRAT, MAAP, ATHLET-CD, SCDAP/RELAP, etc. Despite the model discrepancies, and leaving aside the unrealistic case of homogeneous pool, the average calculated heat fluxes can reach, in many cases, values which are well above 1 MW/m2. This could reduce the residual thickness of the vessel considerably and threaten its strength and integrity. Therefore, it is clear that the safety demonstration of IVR in high power reactors requires a more careful evaluation of the situations which can lead to formation of either a very thin top metal layer provoking the focusing effect or significantly overheated metal, e.g. after oxide and metal layer inversion. Both situations are illustrated in this paper. The demonstration also requires an accurate thermo-mechanical analysis of the ablated vessel. The standard approach based on “yield stress” (plastic behaviour) is compared with more detailed calculations made on realistic profiles of ablated vessels. The validity of the standard approach is discussed.

The current approach followed by many experts for IVR is a compromise between a deterministic analysis using the significant knowledge gained during the last two decades and a probabilistic analysis to take into account large uncertainties due to the lack of data for some physical phenomena, e.g. associated with molten pool transient behaviour, and due to excessive simplifications of models. A harmonization of the positions of safety authorities on the IVR strategy is necessary to allow decision making based on shared scientific knowledge. Some elements that might help to reach such harmonization are proposed in this paper, with a preliminary revision of the methodology that could be used to address the IVR issue. In the proposed revised methodology, the safety criterion is not based on a comparison of the heat flux and the Critical Heat Flux (CHF) profiles as in the current approaches but on the minimum vessel thickness reached after ablation and the maximum pressure load that is applied to the vessel during the transient. The main advantage of this revised criterion is in consideration of both steady-state and transient loads on the RPV. Another advantage is that this criterion is more straightforward to be used in a deterministic approach.

Topics: Vessels
Commentary by Dr. Valentin Fuster
2018;():V007T11A015. doi:10.1115/ICONE26-82310.

The melt jet breakup phenomenon in a pre-flooded reactor cavity during a severe accident is related to the debris bed coolability. It is important to predict the jet breakup length for the evaluation of the debris bed coolability.

A large volume of works on the jet breakup length have been performed. However, the consistency between experiments and correlations was difficult to achieve. Some data follow the Saito correlation (include Froude number in the correlation) and others follow the Epstein correlation (doesn’t include Froude number).

The separation of the jet breakup length correlation along the water subcooling was reported based on the experimental data using the low melting temperature materials in our previous works. Since the previous experiments show an unclear jet shape before entering the water pool which could be an uncertainty factor, a slide gate system for a clear jet shape was additionally installed. Experiments were conducted with the similar condition of previous work and different initial condition of melt jet. With a clear jet shape, the jet breakup length in the subcooled water show different tendency following the Saito correlation.

To figure out the effect of the entry condition of the melt jet, the jet diameter and the method of estimating the jet breakup length are revisited. Our previo0us experiments show large uncertainties on the jet diameter, leading to the large discrepancy of the dimensionless jet breakup length. Also, early broken jet core is reported in subcooled water cases.

Thus, the uncertain characteristics of the jet breakup length analysis is discussed in this paper including the jet diameter and the method to estimate the jet breakup length.

Topics: Subcooling , Water
Commentary by Dr. Valentin Fuster
2018;():V007T11A016. doi:10.1115/ICONE26-82389.

Multi-objective optimization is a powerful tool that has been successfully applied to many fields but has seen minimal use in the design and development of nuclear power plant systems. When applied to design, multi-objective optimization involves the manipulation of key design parameters in order to develop optimal designs. These design parameters include continuous and/or discrete variables and represent the physical design specifications. They are modified across a specific design space to accomplish a number of set objective functions, representing the goals for both system design and performance, which conflict and cannot be combined into a single objective function. In this paper, a non-dominated sorting genetic algorithm (NSGA) and parallel processing in Python 3 were used to optimize the design of the passive endothermic reaction cooling system (PERCS) model developed in RELAP5/MOD 3.3. This system has been proposed as a retrofit to currently-operating light water reactors (LWR) and is designed to remove decay heat from the reactor core via the endothermic decomposition of magnesium carbonate (MgCO3) and natural circulation of the reactor coolant. The PERCS design is currently a shell-and-tube heat exchanger, with the coolant flowing through the tube side and MgCO3 on the shell side. During a station blackout (SBO), the PERCS initially keeps the reactor core outlet temperature from exceeding 635 K and then reduces it to below 620 K for 30 days. The optimization of the PERCS was performed with three different objectives: (1) minimization of equipment costs, (2) minimization of deviation of the core outlet temperature during a SBO from its normal operation steady-state value, and (3) minimization of fractional consumption of MgCO3, a metric that is measurable and directly related to the operating time of the PERCS. The manipulated parameters of the optimization include the radius of the PERCS shell, the pitch, hydraulic diameter, thickness and length of the PERCS tubes, and the elevation of the PERCS with respect to the reactor core. The NSGA methodology works by creating a population of PERCS options with varying design parameters. Using the evolutionary concepts of selection, reproduction, mutation, and survival of the fittest, the NSGA method repeatedly generates new PERCS options and gets rid of less fit ones. In the end, the result was a Pareto front of PERCS designs, each thermodynamically viable and optimal with respect to the three objectives. The Pareto front of options as a whole represents the optimized trade-off between the objectives.

Commentary by Dr. Valentin Fuster
2018;():V007T11A017. doi:10.1115/ICONE26-82593.

The paper describes at first the evolution of the Design Extension Condition (DEC) concept, that started in early 1990ies and was accelerated after the Fukushima accident in 2011. Nowadays the concept is elaborated in EUR document [1], WENRA Reference Levels [4], and finally in the IAEA Safety Standards [6][7]. Number of countries is implementing the DEC concept into the national legislation, safety assessment, design modifications and operation of NPPs. Typically, the DEC events are divided to 2 groups: DEC without core melt (DEC-A) and DEC with core melt (DEC-B).

The second part of the paper deals with safety assessment of DEC events with more detailed information about deterministic analysis of DEC-A events. Legislative and methodology basis, acceptance criteria, computer codes and their validation, list of scenarios to be analyzed, examples of analyses performed, and incorporation of results of new analyses into SAR are described step by step.

Topics: Design
Commentary by Dr. Valentin Fuster

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