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ASME Conference Presenter Attendance Policy and Archival Proceedings

2016;():V002T00A001. doi:10.1115/ICONE24-NS2.
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This online compilation of papers from the 2016 24th International Conference on Nuclear Engineering (ICONE24) 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presenter attendance policy, if a paper is not presented at the Conference by an author of the paper, the paper will not be published in the official archival Proceedings, which are registered with the Library of Congress and are submitted for abstracting and indexing. The paper also will not be published in The ASME Digital Collection and may not be cited as a published paper.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

Smart Grids, Grid Stability, and Offsite and Emergency Power

2016;():V002T05A001. doi:10.1115/ICONE24-60153.

This paper presents the results of statistical and engineering analysis of Loss of Offsite Power (LOOP) events registered in four reviewed databases. The paper includes events registered in IRSN (Institut de Radioprotection et de Sûreté Nucléaire) SAPIDE and GRS (Gesellschaft für Anlagen- und Reaktorsicherheit mbH) VERA database in time period 1992 to 2011. The Nuclear Regulatory Commission (NRC) Licensee Event Reports (LERs) database and the IAEA International Reporting System (IRS) database are screened for the relevant events registered in period 1990 to 2013.

In total, 228 relevant events were identified in the IRSN database, 190 in GRS, 120 in LER and 52 in IRS. The data include events registered both during the critical (at power) and shutdown operation of the plants. The identified events were classified considering nine different categories.

In the three databases (SAPIDE, VERA, IAEA-IRS) the largest numbers of events are registered for the plant centered category. The largest number of the events in the NRC-LER database is found for switchyard centered events. According to the mode of operation, most events were reported during critical power operation, in all four databases. The “Partial loss of external power” events are the most frequent type of event found in the IRSN and NRC databases while the “Physical loss of electrical busbars” is the main type in the GRS and IAEA databases. The largest number of events in all databases is identified for the switchyard failures followed by the interconnections failures (both lines and transformers). Mainly LOOP event are identified by the fault report in the control room. Electrical deficiency is detected as the main direct cause of events. Environment is registered as the main contributor for the electrical grid deficiency in the French and NRC databases. Electrical failures are dominant contributor to the electrical grid deficiency in the German and IAEA databases. The principal root cause for the LOOP events are human failures with the human errors during test, inspection and maintenance as the largest sub-group. The largest number of the LOOP events resulted in reactor trip followed by the Emergency Diesel Generator (EDG) start. The majority of the reported LOOP events lasted for more than 2 minutes.

Main lessons learned from the analysed events and potential actions for decrease of the number of LOOP events are presented.

Commentary by Dr. Valentin Fuster
2016;():V002T05A002. doi:10.1115/ICONE24-60154.

This paper presents the results of the trend analysis of Loss of Offsite Power (LOOP) events registered in two reviewed databases. The reviewed databases include the Nuclear Regulatory Commission (NRC) Licensee Event Reports (LERs) database and the IAEA International Reporting System (IRS). Both databases were screened for the relevant events registered in period 1990 to 2012.

The statistical analysis of the identified relevant LOOP events is done. The analysis includes assessment of the LOOP initiating event frequency, distribution of the events per year in the analysed period and trend analysis of the identified events. The LOOP frequency is calculated for LOOP events registered in NRC LERs subdivided into four types by cause or location: plant centered, switchyard centered, grid related, and weather related. These four LOOP categories are assessed for two modes of operation (critical and shutdown operation).

The number of LOOP events in each year over the analysed period and distribution of events per unit in given year were assessed from the reviewed databases.

Trend analysis of the identified events is performed with the utilization of four trend measures. Analysis is done for events registered during power and shutdown operation and their sum. The obtained LOOP frequency for events registered NRC LERs for critical and shutdown mode is comparable to the one assessed in NUREG/CR-6890.

Decreasing trend is obtained for the LOOP events registered for events in NRC LER database. Different trends are identified during critical and shutdown modes of operation for the events registered in the IAEA database. The sum of the LOOP events reported during critical and shutdown modes in IAEA IRS show no trend.

Topics: Trend analysis
Commentary by Dr. Valentin Fuster
2016;():V002T05A003. doi:10.1115/ICONE24-60255.

The electrical disturbance in Forsmark 2006 [3] led to increased attention being paid to the power supplies of nuclear power plants and their role in safety system reliability, both nationally and internationally. Since then numerous disturbances similar in nature have occurred in the electrical power supply which raises questions whether best available technology (BAT) has been utilised in the design and analysis of the electrical power supply of the safety functions of nuclear facilities. On repeated occasions this type of disturbances has had an impact on redundant parts of several safety systems due to functional dependencies between these. The frequency of these occurrences has been unexpected.

The Swedish Radiation Safety Authority (SSM) has decided to write this document to clarify the regulators position on this issue. The document is also intended to support SSM:s assessments and evaluations of the Licensees efforts regarding degraded power supplies.

An assessment of nuclear power plant electrical power systems is necessary in the light of the past years’ operational experience [1–7], where disturbances in the electrical power supply on repeated occasions have caused a power supply with degradation severe enough to challenge plant safety. As the potential consequences of such a degraded power supply can be severe it must be proven that the frequency of such occurrences is tolerably low. Furthermore, it is important to consider experiences from known situations with degraded power supplies, to enable a reasonable approach to identify and take counter-measures based on the root-cause and ensure utilisation of best available technology.

A sufficient approach to enable prevention, protection and mitigation against this type of disturbances has been difficult to identify. Actual events and conditions causing a degraded power supply have often been complex in nature and difficult to anticipate, wherefore events and conditions which has not yet occurred are difficult to foresee. For this reason it is deemed most effective to identify and implement proportional measures that enhances the independence of the power supplies, such that a degraded power supply with a higher reliability is prevented from propagating to multiple parts of the safety systems.

In this memorandum, SSM describes a state-based approach to analysing electrical power system functionality in different states of degraded power supply. The approach is intended to identify potential design weaknesses and measures to enhance robustness. Such an approach is viewed as more favourable in facilitating the identification of such measures, which may otherwise be neglected due to an estimated low frequency of occurrence, or missed due to incomplete identification of possible events and conditions.

Furthermore this document describes how an assessment of electrical power system design can be performed, where the lowest common denominator from operational experience e.g. [1–7] is identified and counteracted. Actual occurrences of degraded power supplies, which all have been “unknown during the event identification process” but “well-known electrical phenomena”, can be described as unidentified degrading conductive disturbances.

Commentary by Dr. Valentin Fuster
2016;():V002T05A004. doi:10.1115/ICONE24-60514.

The Smart Grid is the further development and improvement of power system, with relating characteristics of informatization, digitization and automation as well as interaction, applications including generating, transmission and dispatching and also transformation and distribution of power. Based on the existing and high speed of communication network, the technical supporting system of Smart Grid in China is built with application of modern equipment technology, control method and decision support.

Topics: China , Smart grids
Commentary by Dr. Valentin Fuster
2016;():V002T05A005. doi:10.1115/ICONE24-60519.

Aiming at a wide range and high-frequency reactive voltage fluctuation problem in 750kV power grid which due to large-scale wind power, a coordinated control strategy of a multiple FACTS (Flexible Alternative Current Transmission Systems) configuration is investigated in the paper. System-level and electromagnetic transient level control method of multiple FACTS devices are proposed in the small wind power fluctuation, big disturbance and off-grid level, and the priorities and the principles are determined. The strategy is applied to “Xinjiang and Northwest 750kV main grid interconnection second channel project”, and the correctness and effectiveness is proved.

Topics: Power grids
Commentary by Dr. Valentin Fuster
2016;():V002T05A006. doi:10.1115/ICONE24-60563.

The reliability of the auxiliary power supply of a nuclear power plant (NPP) is of high importance for safe operation. The loss of the electrical power supply is one of the major contributions to the calculated core damage frequency in probabilistic safety assessments. Among others, the events in Forsmark in 2006 [1] and 2012 [2] as well as in Byron in 2012 [3] illustrate that disturbances in the external power grid can propagate into the NPP and have an impact on the safety important electrical equipment. Therefore, the grid reliability contributes considerably to the reliability of the auxiliary power supply.

In the research work presented in this paper the international operating experience has been evaluated concerning events which include disturbance in the external grid to discover those types of grid disturbances which may have influence on the safe operation of the NPPs. The identified events have then been categorized within a developed classification scheme to determine those with the highest relevance. Based on this scheme representative scenarios of grid disturbances have been developed.

The investigation of the impact of the developed scenarios on the electrical equipment of NPPs will be performed using a grid analysis, planning and optimization tool which also allows executing dynamic simulations of electrical grids [4]. Therefore, a generalized auxiliary power supply of a pressurized water reactor was modeled according to German NPPs of the type Konvoi.

In this paper, an overview of the developed scenarios of grid disturbances and the actual status of the simulation of the auxiliary power supply of NPPs is presented.

Commentary by Dr. Valentin Fuster
2016;():V002T05A007. doi:10.1115/ICONE24-60645.

The nuclear power site resource is very rich in Jiaodong Peninsula of Shandong Province. It is suitable for construction of the large nuclear power base. The transmission scope and direction of Jiaodong Peninsula nuclear power base is analyzed, and optional transmission plans of Plant 1, Plant 2, Plant 3 and Plant 4 are proposed. The transmission plans are recommended based on technical and economic comparison, which provide good references for construction of the large-scale nuclear power base and power grid development planning.

Jiaodong Peninsula nuclear power base is planned to be built in the year of 2016–2030, planning capacity of which is 30500MW. The site of nuclear power base is 100∼400km away from the power load center. The nuclear power will use AC transmission and mainly meet the demand of local power load. The early 15500MW gensets will be accessed to the power grid at 500kV, as the following 15000MW gensets will be accessed at 1000kV UHV (Ultra-high voltage) grid. As the accessing of many large-capacity gensets will produce huge impact to the short-circuit current, sectionalized double-bus configuration is recommended in the 500kV main electrical wiring to reduce the short-circuit current of 500kV bus of nuclear power plant. Double bus section cross wire connection is presented to make sure that every two generators on each bus will be connected to different substations on two transmission lines which are set up on different poles and in different paths, to improve the reliability of the power plant. Through analysis and provement, the construction of large nuclear power base must be based on large and stronge power grid, especially the UHV (Ultra-high voltage) AC grid, to meet the demand of huge nuclear power transmission, and to improve the ability of power exchange and ensure the safety of regional power supply.

Also, as the nuclear power plant should better be in base-load operation, the construction of large-scale nuclear power base, would make the system load-control demands increase, which leads to more prominent problems. In order to avoid adding additional depth of peaking power operation and reducing the overall economic operation of power system, power grid should have the necessary means to load-control. Namely the construction of peaking units, such as pumped storage units or gas-fired units at about 5000MW.

By analyzing and demonstration, large-scale nuclear power base must rely on large-scale power grid, particularly the support of UHV power grid in order to meet the demond of large-scale power transmission and electricity exchange, and also to ensure regional security of electricity supply.

Topics: Nuclear power
Commentary by Dr. Valentin Fuster
2016;():V002T05A008. doi:10.1115/ICONE24-60729.

Nuclear power plants contain hundreds of kilometers of electrical cables including cables used for power, for instrumentation, and for control. It is essential that safety-related cable systems continue to perform following a design-basis event. Wholesale replacement of electrical cables in existing plants facing licensing period renewal may be both impractical and cost-prohibitive. It is therefore important to understand the long term aging of cable materials to have confidence that aged cables will perform when needed. It is equally important in support of cable aging management to develop methods to evaluate the health of installed cables and inform selective cable replacement decisions.

The most common insulation materials for electrical cables in nuclear power plants are cross-linked polyethylene and ethylene-propylene rubber. The mechanical properties of these materials degrade over time in the presence of environmental stresses including heat, gamma irradiation, and moisture. Mechanical degradation of cable insulation beyond a certain threshold is unacceptable because it can lead to insulation cracking, exposure of energized conductors, arcing and burning or loss of the ability of the cable system to function during a design-basis accident.

While thermal-, radiation-, and moisture-related degradation of polymer insulation materials has been extensively studied over the last few decades, questions remain regarding the long term performance of cable materials in nuclear plant-specific environments. Identified knowledge gaps include an understanding of the temperature-dependence of activation energies for thermal damage and an understanding of the synergistic effects of radiation and thermal stress on polymer degradation. Many of the outstanding questions in the aging behavior of cable materials relate to the necessity of predicting long-term field degradation using accelerated aging results from the laboratory. Materials degrade faster under more extreme conditions, but extension of behavior to long term degradation under more mild conditions, such as those experienced by most installed cables in nuclear power plants, is complicated by the fact that different degradation mechanisms may be involved in extreme and mild scenarios. The discrepancy in predicted results from short term, more extreme exposure and actual results from longer term, more mild exposures can be counter intuitive. For instance, due to the attenuation of oxidation penetration in material samples rapidly aged through exposure to high temperatures, the bulk of the samples may be artificially protected from thermal aging. In another example, simultaneous exposure of cable insulation material to heat and radiation may actually lead to less damage at higher temperatures than may be observed at lower temperatures. The Light Water Reactor Sustainability program of the United States (US) Department of Energy (DOE) Office of Nuclear Energy is funding research to increase the predictive understanding of electrical cable material aging and degradation in existing nuclear power plants in support of continued safe operation of plants beyond their initial license periods. This research includes the evaluation and development of methods to assess installed cable condition.

Commentary by Dr. Valentin Fuster
2016;():V002T05A009. doi:10.1115/ICONE24-60987.

The history of building and operating nuclear power plants (NPPs) in Germany dates back to the late 1950s and will come to an end in 2022. By then all NPPs still in operation will have to shut down in a defined sequence, according to the revisions made to the German Atomic Energy Act as a consequence of the accident at the Fukushima Dai-Ichi NPP. Nine out of 17 NPPs have already been shut down permanently as a consequence.

Due to the progress in science and technology, the design of the electrical power supply of German NPPs got more complex and hardened against various scenarios with time. The latest generation of NPPs built in Germany in the late 1980’s — the pressurized water reactor of type Konvoi — was designed with Defence in Depth in mind. They are connected to several voltage levels of the power grid and feature two layers of AC emergency power systems, each of which fulfills the n+2 redundancy criteria. The second of those layers is especially hardened against the influence of certain internal and external events and is part of an emergency control system which can keep the plant in a safe state autonomously for 10 hours under certain conditions.

With this being the state of science and technology at that time in Germany, most of the older NPPs in operation had been retrofitted by 2011 with systems that were designed to partially compensate for these plants’ weaker original design.

Various events such as the accident at the Chernobyl 4 NPP in 1986 and also the accident at the Fukushima Dai-Ichi NPP in 2011 led to changes in the German regulatory framework and recommendations to the NPPs for further retrofitting activities. In the regime of electrical power supply, the latest changes in requirements and corresponding retrofitting of the NPPs in operation include mobile diesel generators with corresponding, redundant feeding points, an enhanced coping capability for station blackouts with only DC-power left and measures to ensure bringing back AC-power within the available time.

In this presentation the author gives an overview over the historic development of the electric design in German NPPs and discuss details of the most recently added requirements on retrofitting — e.g. in the new regulatory framework — to enhance the robustness of the electrical power supply of those NPPs. An update on the progress on the actual retrofitting process of the German NPPs with respect to these new requirements is given.

Commentary by Dr. Valentin Fuster
2016;():V002T05A010. doi:10.1115/ICONE24-61075.

In this paper we present the results of the CSNI task group on the Robustness of Electrical systems in the light of the Fukushima Daiichi accident (ROBELSYS). Based on the conclusion of a workshop organized in April 2014 and on further work, this task group has identified the three following main areas for international cooperation: Enhancement of the robustness of electrical systems, development and improvement in the analysis and simulation of the behavior of NPP’s electrical systems and safety challenges related to the use of power and software based electronics in electrical power systems. The paper concludes on the recently approved CSNI framework that will address these issues.

Commentary by Dr. Valentin Fuster
2016;():V002T05A011. doi:10.1115/ICONE24-61077.

An open phase condition is a known phenomenon in the power industry and is now recognized to have adverse impact on the electrical power systems in several nuclear power plants. An open phase condition may result in challenging plant safety. Operating experience in different countries has shown that the currently installed instrumentation and protective schemes have not been adequate to detect this condition and take appropriate action. An open phase condition, if not detected and disconnected in a timely manner, represents design vulnerability for many nuclear power plants. It may lead to a condition where neither the offsite power system nor the onsite power system is able to support the safety functions, and could propagate to station blackout. The design of electrical power systems needs to be evaluated systematically and improved, where necessary, to minimize the probability of losing electric power from any of the remaining supplies as a result of single or double open phase conditions. The improved design should be coordinated with existing measures to ensure that the electrical power system is able to support the safety functions after the open phase condition is detected and disconnected. In this regard, the IAEA has developed a safety publication dealing with design vulnerability of open phase conditions. This paper summarizes the contents of the report, the rationale and criteria to enhance the safety of nuclear power plants by providing technical guidance to address an open phase condition vulnerability in electrical systems used to start up, operate, maintain and shutdown the nuclear power plant.

Commentary by Dr. Valentin Fuster
2016;():V002T05A012. doi:10.1115/ICONE24-61084.

In 2004, a decision was made to perform a modernization and a new power uprate of unit 2 at Oskarshamn nuclear power plant in Sweden. Among the most important reasons for this decision were new safety regulations from Swedish Radiation Safety Authority and ageing of important components. A project was established and became the largest nuclear power modernization in the world.

The modernization led to the need of analysing the auxiliary power system to ensure that it could supply the unit after the uprate, given tolerances on current, voltage and frequency. During the process of developing models for the diesel generator sets, it turned out that the suppliers could not deliver enough satisfactory material for modelling the diesel engines, the speed controllers and the magnetization systems. Therefore, Oskarshamn nuclear power plant with the help of the manufacturers of the diesel generator sets carried out additional measurements in order to collect data for modelling.

Based on electric circuit diagrams provided by the manufacturers, block diagrams of the magnetization systems were made. For the speed controllers, no information was available at all so it was assumed that the controller was of PI-type. The parameters of the magnetization systems and the speed controllers were then tuned using the measurement results.

Finally, a comparison between simulated results and the measurement results were made, showing good agreement. This is especially true in the most commonly used operating interval of the diesel generator sets.

Commentary by Dr. Valentin Fuster

Advanced and Next Generation Reactors, Fusion Technology

2016;():V002T06A001. doi:10.1115/ICONE24-60017.

The reliability of steam generator is extremely important for the sodium-cooled fast reactor nuclear power plant safety and stable operation. The convective heat transfer mechanism of the once-through steam generator (OTSG) of China Experimental Fast Reactor (CEFR) was researched. The water/steam side was divided into four areas according to the heat flux and steam quality, named subcooled, nucleate-boiling, film-boiling, and superheater. In order to accurate determine the DNB, the CHF table was used in this paper. Based on the homogeneous flow model and fixed boundary method, a thermal-hydraulic simulation system, which named OTAC, was established in this paper. To evaluate its performance, the predictions of this method were compared with PSM-W code. The maximum difference between the temperatures predicted by this model and PSM-W was ∼5K. The calculated results are consistent with the actual experiment data, which indicates the correctness of the mathematical model and simulation method. Static and dynamic characteristic researches of CEFR OTSG have done in the simulation system. And the system can be used to simulate the OTSG dynamic in real-time.

Topics: Simulation , Boilers
Commentary by Dr. Valentin Fuster
2016;():V002T06A002. doi:10.1115/ICONE24-60022.

Due to the high inertia of the metal coolant, the safety concerns of the next generation LMRs (e.g. the Advanced Lead Fast Reactor European Demonstrator - ALFRED) have some connections with the core compaction phenomenon when severe earthquake occurs. In this paper the effects on the fuel assemblies (FAs) are numerically analyzed (by FEM code) taking into account suitable boundary and initial conditions. To characterize the interaction between the internal components, surface-to-surface contact condition has been implemented.

The results indicate that the annular area neighboring the piping penetration ovalizes and so a circumferential buckling occurs. The FAs undergo bending deformation especially in correspondence of the half height of the elements. The displacement varies along the vertical axis (direction of maximum flexibility) reaching, in some time interval, the maximum value of about 9 cm. Vibration phenomenon also appeared.

Topics: Compacting
Commentary by Dr. Valentin Fuster
2016;():V002T06A003. doi:10.1115/ICONE24-60029.

The Vacuum Vessel Pressure Suppression System (VVPSS), a key safety system of the ITER plant, is designed to protect the Vacuum Vessel (VV) from over pressure occurring in the case of LOCA (Loss Of Coolant Accident) or other pressurizing accidents such as LOVA (Loss Of Vacuum Accident). The steam condensation in the Suppression Tanks (main elements of the VVPSS system), occurs at sub-atmospheric pressure.

The steam condensation, at pressures equal or greater than the atmospheric, has been numerically analyzed and experimentally investigated in the past in order to optimize the design of the pressure suppression system of boiling water nuclear reactors.

However, very limited experimental data is available concerning the steam condensation in a water tank at sub-atmospheric pressure. In order to analyze the steam condensation in these operating conditions, an experimental study, funded by ITER Organization, is conducted at the Department of Civil and Industrial Engineering (DICI) of University of Pisa.

The tests analyze the condensation of saturated or superheated steam at sub-atmospheric pressures (4.2 kPa and slightly above the water vapour saturation pressure), and pool temperature up to 50°C at several heights of water head.

The experimental facility, to perform this study, has been set up with a significant scaling factor regarding the full size installation at ITER. In this paper the experimental rig, the conditions of the experiments, and the test matrix are presented. The temperature and pressure measurements with details of the data acquisition system are described.

The tests were performed at different patterns of the sparger exit holes (1, 3 and 9) and for three steam mass flow rates per one hole. The results show very high efficiency of condensation for all examined conditions.

Finally, a comparison between the condensation regimen at sub-atmospheric and at atmospheric pressure is discussed.

Commentary by Dr. Valentin Fuster
2016;():V002T06A004. doi:10.1115/ICONE24-60064.

In the framework of generation IV international forum (GIF), safety design criteria (SDC) and safety design guideline (SDG) for the generation IV sodium-cooled fast reactors (SFRs) have been developed as part of the worldwide deployment of SFRs. Japan Atomic Energy Agency (JAEA) and the Mitsubishi FBR Systems, Inc. (MFBR) have been investigating design study of an advanced loop-type SFR to satisfy SDC in the feasibility study of SDG for SFR. In this study, the ability of the pump-integrated Intermediate Heat Exchanger (IHX) is evaluated as a safety measure for the advanced loop-type SFR. Furthermore, maintainability and reparability of the safety measures are taken into account for the advanced loop-type SFR design study. The pump-integrated IHX has been modified to satisfy these requirements. This paper describes the modifications to withstand severe earthquake, primary coolant leak and sodium-water reaction. Also, this paper includes evaluations of thermal transient, structural vibration with pump rotation and wear-out of IHX tubes for they have been adversely affected by the modifications.

Commentary by Dr. Valentin Fuster
2016;():V002T06A005. doi:10.1115/ICONE24-60076.

CEFR, which is the acronym of China Experimental Fast Reactor, as the fourth generation advanced nuclear power system, plays a very important role in the development of nuclear industry in the future. Currently, as a kind of study, has not yet formed a fuel rod and fuel assembly manufacturing company in China. Realizing the localization of fast reactor fuel rod, fuel assembly and equipment has a very positive meaning to China’s nuclear industry development. At the same time for the CJNF’s development, promoting the manufacture level and scientific research technology of product, the research plays an extremely important role.

In this paper, a fuel rod extrusion pit device and a fuel rod cladding pipe wire equipment have been designed according to the characteristic of fast reactor components. The fuel rod extrusion pit device to ensure pit deepness is 1.8–2.0mm, the radius of pit is 3mm, three pits into each other 120°, deviation is ±5°, the distance is 450.4±0.3mm. And the fuel rod cladding pipe wire equipment meet the product technical requirements for the pitch of 100±5mm, each distance deviation shall not be more than ±15mm, wire tension force for 100±20N. Massive tests on the extrude pits, press plug, load pellet, wire and spot welding, fuel rods ring welding and other important process step have been conducted and develop a set of optimized production plan.

Fast reactor fuel rods structure have a enormous difference with the past other types of products have, the control of the input power for welding and an increase in the pressure of sealed welding chambers have been employed and avoided the problem of weld inflatable and weld porosity successfully. Results showed that the study on manufacturing of fast reactor fuel rod achieves success and realize the localization of fast reactor fuel rod in CJNF.

Commentary by Dr. Valentin Fuster
2016;():V002T06A006. doi:10.1115/ICONE24-60113.

In thermal hydraulics designing and safety analysis of the High Temperature gas-cooled Reactor-Pebble Bed (HTR-PM), the THERMIX code was used to study the behavior of helium in the primary coolant system. Once the helium leaks out of the primary loop through a break on the pressure boundary or an inadvertent open relief valve, it is difficult to simulate the conditions of the room where the release occurred with THERMIX. In this paper, the latest version of RELAP5/MOD4 was used to simulate the behavior of the helium released to the containment rooms. A RELAP5/MOD4 input deck of the HTR-PM, consisting of the core, the primary coolant system, the secondary loop and the containment, was developed and evaluated in this paper. Based on the model, this paper simulated the accidents consequences of large breaks or small breaks near the inlet or the outlet of the helium circulator located inside the steam generator pressure vessel. The calculating results illustrate that the temperature of the helium flowing into the reactor building through the break was no more than 280°C even after an un-isolating large break. The analysis shows that the systems function to scram the reactor and to monitor the core temperature and pressure after accidents would not be affected by breaks.

Topics: Accidents
Commentary by Dr. Valentin Fuster
2016;():V002T06A007. doi:10.1115/ICONE24-60123.

Dual-functional Lithium Lead Test Blanket Module (DFLL-TBM) was proposed by China for testing in the International Thermonuclear Experimental Reactor (ITER).When an in-TBM helium coolant tube breaks, high pressure helium will discharge into the Pb-Li breeding zones. The pressure shock in the TBM will threaten the structural integrity and safety of ITER. Simulation and analysis on helium coolant tube break accident of DFLL-TBM was performed, and two cases with different break sizes were considered. Computational results indicate that intense pressure waves spread quickly from the break to the surrounding structures and the variation of pressure in the TBM breeding box is drastic especially when the pressure wave propagation encounters large resistance such as at the bending corner of the flow channel, the inlet and outlet of Pb-Li, etc. The maximum pressure in the TBM breeding box which is even higher than the operating pressure of helium also occurs in these zones. Although the pressure shock lasts for a very short time, its effect on the structural integrity of DFLL-TBM needs to be paid attention to.

Commentary by Dr. Valentin Fuster
2016;():V002T06A008. doi:10.1115/ICONE24-60144.

China Fusion Engineering Test Reactor (CFETR) is under design recently, in which a conceptual structure of the helium-cooled solid breeder blanket is proposed as one of the candidate tritium breeding blankets. In this concept, three radial arranged U-shaped breeding zones are designed and optimized for higher Tritium Breeding Ratio (TBR) and structure simplification. This blanket uses the Li4SiO4 lithium ceramic pebbles as the breeder, while beryllium pebbles as the neutron multiplier. In this paper, the thermal and fluid dynamic analyses of the optimized typical outboard blanket module are performed by CFD method, where the nuclear heating rate is obtained from the preliminary neutronics calculations. The thermal hydraulic behaviors of the first wall (FW), the temperature distributions of submodule structure material, Li4SiO4 pebble bed and Beryllium pebble bed under normal and critical conditions are calculated, respectively. The results show that the temperature on the blanket module can be effectively cooled below allowable temperature limits of the materials, even if the FW is suffering the maximum surface heat flux, which verified the reasonability of the design of the blanket cooling scheme. In addition, several parametric sensitivity studies are conducted to investigate the influences of main parameters (e.g. coolant mass flow rate, inlet temperature, pebble bed thermal conductivity and fusion power) on the temperature distributions of the blanket components.

Topics: Helium
Commentary by Dr. Valentin Fuster
2016;():V002T06A009. doi:10.1115/ICONE24-60156.

The Sodium-cooled Fast Reactor (SFR) is one of the most promising concepts suggested for Generation-IV nuclear reactor systems. Some SFRs adopt Steam Generator (SG) as their heat exchange system between sodium and water. Sodium-water reaction occurs in the tube failure accident of a SG. The tube failure may propagate to adjacent tubes resulting in a large scale tube failure by this reaction. In an advanced loop-type SFR design promoted by Japan Atomic Energy Agency (JAEA), a straight double-walled tube SG is adopted to prevent this sodium-water reaction [1], [2]. The double-walled tube is expected to prevent water leakage by acting as double wall boundary and mitigate consequences of the sodium-water reaction. It is expected for the outer tubes to practically behave as waste resistant for the adjacent tubes to mitigate sodium water reaction consequences. Mitigation is expected in Design Extension Conditions (DECs) such as the loss of the mitigation function which might lead an initial water leakage to large scale tube failure. In addition to the prevention of the initial leakage, the initial water leakage rate is practically suppressed because of the narrow gap between the inner and the outer tube.

In this paper, tube failure propagation has been calculated to assess property protection performance on outer tubes. The evaluation results showed that the total leakage rate is limited to one double-ended guillotine scale hence the double-walled tube SG has the property protection performance. By additional calculations assuming the loss of the mitigation function, a sever event in DECs is cleared. These calculations suggest that increase of the reliability of water blowdown system and enhancement of the pressure release system are effective for the boundary integrity between primary and secondary cooling systems.

There is an issue to be addressed to adopt the concept described above, that is, the decrease of temperature difference between exchange tubes especially for structural integrity of the straight double-walled tube SG for its thermal contact resistance between double-walled tubes and its lack of bending part to release thermal stress. The dispersion of thermal contact resistance between tubes causes temperature difference there due to their heat transfer rate difference. To suppress this dispersion, the oxidized scale is reduced on the interface between the inner and the outer tubes by applying heat treatment using hydrogen furnace for the tube element production. Then, thermal contact resistance of the double-walled tube is successfully reduced at laboratory scale.

Thus, these results suggest that the double-walled tube SG may suppress water leakage rate and sodium-water reaction consequences in DECs. Furthermore, temperature difference between exchange tubes due to oxidized scale on the interface between the inner and the outer tubes can be reduced at laboratory scale. Hereafter, the specifications of the double-walled tube SG will be determined including tolerance reinforcement of sodium boundary.

Topics: Boilers , Design , Sodium , Water
Commentary by Dr. Valentin Fuster
2016;():V002T06A010. doi:10.1115/ICONE24-60232.

One important safety design consideration for high temperature gas-cooled reactor (HTGR) is air ingress following a rupture of the reactor pressure boundary such as primary piping. The air intrusion to the reactor core held at high temperature through the break will results in significant oxidation of graphite components and fuels. Such oxidation may leads to the weakening of core support structures as well as fuel element damage and subsequent fission product release.

This paper intends to propose a practical solution to protect the reactor from severe oxidation against air ingress accidents without reliance on subsystems. Firstly, a change is made to the center reflector structure to minimize temperature difference during the accident condition in order to reduce buoyancy-driven natural circulation in the reactor. Secondly, a modified structure of the upper reflector is suggested to prevent massive air ingress against a rupture in standpipes. As a preliminary study, a numerical analysis is performed for a typical prismatic-type HTGR to study the effectiveness of the proposed design concept using simplified lumped element models. The analysis considers internal decay heat generation and transient conduction from inner to outer regions at the reactor core, cooling of vessel outer surface by radiation and natural convection, and natural circulation flow in reactor. The results showed that amount of air ingress into the reactor can be significantly reduced with practical changes to local structure in the reactor.

Commentary by Dr. Valentin Fuster
2016;():V002T06A011. doi:10.1115/ICONE24-60328.

An identification method based on growing and pruning radial basis function network (GAP-RBFN) is presented for modeling an accelerator driven system (ADS). Compared with traditional neural networks, GAP-RBFN could automatically adjust the number of hidden neurons to find a suitable network structure by using growing and pruning strategies. In addition, an extended Kalman filter (EKF) algorithm is adopted to update network parameters of neurons in GAP-RBFN, which has a rapid convergence speed during the training process. A numerical calculation code named ARTAP (ADS Reactor Transient Analysis Program) is used to generate data for training GAP-RBFN. After GAP-RBFN is trained by the data, an identification model for ADS is established. The simulation results obtained from the GAP-RBFN model are compared with those obtained from a recurrent neural network (RNN) model. It is shown that the GAP-RBFN model not only has higher prediction accuracy than the RNN model, but also has faster computation speed than the numerical calculation code. Owing to its accuracy, simplicity and fast computation speed, the proposed GAP-RBFN method can be used to model the ADS reactor.

Commentary by Dr. Valentin Fuster
2016;():V002T06A012. doi:10.1115/ICONE24-60333.

Nuclear power plant is a large-scale complicated system, which includes reactor core, steam generator, turbine and other important components. These components are tightly coupled with each other. Among these components, steam generator is the key link of primary circuit system and secondary circuit system. Heat transfers from primary side to secondary side. Once-through steam generator applied in high temperature gas cooled reactor (HTGR) has the properties of small heat capacity and rapid response speed. In HTGR system, the steam generator should match with the properties of reactor core such as large heat capacity and the slow response. Therefore, accurately simulating the steam generator is a complex task and has a great impact on the coupling property of a reactor system. To address this issue, effects of boundary conditions on the output water quality are analyzed and time integration schemes of backward differentiation formula (BDF) are implemented to HTGR steam generator simulation code BLAST in this work. The introduced BDF is a higher-order approximation to a transient term. It can reduce the numerical error from an explicit time integration scheme. The modified code is numerical tested in a noteworthy HTGR accident operation condition: Pressurized Loss Of Forced Cooling (PLOFC) accident. The performance of HTGR steam generator in the accident is analyzed. The accuracy of the improved algorithm is compared with the original BLAST code. Result shows the safety characteristics of steam generator in PLOFC accident and indicates that the numerical accuracy is significantly improved for both helium and water sides by BDF. For the consideration of accuracy and stability, BDF2 is chosen in the modified BLAST code.

Commentary by Dr. Valentin Fuster
2016;():V002T06A013. doi:10.1115/ICONE24-60352.

This research developed a Kinetic Monte Carlo (KMC) method for simulating hydrogen diffusion in tungsten bulk. The KMC inputs such as diffusion paths and energy barriers are taken from DFT calculation results from the literatures. In this simulation model, stable hydrogen interstitial sites in tungsten are the tetrahedral sites on each surface of the bcc lattice, and each site has four tetrahedral neighboring sites, with two neighbors on the same lattice surface and the other two on the adjacent two perpendicular surfaces. A MATLAB script has been developed to perform the diffusion modeling for any given hydrogen concentration and substrate temperature. To compare the simulation results with experiment measurements, modeling configuration of low hydrogen concentration and temperature of 300 K to 2500 K mirroring the experiment conditions was used. The calculated diffusion coefficients at various temperatures match the experiment reference very well. The calculated diffusion coefficients are also fitted to the Arrhenius equation as:

D [m2/s] = 5.59×10−7 exp(−0.426/kBT)

Commentary by Dr. Valentin Fuster
2016;():V002T06A014. doi:10.1115/ICONE24-60355.

The characteristic confirmation test has been demonstrating by using High Temperature engineering Test Reactor (HTTR). The nuclear heat supply performance test, which is one of the characteristic confirmation test is planned to be carried out after restarting of HTTR. Towards the realization of the industrial utilization of a High-Temperature Gas-cooled Reactor (HTGR) cogeneration system as an extension of a nuclear plant, it is important to ensure the reactor safety in the case that thermal-load of the heat application system is fluctuated or lost. The preliminary analysis for the thermal load fluctuation test, which is one of the nuclear heat supply performance test has been investigated. In the analysis, the reactor outlet temperature can continue to be stable against the reactor inlet temperature changing by the thermal fluctuation. It means that HTGR have the capability of absorbing the thermal fluctuation. This paper focuses on the investigation of the mechanism of absorbing the thermal fluctuation. With the reactor inlet temperature increasing, the graphite moderator reactivity keeps negative though the fuel reactivity becomes active. The large negative graphite moderator reactivity enhances the capability of the absorbing thermal fluctuation. In addition, in the middle of the core, the graphite moderator reactivity insertion trend is inverted. This trend is unique to HTGR because of the large temperature difference between top and bottom of HTGR core.

Topics: Stress
Commentary by Dr. Valentin Fuster
2016;():V002T06A015. doi:10.1115/ICONE24-60362.

In many parts of the world, drinking water is not available except through desalination. Most of these areas have an abundance of solar energy, with few cloudy periods. Energy is required for desalination and for producing electricity. Traditionally this energy has been supplied by fossil fuels. However, even in those parts of the world that have abundant fossil fuels, using them for these purposes is being discouraged for two reasons: 1) the emission of greenhouse gases from combustion of fossil fuels, and 2) the higher value of fossil fuels when used for transportation. Nuclear power and solar power are both proposed as replacements for fossil fuels in these locations. Both of these energy systems have high capital costs, and negligible fuel costs (zero for solar) Instead of these two primary forms of energy competing, this paper shows how they can compliment each other, especially where a significant part of the electricity demand is used for desalination.

Topics: Solar energy
Commentary by Dr. Valentin Fuster
2016;():V002T06A016. doi:10.1115/ICONE24-60382.

In the pebble-bed high temperature gas-cooled reactor, there exist randomly located TRISO coated fuel particles in the pebbles and randomly located pebbles in the core, which is known as the double stochastic heterogeneity. In the previous research, the regular lattice pattern was used to approximately simulate the pebble unit cells because the difficulties in modeling the randomly located TRISO geometric. This work aimed at to quantify the stochastic effect of high-temperature gas cooled pebble-bed reactor unit cells, and in view of the strong ability to carry out the accurate simulation of random media, the implicit particle fuel model of Monte Carlo method is applied to analyze to the difference between regular distribution and random distribution. Infinite multiplication factors of the pebble-bed reactor unite cells were calculated by the implicit particle fuel model and simple cube regular lattice pattern at different TRISO packing factor from 0.5%–50%. The results showed that the simple cube regular lattice pattern underestimates the infinite multiplication factors for most packing fractions, but overrates the infinite multiplication factors when the packing fraction is very low.

Commentary by Dr. Valentin Fuster
2016;():V002T06A017. doi:10.1115/ICONE24-60392.

China Fusion Engineering Test Reactor (CFETR) is a test tokamak reactor being designed in China to bridge the gap between ITER and future fusion power plant. As one of the candidates, a water-cooled solid breeder blanket based on PWR (pressurized water reactor) and SCWR (super-critical water reactor) water conditions were proposed. In the concept, multiplying layers separated by three breeding layers were designed and optimized for higher Tritium Breeding Ratio (TBR) and uniform heat distribution. This blanket uses the Li2TiO3 lithium ceramic pebbles as the breeder, while beryllium pebbles as the neutron multiplier. In this paper, the thermal and fluid dynamic analyses of the optimized blanket on both water conditions were performed by numerical simulation, to discuss thermo-hydraulic performance of the blanket using pressurized water/supercritical water as its coolant. The nuclear heating distribution was obtained from the neutronics calculations by MCNP. The thermal hydraulic behaviors of the first wall (FW), structure material, Li2TiO3 pebble bed and Beryllium pebble bed under normal condition were calculated, respectively. It was found that the temperature on the blanket can be effectively cooled on both water conditions, certified the feasibility of the blanket design with pressurized/supercritical water cooling scheme. It indicated that SCWR case had smaller safety margin than PWR case, but SCWR case would lead higher outlet temperature, thermal conductivity and heat exchange efficiency also. In addition, it was found that beryllium was the dominant factor leading a higher TBR. The results would be important to water condition choice for solid blanket in the future.

Commentary by Dr. Valentin Fuster
2016;():V002T06A018. doi:10.1115/ICONE24-60396.

The helium-cooled Gas Fast Reactor (GFR) is one of the six reactor concepts selected for further development in the frame of the Generation IV International Forum (GIF). Since no gas cooled fast reactor has ever been built, a small demonstration reactor is necessary on the road towards the full-scale GFR reactor. A concept of this demonstrator is called ALLEGRO.

The French Commissariat à l’énergie atomique et aux énergies alternatives (CEA) developed between 2001–2009 a pre-conceptual design of both the full-scale GFR called GFR2400 and the small demonstration unit called ALLEGRO (75 MWt). Since 2013 ALLEGRO has been under development by several partners from Czech Republic, France, Hungary, Poland and Slovakia.

No severe accident study of ALLEGRO using a dedicated computer code has been published so far. This paper is the first attempt to perform computer simulations of the ALLEGRO CEA 2009 concept, using MELCOR version 2.1.

A model of the ALLEGRO CEA 2009 concept has been developed with the aim to perform safety analyses; to confirm that MELCOR can be used for such a study, to investigate what scenarios lead to a severe accident and to study in detail the progression of the severe accident during the in-vessel phase.

Several pressurized and depressurized protected scenarios were investigated; four of them are presented in this paper. It was observed that even long-lasting station blackout (SBO) without further failures of the passive safety systems does not lead to a severe accident as long as there is enough water in the decay heat removal (DHR) system. Loss of coolant (LOCA) transients with DHR system in the forced-convection mode can lead to peak cladding temperatures causing limited core damage in the early phase of the accidents, but without further development into core meltdown. On the other hand, LOCA combined with SBO leads to excessive core melting in orders of minutes, which represents a weak point of ALLEGRO 2009 concept. Recommendations were formulated for the further development of the ALLEGRO concept.

Topics: Accidents
Commentary by Dr. Valentin Fuster
2016;():V002T06A019. doi:10.1115/ICONE24-60429.

Fluoride-salt-cooled high-temperature reactors (FHRs) are a new concept that uses solid fuel and employs high temperature liquid salts as both primary and secondary side working fluids. The thermo-physical properties of the fluoride salts and specific heat transfer correlations have been implemented into the RELAP5-MOD3.2 code to enable the code to simulate the transient thermal-hydraulic response in accidents. The thermo-physical properties calculated by the modified code have been benchmarked versus the experimental data. The results of thermal-hydraulic responses of the PB-AHTR in the steady state and simple loss of forced flow from the publications using the RELAP5-3D are used to verify the ability of the modified RELAP5-MOD3.2 to simulate the transients of the FHRs. The code can predict the variations of the thermal-hydraulic parameters well whereas some results may distinguish from that calculated by the RELAP5-3D. The unprotected loss of flow is analyzed by the modified code, indicating that the passive residual heat removal system can mitigate the consequence of the accidents.

Commentary by Dr. Valentin Fuster
2016;():V002T06A020. doi:10.1115/ICONE24-60455.

Under the cooperative effort of the Civil Nuclear Energy R&D Working Group within the framework of the U.S.-Japan bilateral, Argonne National Laboratory (ANL) and Japan Atomic Energy Agency (JAEA) have been performing benchmark study using Japan Sodium-cooled Fast Reactor (JSFR) design with metal fuel. In this benchmark study, core characteristic parameters at the beginning of cycle were evaluated by the best estimate deterministic and stochastic methodologies of ANL and JAEA. The results obtained by both institutions are agreed well with less than 200 pcm of discrepancy on the neutron multiplication factor, and less than 3% of discrepancy on the sodium void reactivity, Doppler reactivity, and control rod worth. The results by the stochastic and deterministic were compared in each party to investigate impacts of the deterministic approximation and to understand potential variations in the results due to different calculation methodologies employed. Impacts of the nuclear data libraries were also investigated using a sensitivity analysis methodology.

Commentary by Dr. Valentin Fuster
2016;():V002T06A021. doi:10.1115/ICONE24-60502.

The Generation IV (GEN-IV) international forum is a framework for international cooperation in research and development (R&D) for the next generation of nuclear energy systems. Concerning the sodium-cooled fast reactor (SFR) system, there are five cooperation projects for R&D. The SFR Safety and Operation (SO) project addresses the area of the safety technology and the reactor operation technology developments. The aim of the SO project includes (1) analyses and experiments that support establishing safety approaches and validating performance of specific safety features, (2) development and verification of computational tools and validation of models employed in safety assessment and facility licensing, and (3) acquisition of reactor operation technology, as determined largely from experience and testing in operating SFR plants. The tasks in the SO topics are categorized into the following three work packages (WP): WP-SO-1 “Methods, Models and codes” is devoted to the development of tools for the evaluation of safety, WP-SO-2 “Experimental Programs and Operational Experiences” includes the operation, maintenance and testing experiences in experimental facilities and SFRs (e.g., Monju, Phenix, BN-600 and CEFR), and WP-SO-3 “Studies of Innovative Design and Safety Systems” relates to safety technologies for GEN-IV reactors such as passive safety systems. In this paper, recent activities in the SO project are described.

Commentary by Dr. Valentin Fuster
2016;():V002T06A022. doi:10.1115/ICONE24-60523.

iB1350 stands for an innovative, intelligent and inexpensive boiling water reactor 1350. It is the first Generation III.7 reactor after the Fukushima Daiichi accident. It has incorporated lessons learned from the Fukushima Daiichi accident and Western European Nuclear Regulation Association safety objectives. It has innovative safety to cope with devastating natural disasters including a giant earthquake, a large tsunami and a monster hurricane. The iB1350 can survive passively such devastation and a very prolonged station blackout without any support from the outside of a site up to 7 days even preventing core melt. It, however, is based on the well-established proven Advance Boiling Water Reactor (ABWR) design. The nuclear steam supply system is exactly the same as that of the current ABWR. As for safety design it has a double cylinder reinforced concrete containment vessel (Mark W containment) and an in-depth hybrid safety system (IDHS). The Mark W containment has double fission product confinement barriers and the in-containment filtered venting system (IFVS) that enable passively no emergency evacuation outside the immediate vicinity of the plant for a severe accident (SA). It has a large volume to hold hydrogen, a core catcher, a passive flooding system and an innovative passive containment cooling system (iPCCS) establishing passively practical elimination of containment failure even in a long term. The IDHS consists of 4 division active safety systems for a design basis accident, 2 division active safety systems for a SA and built-in passive safety systems (BiPSS) consisting of an isolation condenser (IC) and the iPCCS for a SA. The IC/PCCS pools have enough capacity for 7-day grace period. The IC/PCCS heat exchangers, core and spent fuel pool are enclosed inside the containment vessel (CV) building and protected against a large airplane crash. The iB1350 can survive a large airplane crash only by the CV building and the built-in passive safety systems therein. The dome of the CV building consists of a single wall made of steel and concrete composite. This single dome structure facilitates a short-term construction period and cost saving. The CV diameter is smaller than that of most PWR resulting in a smaller R/B. Each active safety division includes only one emergency core cooling system (ECCS) pump and one emergency diesel generator (EDG). Therefore, a single failure of the EDG never causes multiple failures of ECCS pumps in a safety division. The iB1350 is based on the proven ABWR technology and ready for construction. No new technology is incorporated but design concept and philosophy are initiative and innovative.

Commentary by Dr. Valentin Fuster
2016;():V002T06A023. doi:10.1115/ICONE24-60571.

Japan Atomic Energy Agency (JAEA) has carried out research and development to establish the technical basis of High Temperature Gas-cooled Reactor (HTGR) by using High Temperature engineering Test Reactor (HTTR). On March 11th, 2011, the Great East Japan Earthquake of magnitude 9.0 occurred. When the great earthquake occurred, the HTTR had been stopped under the periodic inspection and maintenance of equipment and instrument. In the great earthquake, the maximum seismic acceleration observed at the HTTR exceeded the maximum value in seismic design. The visual inspection of HTTR facility was carried out for the seismic integrity conformation of HTTR. The seismic analysis was also carried out using the observed earthquake motion at HTTR site to confirm the integrity of HTTR.

The concept of comprehensive integrity evaluation for the HTTR facility is divided into two parts. One is the “inspection of equipment and instrument”. The other is the “seismic response analysis” for the building structure, equipment and instrument using the observed earthquake. For the basic inspections of equipment and instrument were performed for all them related to the operation of reactor. The integrity of the facilities is confirmed by comparing the inspection results or the numerical results with their evaluation criteria.

As the result of inspection of equipment and instrument and seismic response analysis, it was judged that there was no problem to operate the reactor, because there was no damage and performance deterioration, which affects the reactor operation. The integrity of HTTR was also supported by the several operations without reactor power in cold conditions of HTTR in 2011, 2013 and 2015.

Topics: Earthquakes
Commentary by Dr. Valentin Fuster
2016;():V002T06A024. doi:10.1115/ICONE24-60675.

The use Total Capital Investment Cost (TCIC) as a figure of merit to evaluate the design of a Nuclear Power Plant can help lead to more economically competitive designs. TCIC includes costs of equipment, labor, materials and the associated time value of capital. The team developed the software tool EVAL, which is capable of determining TCIC impacts for any nuclear island (NI) of any design. EVAL was first used to estimate the effect of modularization on TCIC in constructing the Westinghouse Small Modular Reactor (SMR). In particular, three different construction cases were identified. In the first case, modules are manufactured in the fabrication facility and assembled into Super Modules (SMs) in the on-site assembly area, while SMs are assembled in the hole to form the NI. The second case differs from the first case in the fabrication process, as modules are manufactured in the on-site assembly area. In the third case, the NI is ‘stick built’; i.e., the modules are assembled in the hole, where all connections are performed and the structures are built. The analysis highlighted the positive impact of off-site modular construction on TCIC.

EVAL is based on an open evaluation methodology. In this paper, we present an extension of EVAL that aims to analyze the impact of testing on TCIC. As only few Nuclear Power Plants (NPP) were built in the recent years, testing costs and durations are characterized by a high uncertainty. EVAL was used to evaluate the impact of testing on TCIC, considering a range of realistic data points. Testing costs were expressed as a percentage of total labor costs and TCICs were calculated for the three construction strategies.

EVAL was also used to evaluate the impact of modular testing on TCIC. Modularization allows functional testing and system testing activities to be moved from the installation stage to the fabrication and assembly stages, with a subsequent reduction in labor cost and total construction time. TCIC sensitivities were performed on the fraction of testing activities that can be moved from the installation stage. The number of these activities is dependent on both the design and the technologies used during construction. The analysis showed the positive impact of modular testing on TCIC and demonstrated how EVAL can be a tool capable of helping stakeholder decisions.

Commentary by Dr. Valentin Fuster
2016;():V002T06A025. doi:10.1115/ICONE24-60711.

An experimental campaign investigating the postulated Steam Generator Tube Rupture (SGTR) event, in relevant configurations for Heavy Liquid Metal Reactors (HLMRs), was carried out in the separate-effect facility LIFU5/Mod2, at ENEA CR Brasimone. Ten tests were performed injecting pressurized subcooled water into the reaction tank partially filled by Lead-Bismuth Eutectic alloy (LBE) at 400°C with a cover gas of argon at about 2 bar. Fast pressure transducers, thermocouples and strain gages provided high-quality measurement data for improving the phenomena understanding and supporting the development and validation phase of computer codes for SGTR numerical simulation.

The experimental campaign is composed by two series of tests, characterized by different water pressure: 40 and 16 bar. The first two tests belonging to the low pressure experiments are presented, highlighting the pressurization time trends of the water injection tank, injection line and reaction vessel. The injected water mass flow rate and temperature trends in the reaction vessel were measured. The former test is the reference one and the latter was carried out for investigating the injection of water with higher sub-cooling.

A post-test analysis of the two mentioned tests was carried out by SIMMER-III code. The pressure profile in the water injection tank was set as boundary condition of the calculation. The numerical analysis provided injection line and reaction tank pressurization in agreement with the experimental data. The lower water temperature test provided a better accordance with the measured data, due to the lower evaporation along the injection line. The SIMMER-III analysis also studied the water-LBE interaction from the volume fraction point of view and the energy released in the total reaction tank and in its cover gas.

Topics: Boilers , Rupture
Commentary by Dr. Valentin Fuster
2016;():V002T06A026. doi:10.1115/ICONE24-60715.

In the framework of the EC FP7 LEADER project, an experimental campaign was performed in the LIFUS5/Mod2 facility, at ENEA CR Brasimone, for investigating the postulated Steam Generator Tube Rupture (SGTR) event in a relevant configuration for the spiral tube Steam Generator (SG) of the European Lead Fast Reactor (ELFR). Two tests are analysed.

The LIFUS5/Mod2 facility implemented a test section composed by 188 tubes, vertically disposed with triangular pitch, in a shell closed by top and bottom flanges and having a perforated cylindrical wall. The central tube injected water at about 180 bar and 270°C, at middle height of the tube bundle, in the reaction tank partially filled by Lead-Bismuth Eutectic alloy (LBE) at 400°C with an argon cover gas at about 2 bar. It was connected to a 2 m3 dump tank, due to the high injection pressure.

In the reaction tank fast instrumentation was set: 6 fast Pressure Transducers (PTs) acquiring data at 10 kHz for precisely characterize the first injection peaks; 70 low constant time Thermocouples (TCs) to understand the vapour evolution path; and 13 strain gages (SGGs) to measure the strain of the bundle and main vessel.

The first test analysed showed a first pressure peak of about 25 bar, due to pressure wave propagation at the cap rupture instant. It did not appear in the second test as consequence of a leakage from the cap before the complete rupture. The following pressurization caused by the entering of water into the reaction vessel was of an analogues magnitude for both the tests (about 30 bar). The water/LBE interaction lower temperature was reached on the inner ranks of tubes, about 160°C. The outer rank was cooled down to 340°C. The strain gage measurements showed a decreasing deformation on the tubes toward the outer positions. No ruptures were observed on tubes surrounding the injector. The amount of LBE transported into the dump tank was strongly dependent on the LBE level in the reaction tank at the start of the tests.

Topics: Boilers , Rupture
Commentary by Dr. Valentin Fuster
2016;():V002T06A027. doi:10.1115/ICONE24-60716.

In the framework of MAXSIMA project, the design of a large-scale Test Section (TS), aiming to experimentally investigate the Steam Generator Tube Rupture (SGTR) postulated event in a relevant configuration for Gen IV MYRRHA reactor, was carried out. The TS will be implemented in the large pool CIRCE facility, at ENEA CR Brasimone. The TS is composed of four tube bundles representing a full scale portion of the Primary Heat eXchanger (PHX) of MYRRHA plant. They allow the execution of four SGTR tests, one at a time, excluding the necessity to extract the TS from the facility after each test. Water is foreseen to be injected at 16 bar and 200°C in the pool, partially filled by LBE at 350°C with a cover gas of argon at about 1 bar.

The pressurization transients of CIRCE vessel and the sizing of the discharge lines and relative rupture disks were numerically predicted by SIMMER-III code on the base of a preliminary simplified configuration of the TS. The obtained results showed that the design pressure of CIRCE main vessel was not reached during more than 10 s of water injection, implementing a singular rupture disk having a diameter of 2 inch activated at 6.5 bar. It appears more than enough to notice, in a real reactor, the occurrence of the SGTR event and stop the water supply, interrupting the accidental scenario. These numerical results were adopted to support the design of the presented TS.

Topics: Design
Commentary by Dr. Valentin Fuster
2016;():V002T06A028. doi:10.1115/ICONE24-60734.

Many researchers are investigating the potential of lead-bismuth cooled fast reactors for producing electricity, as well as for the safe transmutation of minor actinides and the nuclear incineration of long-lived fission products.

BelV is a Technical Support Organization to the Belgian Nuclear Regulator, FANC, and has begun a project with the goal to identify and analyze key technical issues in core neutronics, thermal-hydraulics, fuels, and materials associated with the development of a fast nuclear reactor with a lead-bismuth coolant.

The paper presents the results from simulating with the RELAP5-3D code of natural circulation in a generic design of a pool-type nuclear reactor with lead-bismuth eutectic alloy (LBEA) as a primary, and water/steam as a secondary coolant.

The simulation results provide valuable insights in the evolution of key reactor safety-relevant phenomena and support also the qualified use of system analysis codes as RELAP5-3D for the simulation of transients in pool-type reactor systems.

Commentary by Dr. Valentin Fuster
2016;():V002T06A029. doi:10.1115/ICONE24-60749.

GE Hitachi Nuclear Energy (GEH) and Argonne National Laboratory are currently engaged in a joint effort to modernize and develop probabilistic risk assessment (PRA) techniques for advanced non-light water reactors. At a high level, the primary outcome of this project will be the development of next-generation PRA methodologies that will enable risk-informed prioritization of safety- and reliability-focused research and development, while also identifying gaps that may be resolved through additional research. A subset of this effort is the development of PRA methodologies that can be used for the determination of passive system reliability while integrating quantitative success criteria into the risk analysis framework. An updated passive system reliability approach has been developed for utilization in the PRISM PRA that systematically characterizes the impact of passive safety systems on key success criteria. This methodology is derived from the Reliability Method for Passive Systems (RMPS), but is refined to explicitly include consideration of overall mission success through satisfaction of success criteria, rather than only focusing on the passive system itself. This paper provides details on the integrated methodology, focusing on the interface between passive system reliability and success criteria. Specific examples for the passive systems/features of interest, RVACS and inherent reactivity feedback, are included. Additionally, aspects of the integrated passive system and success criteria methodology as they relate to the ASME/ANS Non-LWR PRA Standard are identified and discussed.

Topics: Reliability
Commentary by Dr. Valentin Fuster
2016;():V002T06A030. doi:10.1115/ICONE24-60759.

GE Hitachi Nuclear Energy (GEH) and Argonne National Laboratory are currently engaged in a joint effort to modernize and develop probabilistic risk assessment (PRA) techniques for advanced non-light water reactors. At a high level, the primary outcome of this project will be the development of next-generation PRA methodologies that will enable risk-informed prioritization of safety- and reliability-focused research and development, while also identifying gaps that may be resolved through additional research. A subset of this effort is the development of PRA methodologies to conduct a mechanistic source term (MST) analysis for event sequences that could result in the release of radionuclides. The MST analysis seeks to realistically model and assess the transport, retention, and release of radionuclides from the reactor to the environment. The MST methods developed during this project seek to satisfy the requirements of the Mechanistic Source Term element of the ASME/ANS Non-LWR PRA standard. The MST methodology consists of separate analysis approaches for risk-significant and non-risk significant event sequences that may result in the release of radionuclides from the reactor. For risk-significant event sequences, the methodology focuses on a detailed assessment, using mechanistic models, of radionuclide release from the fuel, transport through and release from the primary system, transport in the containment, and finally release to the environment. The analysis approach for non-risk significant event sequences examines the possibility of large radionuclide releases due to events such as re-criticality or the complete loss of radionuclide barriers. This paper provides details on the MST methodology, including the interface between the MST analysis and other elements of the PRA, and provides a simplified example MST calculation for a sodium fast reactor.

Commentary by Dr. Valentin Fuster
2016;():V002T06A031. doi:10.1115/ICONE24-60760.

GE Hitachi Nuclear Energy (GEH) and Argonne National Laboratory are currently engaged in a joint effort to modernize and develop probabilistic risk assessment (PRA) techniques for advanced non-light water reactors. At a high level the primary outcome of this project will be the development of next-generation PRA methodologies that will enable risk-informed prioritization of safety- and reliability-focused research and development, while also identifying gaps that may be resolved through additional research. A subset of this effort is the development of a reliability database (RDB) methodology to determine applicable reliability data for inclusion in the quantification of the PRA. The RDB method developed during this project seeks to satisfy the requirements of the Data Analysis element of the ASME/ANS Non-LWR PRA standard. The RDB methodology utilizes a relevancy test to examine reliability data and determine whether it is appropriate to include as part of the reliability database for the PRA. The relevancy test compares three component properties to establish the level of similarity to components examined as part of the PRA. These properties include the component function, the component failure modes, and the environment/boundary conditions of the component. The relevancy test is used to gauge the quality of data found in a variety of sources, such as advanced reactor-specific databases, non-advanced reactor nuclear databases, and non-nuclear databases. The RDB also establishes the integration of expert judgment or separate reliability analysis with past reliability data. This paper provides details on the RDB methodology, and includes an example application of the RDB methodology for determining the reliability of the intermediate heat exchanger of a sodium fast reactor. The example explores a variety of reliability data sources, and assesses their applicability for the PRA of interest through the use of the relevancy test.

Commentary by Dr. Valentin Fuster
2016;():V002T06A032. doi:10.1115/ICONE24-60829.

An system analysis code coupled with the heat pipe model is developed to analyze the thermal-hydraulic characteristics of the improved TOPAZ-II space reactor power system with a heat pipe radiator. The core thermal-hydraulic model, neutron physics model, and the coolant loop components models (including pump, volume accumulator, pipes and plenums) are established. The designed heat pipe radiator, which replaces the original coolant loop radiator, is also modeled, including the two-dimensional heat pipe analysis model, the fin model and the coolant transport duct model. The start-up procedure of the improved TOPAZ-II system are calculated. The results show that the code can be used to obtained the thermal characteristics of the system start-up process.

Commentary by Dr. Valentin Fuster
2016;():V002T06A033. doi:10.1115/ICONE24-60842.

In the HTTR, 252Cf is loaded in the reactor core as a neutron startup source and changed at the frequency. In this exchange work, there were two technical issues; slightly higher radiation exposure of workers by neutron leakage and reliability of neutron source transportation container in handling.

To reduce the radiation dose by the neutron leakage, detailed numerical evaluations using PHITS code were carried out, the effective shielding method for fuel handling machine was proposed. Easily removable poly-ethylene blocks and particles were used as the neutron shieling, and installed in the cooling paths of the fuel handling machine. As a result, the collective effective dose by neutron was reduced from about 700man-μSv to about 300man-μSv.

As to the neutron source transportation container, the handling performance was improved and the handling work was safely accomplished by downsizing.

Commentary by Dr. Valentin Fuster
2016;():V002T06A034. doi:10.1115/ICONE24-60843.

The future HTGR has been designed in JAEA. The HTGR uses the helium gas as the coolant in primary cooling system. The reactor has many merging points of the higher-temperature helium gas and the lower-temperature helium gas in the cooling system.

Previously, the reactor inlet coolant temperature was controlled lower than the specific one in the HTTR constructed in JAEA. It was confirmed that this event was caused by the lack of mixing of helium gas thermally at the measurement point of the reactor inlet coolant temperature. From this operational experience, it is needed to clear the thermal mixing characteristics of the helium gas at the annulus of the co-axial double-walled piping in HTGR from the viewpoint of the appropriate temperature control in future HTGR.

In this paper, thermal-hydraulic analysis on the helium gas at the annular flow path of the co-axial double pipe with T-junction was conducted to clarify the thermal mixing behavior of the helium gas.

It is shown that the thermal mixing behavior is not so much affected by the flow rate helium gases. Moreover it is difficult to mix the helium gas with the smaller height of the annular flow path. This is caused by smaller contact area of higher- and lower-temperature helium gas and the lack of helium gas flow in radial direction.

It is confirmed that it is difficult to mix the higher- and the lower-temperature helium gas in the annular flow path of the co-axial double-walled piping by using the hydraulic behavior, and it is necessary to arrange the mixing promotor in the annular flow path in order to mix the higher- and the lower-temperature helium gas.

Commentary by Dr. Valentin Fuster
2016;():V002T06A035. doi:10.1115/ICONE24-60858.

In the High Temperature engineering Test Reactor (HTTR), the Vessel Cooling System (VCS) which is arranged around the reactor pressure vessel, removes residual heat and decay heat from the reactor core passively when the forced core cooling is lost. The test was carried out at the reactor thermal power of 9 MW, under the condition that the reactor power control system and the reactor inlet coolant temperature control system was isolated, and three helium gas circulators in the primary cooling system were stopped to simulate the loss of forced cooling flow in the core. Moreover, one cooling line of the VCS was stopped to simulate the partial loss of cooling function from the surface of the reactor pressure vessel. The test results showed that the reactor power immediately decreased to almost zero and was stable as soon as the helium gas circulators were stopped. The power decrease is caused by negative feedback effect of reactivity. On the other hand, temperature changes of core internal structures, the reactor pressure vessel and the biological shielding concrete were slowly during the test. The measured temperature of the reactor pressure vessel decreased for several degrees during the test. The measured temperature increase of biological shielding made of concrete was small within 1 °C. The numerical analysis showed that the temperature increase of VCS cooling tube was about 15°C which is sufficiently small, which did not significantly affect the temperature of biological shielding. As the results, it was confirmed that the cooling ability of VCS can be kept sufficiently even in case that one of two water cooling lines of VCS is lost.

Commentary by Dr. Valentin Fuster
2016;():V002T06A036. doi:10.1115/ICONE24-60876.

In view of the high interest among a number of Member States in the Supercritical Water Cooled Reactor (SCWR) concept, the IAEA launched the second Coordinated Research Project (CRP) on thermal-hydraulics of SCWRs, entitled “Understanding and Prediction of Thermal-Hydraulics Phenomena Relevant to SCWRs” in 2014 to foster international collaboration. The key objectives of this new CRP are to (i) improve the understanding and prediction accuracy of thermal-hydraulics phenomena relevant to SCWRs and (ii) benchmark numerical toolsets for their analyses. At present, 12 institutes participate in the CRP from 10 IAEA Member States, and the OECD/NEA is in cooperation, based on a special agreement with the IAEA, to host a database housing experimental and analytical results contributed from the CRP participants. The expected outcomes from this CRP include (i) enhancement of the understanding of thermal-hydraulics phenomena, (ii) sharing of experimental and analytical results, and the prediction methods for key thermal-hydraulics parameters, and (iii) cross-training of personnel between participating institutes through their close interactions and collaboration.

This paper describes the plan of the new CRP: overall and specific research objectives; tasks and sub-tasks; schedule; and expected outcomes and outputs. It also introduces briefly other IAEA activities to facilitate and support R&D for SCWR technology in Member States, which include technical meetings and training courses.

Commentary by Dr. Valentin Fuster
2016;():V002T06A037. doi:10.1115/ICONE24-60893.

For the prototype sodium-cooled fast reactor, MONJU, the mechanical energy and structural response under energetics caused by neutronic power excursion during Unprotected Loss of Flow accident (ULOF) were preliminarily analyzed. The objective of this study is to demonstrate the integrity of the reactor vessel against the mechanical load induced by the energetics. Conservative energy production was assumed in order to confirm the robustness of the safety design of MONJU. Mechanical energy was evaluated with the code in which mechanistic modelling of core expansion was implemented. The mechanical energy, which were obtained by analyzing the expanding behavior of core materials after energetics, were about one order of magnitude below the thermodynamic work potential calculated by assuming isentropic expansion of the fuel vapor to one atmosphere, which was often used as an indicator to express the severity of the energetics. Structural integrity was then evaluated with coupled fluid-structure dynamics code using the obtained mechanical energy. No or very small circumferential residual strain of the reactor vessel was evaluated in most analytical cases, and even in the most conservative energy production case, the residual strain was only 0.008 % so that the integrity of the reactor vessel is maintained. The result obtained in the present study shows that MONJU has enough robustness against the mechanical load under energetics.

Commentary by Dr. Valentin Fuster
2016;():V002T06A038. doi:10.1115/ICONE24-60914.

In the context of GEN-IV heavy liquid metal-cooled reactors safety studies, the flow blockage in a fuel sub-assembly is considered one of the main issues to be addressed and one of the most important and realistic accident for Lead Fast Reactors (LFR) fuel assembly. The blockage in a fast reactor Fuel Assembly (FA) may have serious effects on the safety of the plant leading to the FA damaging or melting. The temperature of the coolant leaving the FA is considered an important indicator of the health of the FA (i.e. the effective heat removal) and is usually monitored via a dedicated, safety-related system (e.g. thermocouple). The external or internal blockage of the FA may impair the correct cooling of the fuel pins, be the root cause of anomalous heating of the cladding and of the wrapper and potentially impact also fuel pins not directly located above or around the blocked area. In order to model the temperature and velocity field inside a wrapped FA under unblocked and blocked conditions, detailed experimental campaign as well as 3D thermal hydraulic analyses of the FA is required.

The present paper is focused on the CFD pre-test analysis and design of the new experimental facility ‘Blocked’ Fuel Pin bundle Simulator (BFPS) that will be installed into the NACIE-UP (NAtural CIrculation Experiment-UPgrade) facility located at the ENEA Brasimone Research Center (Italy).

The BFPS test section will be installed into the NACIE-UP loop facility aiming to carry out suitable experiments to fully investigate different flow blockage regimes in a 19 fuel pin bundle providing experimental data in support of the development of the ALFRED (Advanced Lead-cooled Fast Reactor European Demonstrator) LFR DEMO. In particular, the ‘Blocked’ Fuel Pin bundle Simulator (BFPS) cooled by lead bismuth eutectic (LBE), was conceived with a thermal power of about 250 kW and a uniform wall heat flux up to 0.7 MW/m2, relevant values for a LFR. It consists of 19 electrical pins placed on a hexagonal lattice with a pitch to diameter ratio of 1.4 and a diameter of 10 mm.

The geometrical domain of the fuel pin bundle simulator was designed to reproduce the geometrical features of ALFRED, e.g. the external wrapper in the active region and the spacer grids. Pre-tests calculations were carried out by applying accurate boundary conditions; the conjugate heat transfer in the clad is also considered.

The numerical simulation test matrix covered the envisioned experimental range in terms of mass flow rate; the wall heat flux was imposed in order to have a fixed temperature difference across the BFPS in unblocked conditions.

The blockages investigated are internal blockages of different extensions and in different locations (central subchannel blockage, corner sub-channel blockage, edge subchannel blockage, one sector blockage, two sector blockage).

High resolution RANS simulations were carried out adopting the ANSYS CFX V15 commercial code with the laminar sublayer resolved by the mesh resolution. The loci of the peak temperatures and their width as predicted by the CFD simulations are used for determining the location of the pin bundle instrumentation. The CFD pre-test analysis allowed also investigating the temperature distribution in the clad to operate the test section safely.

Commentary by Dr. Valentin Fuster
2016;():V002T06A039. doi:10.1115/ICONE24-60919.

iB1350 stands for an innovative, intelligent and inexpensive BWR 1350. The iB1350 uses innovative passive containment cooling system (iPCCS). The iPCCS is a part of the in-containment filtered venting system (IFVS). The vent pipe is submerged in the IFVS tank in the outer well (OW) of the Mark W containment. The conventional PCCS has a suction pipe only from the dry well (DW). On the contrary, the iPCCS has two suction pipes. One is normally opened to the wet well (WW) and another normally closed to the DW. The suction pipe in the conventional design cannot be connected to the WW because the PCCS vent pipe is connected to the WW. A PCCS functions using differential pressure between two nodes to discharge noncondensable gases in a PCCS heat exchanger (Hx). A suction pipe and a vent pipe must be connected to different nodes to use differential pressure. Therefore, the conventional PCCS never can cool the S/P. Although the S/P is the in-containment heat sink, heat up of the S/P is the most unfavorable for the conventional PCCS. In order to use the PCCS the conventional design must discharge steam directly into the DW instead of the S/P. Therefore, the conventional PCCS must open depressurization valves (DPV) at a SBO if the isolation condenser (IC) fails. On the contrary, the iPCCS can cool the S/P directly using the suction pipe connected to the WW and without DPV. Instead of DPV the iB1350 has modulating valves (MV) of which discharge lines are submerged in the S/P. Even if the IC fails during a SBO, the iB1350 can cool the core using the severe accident feedwater system (SAFWS), the SRV or the MV, and the iPCCS. The SAFWS makes up the core. The decay heat is carried by steam to the S/P through the SRV or the MV. The S/P works as in-containment heat sink. Once the S/P starts boiling the iPCCS automatically initiates cooling of the steam from the S/P. In the case of a core melt accident, a certain amount of FP is released into the S/P and heats up the S/P. Once the S/P starts boiling, the noncondensable gases in the WW is purged by the steam into the DW and then into the PCCS Hx. In order to purge the stagnant gases, the conventional PCCS needs an active fan in the long term. On the contrary, the iPCCS can easily purge noncondensable gases in the heat exchanger using differential pressure to the OW and need not any active fan even in the long term.

Commentary by Dr. Valentin Fuster
2016;():V002T06A040. doi:10.1115/ICONE24-61081.

After the full power operation of the 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10), several safety demonstration tests, representing the anticipated transient without scram (ATWS) conditions, were successfully performed on this reactor. Among these tests, two reactivity insertion ATWS tests were conducted by withdrawing a single control rod without reactor scram at 30% rated power.

In the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University, these two tests have been reanalyzed using the THERMIX code, and the code itself was strictly checked through the test data. According to the previous code benchmark activities utilizing the HTR-10 tests, the temperature coefficient of reactivity (TCR), the residual heat level (RHL) and the xenon poisoning effect (XPE) could be considered the most important influencing factors of the THERMIX simulation accuracy for the core dynamics. In this study, sensitivity analyses are performed on the basis of the assumed variations of TCR, RHL and XPE. The impacts of these concerned parameters on the reactor power transient are qualitatively identified.

Commentary by Dr. Valentin Fuster
2016;():V002T06A041. doi:10.1115/ICONE24-61166.

The method for ATWS (anticipated transient without scram) analysis was completely developed for commercial pressurized water (PWR) reactor plants, especially for selecting of typical initial events. For accident analysis of ATWS, it is different between PWR and small modular reactor (SMR), as different structures and characters, and it is necessary to study the typical initial events for these reactors. Based on the standard of PWR, the demanding for ATWS analysis was studied and the consequences for typical anticipated transient was calculated using RELAP5/MOD3.2 code, “maintain reactor coolant pressure boundary integrity” was selected as limiting criterion. The results shows for SMR, anticipated transient with the most serious consequence for ATWS are loss of offsite power and inadvertent control rod withdraw event, this conclusion will support to prepare the safety analysis report and optimum design of diversity activation system (DAS) for SMR.

Commentary by Dr. Valentin Fuster

Safety, Security, and Cyber Security

2016;():V002T07A001. doi:10.1115/ICONE24-60024.

ACP600 is a 600WMe power, self-reliance developed pressurized water reactor design of third generation technology, which is the China National Nuclear Corporation’s (CNNC) achievement of continuous R&D and engineering feedback started with M310 series and 600WMe designs. In general technical project of ACP600, there is a plan to cancel Boron Injection Tank (BIT) in safety injection system and set Reactor Emergency Borating (REB) system in order to enhance safety of nuclear power plant. This paper discusses function determination and system capacity of ACP600’s REB system based on the investigations of both international and domestic advanced third generation nuclear power plant emergency borating system’s safety functions and according to ACP600’s top design and general project along with operation and typical accident analysis, bringing forward to preliminary suggestions about function determination and system capacity of ACP600’s REB system as well as further aspect and content for investigations. This provides concrete information for further design of ACP600’s REB system.

Topics: Emergencies
Commentary by Dr. Valentin Fuster
2016;():V002T07A002. doi:10.1115/ICONE24-60091.

Seismic analysis of the primary system in nuclear plant design is very important. The change of reactor coolant pump has some impact on the support seismic loads of primary equipment. Using the numerical calculation method, preliminary analysis was completed and some valuable results are obtained. The results show the natural frequency of system should be far away from the spectral peak with optimizing the reactor coolant pump configuration to raise the safety and economy of the plant.

Commentary by Dr. Valentin Fuster
2016;():V002T07A003. doi:10.1115/ICONE24-60173.

Ship nuclear power platform is a small and movable power plant on the sea, aiming at generating electric energy and producing fresh water, it provides support for the national energy strategy. Subsequent to a loss of coolant accident (LOCA), steam is vented in the reactor containment following vaporization of liquid and/or steam expansion. The temperature as well as pressure in the condensation rises synchronously. For removing heat and reducing pressure inside containment subsequent to a LOCA, the Passive containment cooling system of Ship nuclear power platform is designed. In order to establish and maintain the passive heat removing channel, steam condenses on the containment condenser tube surface, coupling natural convection of the seawater inside the tubes. The heat transfer mechanism of Passive containment cooling system is very complex. To solve this problem, a three dimensional heat exchanging/one dimensional natural circulation coupling numerical computing method is proposed to obtained the safety performance of the reactor containment. Models of heat exchanging process between steam which contains non-condensable gas inside the reactor containment and sea water outside are firstly established. Then the thermal-hydraulic characteristics of the steam and sea water beside the heat transfer tubes are obtained by a simulation which is carried out in a LOCA.

Commentary by Dr. Valentin Fuster
2016;():V002T07A004. doi:10.1115/ICONE24-60249.

A new experimental program using non-transfer type plasma heating is under consideration in JAEA to clarify the uncertainty on core-material relocation (CMR) behavior of BWR. In order to confirm the applicability of this new technology, authors performed preparatory plasma heating tests using small-scale test pieces (107 mm × 107 mm × 222 mmh). Based on these preliminary results, an excellent perspective in terms of applicability of the non-transfer plasma heating technology to the SA (Severe Accident) experimental study was obtained. Furthermore, JAEA is preparing for the next step intermediate-scale preparatory tests in 2016 using about 50 rods and a control blade that would not only confirm its technical applicability, but also some insights relevant to the issue on CMR itself.

Commentary by Dr. Valentin Fuster
2016;():V002T07A005. doi:10.1115/ICONE24-60250.

Based on lessons learned from the Fukushima Daiichi nuclear power plant accident, pursuit of accident tolerant fuel (ATF) has been discussed by many institutions in the world. Toshiba identified a silicon carbide (SiC) ceramic as the most promising material for accident tolerant fuel. Since SiC has less active characteristics in the presence of high temperature water steam (H2O) and is expected to be tolerant of severe accident conditions. Moreover, SiC has a smaller neutron absorption cross-section which is advantageous feature in terms of neutron economy.

Zirconium alloys (Zry) are one of the main structural materials in LWR core. In high temperature H2O environment under severe accident conditions, Zry rapidly reacts with H2O and oxidation reaction accompanied by release of hydrogen gas occurs. Since SiC may inhibit the progress of oxidation reaction compared to Zry metal alloys, hydrogen and heat generation is expected to decrease in the case of core uncovered accident conditions. In order to confirm the advantage of SiC over Zry as core materials, transient analysis and safety analysis are carried out. For transient analysis, analyses of temperature behavior of cladding at plant transient condition are carried out with best-estimate transient analysis code. This analysis confirmed the effect of physical properties differences between SiC and Zry on cladding temperature behavior. Moreover to indicate the effectiveness of SiC under the core uncovered condition with oxidation reaction, safety analysis with latest “MAAP” code is carried out and the whole plant behavior during severe accident sequence is simulated. This analysis showed the effectiveness of SiC to mitigate the oxidation reaction. As the result of these analyses, the advantage of SiC over Zry can be perceived. And also, future challenges of SiC application as ATF can be clarified through these analyses.

Commentary by Dr. Valentin Fuster
2016;():V002T07A006. doi:10.1115/ICONE24-60267.

For spent nuclear fuel stored within a cooling pond, the essential nuclear safety functions of control, cooling and containment are fulfilled by maintaining an appropriate depth of water above the fuel. External cooling systems remove the decay heat generated by the spent fuel stored within the pond, in order to maintain the temperature of the water at a constant level. In the event of a fault within these external cooling systems, there is the potential for a temperature excursion within the pond. Historically the UK nuclear industry has considered that such faults would pose no threat to the structural integrity of the pond containment and hence the only loss of water would be due to evaporation following a loss of cooling. However, more recently, it has been recognised that such temperature excursions may result in through-wall cracking leading to a loss of water and undermining of these essential safety functions. This paper outlines the safety case implications of these realisations and the way in which they are being addressed within the UK’s nuclear power stations. The paper considers the effects of thermal transient faults on the concrete pond structure and the potential nuclear safety issues which may occur as a result of this.

In response to potential pond cooling faults, consideration is given to the requirement for engineered protection systems along with the safety role of the operator in identifying and responding to faults of this kind. Operators provide a versatile mechanism for identifying fault conditions and taking remedial actions, however, the benefit which can be formally claimed for their role within a safety case is generally limited by the availability or reliability of instrumentation to reveal a fault condition. Post fault operator actions may also be limited by the timescales available following a fault, or by other demands on the operators, which may occur in the event of an external hazard which affects multiple site systems. To quantify the timescales available for post fault remedial action, it is necessary to quantify the rate of water loss from the pond, along with the relationship between pond water depth and the radiological consequences both on-site and off-site. This paper investigates the difficulties which may be encountered in quantifying the role of post fault operator actions within such a safety case, and in demonstrating that the overall nuclear safety risk is acceptably low and as low as reasonably practicable (ALARP).

Commentary by Dr. Valentin Fuster
2016;():V002T07A007. doi:10.1115/ICONE24-60268.

This study investigates the heat transfer characters of a volumetrically-heated melt pool in LWR lower plenum under different top and side boundary conditions. The inconsistency in parameter definitions in Nu-Ra correlations is addressed, and the effect of 2D geometry and crust formation are analyzed. A summary of test conditions of the previous experimental studies are given. The average upwards and downward Nusselt number from LIVE tests in 3D and 2D geometries are compared with each other and with other well-known correlations. The differences of heat flux distribution along the vessel wall are analyzed regarding boundary cooling condition and crust formation. This paper provides some explanations about the discrepant among the exiting heat transfer correlations and recommends most suitable descriptions of melt pool heat transfer under different accident management scenarios.

Commentary by Dr. Valentin Fuster
2016;():V002T07A008. doi:10.1115/ICONE24-60359.

Pool scrubbing is one of the effective mechanisms to filter out radioactive aerosols in a severe accident of a nuclear reactor. A lot of work has been done on the pool scrubbing models and experiments. However, large discrepancies still exist between the simulation and experimental results. To improve the pool scrubbing model, an accurate decontamination factor (DF) evaluation by an aerosol measurement and a detailed two-phase flow structure measurement is inevitable. A pool scrubbing experimental apparatus was constructed by the thermohydraulic safety research group in Japan Atomic Energy Agency. The test section is a transparent pipe with the inner diameter of 0.2 m and the length of about 4.5 m. The aerosol laden air flow was injected upwardly into the pool water. The aerosol particle diameter distribution was measured by a light scattering aerosol spectrometer. White polydisperse BaSO4 particles were used as the aerosol test particles. In the first step, we focused on investigating and reducing the error of DF experimentally. Several problems resulting in the error and their solutions for the error reduction were summarized in this paper. Based on the error reduction methods, the DFs of pool scrubbing were measured in two water submergences. The results showed that the DFs for the aerosol with small diameter were independent of the injecting air velocity in the submergence of 0.3 m. In addition, it was found that the DFs increased with increasing the air flow rate in the submergence of 2.9 m. It was presumed that the increase of DF was dominated by the increase of bubble surface area and/or turbulence intensity with the air flow rate increase, while the effect of the reduced bubble traveling time in the water, which may reduce the DF, was smaller than the increasing effect.

Topics: Aerosols , Errors
Commentary by Dr. Valentin Fuster
2016;():V002T07A009. doi:10.1115/ICONE24-60501.

In this paper, operating characteristics of the safety system of Chinese Supercritical Water-cooled Reactor (CSR1000) is described. Selecting CSR1000 as the focus of research, and it’s active and passive safety systems are analyzed in turn. A comparison is given between these two types of safety systems. Henceforth, the features of the safety control systems of CSR1000 are obtained. The results show that for the active systems, the control speed of the pressure control system is the fastest and that of the power control system is the slowest. It is observed that the active control system exhibits simple harmonic oscillation. On the other hand, the control feature of passive control system is stable. In addition, coupling the safety systems can ensure the safety of CSR1000 in the event of a loss of flow accident (LOFA).

Commentary by Dr. Valentin Fuster
2016;():V002T07A010. doi:10.1115/ICONE24-60676.

After the Fukushima accident the development and implementation of additional safety features has become more and more important.

In March 2014 AREVA filed a patent for the invention of a mobile heat exchanger module in the outer shape similar to the outer geometry of a fuel assembly called Advanced Cooling Tube (ACT) [1]. Since January 2015 there is a full scale test facility in the German headquarters of AREVA in Erlangen to test the performance of such a mobile heat exchanger. In spring 2015 there were investigations running to implement the ACT into the spent fuel pool of a nuclear power plant in Switzerland. In the meantime, these investigations were finished and a contract has been signed.

This paper gives an overview about the main facts of the Advanced Cooling Tube as a mobile heat exchanger in the shape of a fuel assembly for spent fuel pools, how it can be implemented in any nuclear power station and what the advantages of this product are. Further information about the performance and requirements of this special heat exchanger will be given.

Commentary by Dr. Valentin Fuster
2016;():V002T07A011. doi:10.1115/ICONE24-60773.

The cyber security problem is posing new challenges to the current safety analysis of nuclear power plants. Historically, analogue control systems in the absence of interactive communications are immune to cyber-attacks; however, digital control systems with extensive interconnection of reprogrammable components are intensely vulnerable to cyber-attacks which shed light on the significance and urgency of the cyber security. The current cyber security approaches, which merely focus on information networks, have not given multi-faceted considerations to instrumentation and control (I&C) systems. The cyber-attack on I&C systems may lead to more severe consequences, including the abnormal change of parameters, the malfunction of equipment, and even the accident condition. The existing cyber security approaches for information networks, such as firewalls, encryption, can enhance the cyber security of I&C systems, but are often insufficient in addressing challenges associate with the I&C systems which link cyber space and physical systems. The defense approach based on physical information should be developed to meet the emerging challenges. In this paper, we propose the cyber-physical security (CPS) approach based on the physical process data for the cyber defense. This approach does not intend to replace current cyber defense mechanisms. It could be served as the last barrier for security defense. The goal of the CPS defense approach is to detect attacks at the beginning of the occurrence of physical process anomalies cause by cyber-attacks. A practical implementation of the CPS approach is proposed and its influence on the existing infrastructure is discussed. The statistical analysis techniques are utilized on physical process data for attack detection. The method of dynamic principal component analysis (dynamic PCA) is employed to characterize the correlation of multiple variables in the normal operational condition. In the abnormal operational occurrence, the chi-square detector is able to distinguish adversarial cyber-attacks from ordinary random failures.

Commentary by Dr. Valentin Fuster
2016;():V002T07A012. doi:10.1115/ICONE24-60917.

In the course of a severe accident, a large amount of hydrogen gas is generated by a metal-water reaction in a PCV (Primary Containment Vessel) of Light Water Reactors. Although the filter vent of gas mixture, which includes hydrogen and steam, is an effective method for the accident management of BWR that prevents the PCV overpressure, the filter vent at the early stage of severe accident may cause releasing radioactive material to environment. We have been developing the hydrogen treatment system to prevent excessive pressure without PCV vent and releasing radioactive material to environment.

We focus on the oxidation-reduction reaction of metal oxides with high reaction rates, for the hydrogen treatment system. Metal oxide material would be an effective device under low-oxygen conditions like PCV of BWR. The hydrogen treatment system mainly consists of a hydrogen processing unit, a blower and pipes. The hydrogen treatment unit has a lot of reaction pipes in which metal oxides are filled. Some fundamental chemical experiments which we have done have revealed that copper oxides (CuO) rapidly react with hydrogen to form cupper (Cu). Their results show that metal oxides are effective as hydrogen treatment elements. On the other hand, there are few evaluations for the characteristics of hydrogen treatment unit. The dependency of hydrogen treatment performance on gas temperature, hydrogen concentration and pressure is investigated in the present study.

We conducted experiments using a test section with one reaction pipe, which simulated a hydrogen processing unit. The processing materials granulated CuO, MnO2 and Co3O4 with 2mm diameter were used. Gradual increase of processing material temperature in the test section was observed along the gas streams caused by oxidation-reduction reaction after the mixing gases were supplied. Consequently, the hydrogen concentration at the outlet of the test section decreased with time. The increase of the hydrogen reaction rate was also observed with increase of gas temperature, hydrogen concentration and pressure. We have developed the thermal-chemical model of hydrogen processing unit from these experiment results, and confirmed that the model could predict the characteristics of a hydrogen processing unit qualitatively.

Topics: Accidents , Hydrogen
Commentary by Dr. Valentin Fuster
2016;():V002T07A013. doi:10.1115/ICONE24-60927.

Successive safety improvements for nuclear power plants (NPPs) have been required by society as well as by regulatory agencies because of the nuclear accident at the Fukushima Daiichi Nuclear Power Plant due to the Great East Japan Earthquake and Tsunami. The establishment of a methodology for the fragility evaluation of seawalls is essential for developing a probabilistic risk assessment (PRA) for tsunamis that is applicable to NPPs where the hazard level of tsunamis is high.

In the present study, fragility evaluation methods of reinforced concrete (RC) seawalls are documented. Two main damage modes of the seawall, namely overflow and physical damage caused by tsunami wave pressure, were the primary focus. Using the documented fragility evaluation methodology, a conceptual study for evaluating the fragility of a RC seawall against overflow and the impact of tsunami wave pressure is performed, and fragility curves are obtained by considering the following uncertainties: evaluation accuracy of the inundation level and tsunami wave pressure, density of the seawater, compressive strength of concrete, yield strength of reinforcement, and evaluation accuracy of the shear capacity.

Commentary by Dr. Valentin Fuster
2016;():V002T07A014. doi:10.1115/ICONE24-60983.

Monitoring changes in important parameters has been suggested as a potentially useful condition monitoring (CM) method for the accidents occurred in nuclear power plants (NPPs). The reactor core power is believed to be an important parameter governing the performance of reactor during transient response. The accurate prediction of reactor behavior and power is very important for nuclear power plant operators, especially during the major severe accident scenarios following an initiating event such as rod ejection accident (REA) or rod drop accident (RDA). REA and RDA together are found to be the worst postulated power transients in reactor licensing and are referred to as reactivity-initiated accidents (RIAs). RIAs is a postulated event of very low probability and involves inadvertent removal of a control element from an operating reactor, leading to a rapid power excursion in nearby fuel elements. On the basis of our previous work, in which only REA scenario is analyzed, the primary objective of this study is to develop and implement fuzzy weighted support vector regression (FWSVR) for condition monitoring under REA and RDA in NPPs. FWSVR is an extension of support vector regression (SVR) which introduces fuzzy weights in traditional SVR formulation. The accidents simulated in this study are based on the same model used in our previous work. This model can be accomplished by two procedures. First, the neutron flux and enthalpy distributions of the core can be obtained from a solution to the three-dimensional nodal space time kinetics equations and energy equations for both single and two-phase flows respectively. Second, the reactivity effects of the moderator temperature, boron concentration, fuel temperature, coolant void, xenon worth, samarium worth, control element positions (CEAs) and core burnup status can be calculated and determined. For the purpose of condition monitoring, it’s a fundamental issue to acquire condition monitoring data for useful data representation. Otherwise, it’s hard to predict the power accurately if enough data is not available. The data used in this study is collected from computer generated accident scenarios. Then the obtained data is split into two subgroups, training and test. The training subgroup data is utilized to train the FWSVR model and the fuzzy weights introduced in FWSVR are employed to extract multiple linear structures in a training dataset and assign to each data point a cluster index determined by its trained kernel radius function. The test is employed to validate the model. Finally the results of FWSVR model are compared with that of other models such as traditional SVR and back propagation network (BPN) which is one type of artificial neural networks (ANNs). Comparison of results among the three methods indicates that FWSVR model not only outperforms traditional SVR and BPN, but also has a better agreement with the general understanding than them.

Commentary by Dr. Valentin Fuster

Codes, Standards, Conformity Assessment, Licensing, and Regulatory Issues

2016;():V002T08A001. doi:10.1115/ICONE24-60004.

For licensing, design, and construction of nuclear facilities, the investigation of engineering properties of soil borrow and backfill materials is a regulatory requirement, for they impact the performance of these facilities. For instance, the U.S. Nuclear Regulatory Commission (NRC) specifically requires that all applications for nuclear power plants address the source, quantity, static, and dynamic engineering properties of borrow and backfill materials. However, in the nuclear industry, a clear road map does not exist for meeting these requirements. As such, planning, investigation, and/or processes that are needed to address these regulatory requirements become a challenge to applicants. The absence of such a road map can also result in incomplete or unnecessary investigations, licensing cycles, and/or delays. This paper outlines a recommended practice, including steps to design a geotechnical sampling and laboratory investigation program toward addressing these regulatory requirements, with nuclear quality assurance and licensing requirements in perspective. While the steps in this paper may serve as useful guides, requirements vary from project to project; therefore, all efforts should be on developing an investigation program that is project specific in order to meet the actual project objectives. Additionally, this paper provides guidance on presenting the investigation results in regulatory documents.

Commentary by Dr. Valentin Fuster
2016;():V002T08A002. doi:10.1115/ICONE24-60119.

The need to provide standardized guidance for the use of air cleaning systems in nuclear facilities was recognized in the 1960’s when plans for nuclear facilities were at their peak. The American National Standards Institute (ANSI) asked a group of experts to generate documents for boiling-water-reactor standby gas treatment systems for accident mitigation. These experts immediately recognized that their scope needed expansion to include all air cleaning safety systems at all types of nuclear facilities. Their efforts resulted in the issuance of documents that provided guidance for the components in air cleaning systems (ANSI N509), and guidance for the testing of these systems (ANSI N510). Subsequently, it was recognized that this guidance needed to be formalized and implemented in a code format, with requirements instead of recommendations. Various other organizations were also providing guidance in different forms. The US government (the Nuclear Regulatory Commission) issued regulatory guides, and the American Society for Testing of Materials (ASTM) issued many documents providing specifications for activated carbon used in air cleaning systems for radioiodine removal. Acknowledging the need to consolidate all of these documents in a single source, the American Society for Mechanical Engineers (ASME) requested the air cleaning experts to work on publishing a code section under ASME auspices. This code section, designated AG-1, was assigned to a newly formed committee, the ASME Committee on Nuclear Air and Gas Treatment (CONAGT). This Committee started work in the mid 1970’s, and has issued various sections of the code since then, and the document now totals 600 pages. This code section covers the design, construction, installation, operation and testing of air cleaning systems in nuclear facilities. Work continues on updates to these sections of the AG-1 Code, as well as new sections specifically addressing gas processing systems. ASME code section AG-1 is the main international document for nuclear air cleaning systems for safe operation of nuclear power facilities, to ensure the safety of workers, and to protect public health and safety and the environment. The Code has four divisions, and a membership of over 100 of the premier air cleaning experts from 11 different countries.

Commentary by Dr. Valentin Fuster
2016;():V002T08A003. doi:10.1115/ICONE24-60282.

This paper provides a comparison between the requirements for the qualification of welding procedures, welders and welding operators, as well as for welding supervision/coordination and welding quality according to the ASME Boiler & Pressure Vessel code (ASME BPVC Sections III and IX), the Afcen RCC-M code (France) for nuclear plant components, and ISO standards referenced to by the latter. The work was carried out within an international project that included other major codes and standards applied in the Nuclear sector, ie CSA (Canada), JSME (Japan), KEPIC (Korea) and Rostechnadzor PNAE G-7 (Russia), as part of a larger effort towards harmonization of nuclear pressure-boundary codes and standards. The comparison work aimed at identifying significant differences, as well as areas of current or potential convergence and future harmonization of the abovementioned codes and standards.

A line-by-line comparison was carried out, using the ASME BPVC as the baseline. Corresponding requirements in the different codes and standards were identified and categorized as ‘identical’, ‘equivalent’, ‘technically different’ or ‘not specified’ (when addressed by the ASME BPVC only). The most significant differences are discussed and the main findings are presented, including upcoming code and standard changes that may affect harmonization. Concluding remarks are provided with regard to the future effort required for harmonization of the codes and standards considered, through convergence and/or reconciliation.

Topics: Welding
Commentary by Dr. Valentin Fuster
2016;():V002T08A004. doi:10.1115/ICONE24-60472.

Nuclear safety criteria are based upon the concept that plant situations that are expected to have a high frequency of occurrence must not pose a danger to the public, and that plant situations that pose the greatest danger to the public must be limited to situations that have a very low expected frequency of occurrence. This concept is implemented by grouping plant situations (or events) into categories that are defined according to their expected frequencies of occurrence. It is important not to allow the boundaries of these categories to be crossed. Plant designs cannot allow low-consequence events of high expected frequency to develop into events of more serious, high-consequence categories.

The development of this system of frequency-based categorization is discussed, followed by an evaluation of various methods that have been proposed and applied to show how minor events could be prevented from becoming major events.

Commentary by Dr. Valentin Fuster
2016;():V002T08A005. doi:10.1115/ICONE24-60700.

The official report of The Fukushima Nuclear Accident Independent Investigation Commission concluded that “The TEPCO Fukushima Nuclear Power Plant accident was the result of collusion between the government, the regulators and TEPCO, and the lack of governance by said parties. They effectively betrayed the nation’s right to be safe from nuclear accidents. Therefore, we conclude that the accident was clearly ‘manmade.’ We believe that the root causes were the organizational and regulatory systems that supported faulty rationales for decisions and actions, rather than issues relating to the competency of any specific individual.

This wakeup call for the nuclear power utilities should require a public review of their relationship with of regulators. However, severe accident related risk reduction is a relatively uncharted territory and given the apparent lack of in-house technical expertise, the regulators are heavily relying on the qualitative and ‘hand waving’ arguments being presented by the utilities inherently disinterested in further investments they are not required to make under original license conditions. As a result, it has accelerated further deterioration of the safety culture and emboldened many within the regulatory staff to undertake or support otherwise questionable decisions in support of the utilities that prefer status quo. Case in point is the Canadian Nuclear Safety Commission (CNSC) which mostly accepts any and all requests by the nuclear power industry. After Fukushima, the CNSC took a year to publish a set of ‘Action Items’ for the Canadian Nuclear industry to prepare plans over 3 years and then accepted most if not all submissions that in many cases barely addressed the already watered down recommendations. In some cases the solutions proposed by the industry were economically expedient but technically flawed; and some could even be considered dangerous. CNSC also published a study on consequences of a severe accident with a source term that was limited to the desirable safety goal (100 TBq of Cs-137), which coincidently years later matched the utility ‘calculations’, but orders of magnitude smaller than predicted by independent evaluations. As a result, some well publicized conclusions on the benign nature of consequences of a CANDU severe accident were made and the local and provincial agencies that actually are supposed to prepare off-site emergency measures were left with an incorrect picture of what havoc a severe accident can cause otherwise. CNSC then published a much publicized video highlighting the available operator actions to terminate the accident early and later a report outlining the accident progression for a severe accident without operator action with conclusions that were immediately technically suspect from a variety of aspects. The aim was to claim that a severe core damage accident has no unfavorable off-site consequences. The regulator effectively, in this case, comes across as a promoter for the industry it is legislated to regulate. The paper outlines examples of actions being taken by the regulators that hinder development of effective risk reduction measures by the industry which otherwise would be forced to undertake them if the regulators had not stepped on the plate to bat for them. They vary from letters to editors to silence any safety concerns raised by the public, muzzling of its own staff, trying to silence external specialists who question their wisdom on to blatant disregard for any intervention by public they are required to entertain by law but are accustomed to factually ignore or belittle. The paper also outlines a number of examples of actions that an independent regulator would undertake to reduce the risk and enhance the safety culture. The nuclear regulatory regimes work well generally but in cases where it does not, the results can be disastrous as evident from the events in Japan and as is building up in Canada. The paper also summarizes the disparities between the number of Regulatory Actions instituted by the CNSC against small companies that use nuclear substances for industrial applications and almost none actions against the nuclear power plant utilities it regularly grants a pass in spite of the larger risk their operations pose to public.

Commentary by Dr. Valentin Fuster
2016;():V002T08A006. doi:10.1115/ICONE24-60821.

The Atomic Energy Society of Japan (AESJ) has established and issued the “Proactive Safety Review (PSR+) Guideline for Continuous Improvement on Nuclear Power Plants: 2015” 1) through the discussion at the Periodic Safety Review Subcommittee under the System Safety Technical Committee of the AESJ Standards Committee (SC), and then approved by the both committees. This guideline defines the review concept, perspectives, and methodologies, relating to identifying areas of safety improvement when the licensees implement periodic review for continuous improvement of safety of their nuclear power plants, so called “Proactive Safety Review (PSR+).

This paper discusses the details of this newly established standard, including the concept, perspectives, and methodologies, relating to identifying areas of safety improvement. In addition, specific examples of applying this standard and the policy of utilizing this standard in continuously improving the safety of all the nuclear power plants in Japan are also described.

Commentary by Dr. Valentin Fuster
2016;():V002T08A007. doi:10.1115/ICONE24-60936.

This paper describes the current status and path forward of the ongoing activities in the Japan Society of Mechanical Engineers to implement reliability evaluation methodologies into the fast reactor codes and standards. The activities are going on on two aspects; design and inservice inspection. For design, methodologies are being developed to implement a reliability-based method for the evaluation of buckling of vessels using the Load and Resistance Factor Design method. With regards inservice inspection, in order to establish a methodology that takes account of the reliability of components when setting inspection requirements, evaluations are being performed on major passive components of fast reactors. To support the process, guidelines for reliability evaluation are being developed. The efforts to standardize reliability evaluation methodologies will be continued. One of the most important remaining issues is establishing a framework on which target reliabilities could be determined by the consensus of stakeholders.

Commentary by Dr. Valentin Fuster
2016;():V002T08A008. doi:10.1115/ICONE24-60958.

“Rules on Concrete Containment Vessels for Nuclear Power Plants” was first published in 2003. It was revised in 2011 and the latest Edition was published in 2015. Endorsement of this Code is expected in the very near future.

There are two types of reactor containments: steel containment vessels (hereinafter referred to as “SCV”) and concrete containment vessels (hereinafter referred to as “CCV”). The former is addressed by Rules on Design and Construction for Nuclear Power Plants, and the latter by Rules on Concrete Containment Vessels for Nuclear Power Plants, including Reinforced Concrete Containment Vessel (hereinafter referred to as “RCCV”), Pre-stressed Concrete Containment Vessel (hereinafter referred to as “PCCV”), and hybrid containment vessels which are required further safety and reliability.

Commentary by Dr. Valentin Fuster
2016;():V002T08A009. doi:10.1115/ICONE24-60960.

Appendix N9 to AISC N690s1 presents the design provisions for steel-plate composite (SC) walls in safety related nuclear facilities. AISC N690s1 is Supplement No. 1 to AISC N690-12 specification for safety related steel structures in nuclear facilities and was published in October 2015. This paper discusses the outline of Appendix N9 as well as how the appendix can be used for the design of SC wall structures.

Appendix N9 establishes the minimum requirements that SC walls need to meet in order for the specification to be applicable. The requirements include minimum and maximum wall thickness and steel reinforcement ratio. Detailing requirements for SC wall panel sections are also discussed. The faceplate slenderness requirement to prevent the limit state of buckling before yielding is provided. Steel anchor requirements are based on developing adequate composite action, and preventing interfacial shear failure. Requirements for tie bars connecting the steel plates (faceplates) are provided to prevent splitting failure and out-of-plane shear failure. The detailing and design provisions for regions around openings in SC walls are also included. Appendix N9 provides a method of checking the design of SC walls for impactive and impulsive loads.

A discussion of the analysis requirements for SC walls is presented. The provisions include effective stiffnesses, accident thermal loading and model parameters for analysis.

The design strength equations for axial tension, axial compression, out-of-plane shear, out-of-plane flexure, in-plane shear, and for combined in-plane forces and out-of-plane moment demands are parts of the provisions of the appendix. The provisions also include interaction equations for evaluating tie bars resisting demands due to combination of out-of-plane and interfacial shear forces. Performance requirements for the anchorage of SC walls to concrete basemat, SC wall-to-wall connections and SC walls to floor slab connections are given in the appendix.

The provisions also include requirements for fabrication, inspection, and quality control of SC walls constructed for safety-related nuclear facilities.

Commentary by Dr. Valentin Fuster
2016;():V002T08A010. doi:10.1115/ICONE24-60964.

In a previous study, we proposed an evaluation method of seismic buckling probability of a reactor vessel considering seismic hazard and showed seismic load had the most significant impact on buckling probability among the random variables considered in the evaluation. It means that more rational design of vessels can be realized by taking account of seismic load variation. The load and resistant factor design (LRFD) method enables us to determine design factors corresponding to target reliability by considering variations of random variables and it is used widely in civil engineering. Therefore, in this study, we have developed a draft code case of the rule for preventing buckling of vessels in JSME fast reactor codes based on LRFD. The equation in the draft code case is almost the same as the original but every random variables, seismic load and yield stress, have their own design factors. In addition, mean or median values are used for evaluation instead of design values including conservativeness. This is the first attempt to develop a code case based on reliability evaluation and is expected to lead further implementation of reliability evaluation into JSME fast reactor codes.

Commentary by Dr. Valentin Fuster

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